IR 05000029/1988009
| ML20196D184 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 06/21/1988 |
| From: | Haverkamp D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20196D152 | List: |
| References | |
| 50-029-88-09, 50-29-88-9, IEB-80-24, NUDOCS 8807010327 | |
| Download: ML20196D184 (29) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.: 50-29/88-09 Docket No.: 50-29 Licensee No.: OPR-3 Licensee: Yankee Atomic Electric Company 1671 Worcester Road Fr.imingham, Massachusetts 01701 Facility Name: Yankee Nuclear Power Station Inspection at: Rowe, Massachusetts Inspection Conducted: April 1, 1988 - May 31, 1988 Inspectors: Harold Eichenholz, Senior Resident Inspector Cynt a A. Carpenter, Resident Inspector Approved By: 4 Mkd CM<[wI Donald R. Haverkamp, Chief 4/u /M Date ReactorProjectsSectionNo.y3C Inspection Summary: Inspection on April 1, 1988 - May 31, 1988 Report No. 50-29/88-09 Areas Inspected:_ Routine on-site regular and backshif t inspection by resident inspectors (204 hours0.00236 days <br />0.0567 hours <br />3.373016e-4 weeks <br />7.7622e-5 months <br />). Areas inspected included licensee action on previous inspection findings, operational safety verification, bi-monthly safety system verification, radiological controls, events requiring telepnone notification to the NRC, plant events, maintenance observations, surveillance observations, on-site review committee activities, plant information reports, licensee event reports, emergency exercise and part 21 report Results: One violation was identified by the inspector and involved the fail-ure to aaintain written meeting minutes for the plant operation review commit-tee in accordance with Technical Specification requirements (Section 11). The failure of the licensee to maintain the fire rating for six fire doors under all operating conditions was classified as a licensee identified Violation (Section 4.c.2).
Regarding an overall facility assessment for this inspection period, the NRC has noted a generally strong performance in plant operations and surveillance activitie This appeared to be attributable to knowledgeable licensed opera-tors and the overall notable performance and knowledge by personnel performing surveillance activities. An area that warrants increas?d licensee attention is the inconsistency between actual system line-ups and system drawings showing valve positions (Section 5).
s907010327 880621 PDR ADOCK 05000029 o DCD
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TABLE OF CONTENTS Page, Persons Contacted. . . . . . . . _ . . . . . . . . . . . . . . . . . I Summary of Facility and NRC Activities , . . . . . . . ...... 1
- Licensee Action on Previous Inspection Findings (IP 92701, 92702)* . 2
' Operational Safety Verification (IP 71707) . . . . . . . . . . . . . 2 Daily Inspection. . . ..................... 2 System Alignment Inspection . . . . . . . . . . . . . . . . . . 5 Biweekly and Other Inspections. . . . . . . . . . . . . . . . . 5 Backshift Inspection. . . . . . . . . . . . . . ........ 9 Bimonthly Safety System Verification (71710) . . . . . . . . . . . . 9
Radiological Controls (IP 71707) . . . . . . . . . . . . . . . . . . 11 Events Requiring Telephone Notification to the NRC (IP 93702) ... 11 8.' Plant Events (IP " 702, 62703) . . . . . . . . . . . . . . . . . . . 12 Diesel Fuel Oil Spill . . . . . . . . . . . . . . . . . . . . . 12 planned Load Reduction fc- Main Coolant System Inspection . .. 13 Load Reduction for No. . '. .ation Battery Replacement. ..... 14 Load Reduction to Remove No. 3 Station Service Voltage Regulator from Service. . . ................. 15 Reactor Scram Due to Fault on Transmiscion Line . ...... 17 Maintenance Observations (IP 62703). . . . . . . . . . . . . . . . . 18 1 Surveillance Observations (IP 61726, 62703). . . . . . . . . . . . . 19 1 On-site Review Committee Activities (IP 40700) . . . . . . . . . . . 22 1 Plant Information Reports (IP 90712) . . . . . . . . . . . . . . . . 24 1 Review of Licensee Event Reports (IP 90712, 92700) . . . . . . . . ' . 26 1 Emergency Exercise (IP 82301). . . . . . . . . . . . . . . . . . . . 27 15. Management Meetings (IP 30703) . . . ............... 27
- The NRC Inspection Manual inspection procedure (IP) or temporary instruction (TI) or the Region I temporary instruction (RITI) that was used as inspection l
guidance is listed for each applicable report section.
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DETAILS
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1. . Persons Contacted
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Yankee Nuclear Power-Station (YNPS)
\ N. St. Laurent, Plant Superintendent
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T. Henderson, Assistant Plant Superintendent R. Mellor, Technical Director x
rThe inspector also -interviewed. other licensee employees during the
. inspection, including. members of the operations, radiation protection, chemistry, instrument and control, maintenance, reactor engineering, security, training, technical services and general office. staff . Summary of Facility and NRC Activities At the start of the inspection period on April 1,1988 the plant was at 100*f, of rated power. The plant remained at full power until April 7, 1988 when -a _ planned load reduction was initiated to Mode 2 (Startup -Mode) in order to perform an inspection of the No. 2 hot leg loc.p isolation valve
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and the charging line check valve. The plant returned to full power on April 10,19881where it remained until April 29,1988 when a controlled shutdown was initiated to Mode 3 (Hot Standby) to replace Station Battery
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N . The generator. was phased to the grid on May 1,1988, with the plant being returned to fur power on May 10, 1988. The delay in return-ing to full power operat_ ion was due to an out-of-round shaf t on the motor of the No. 1 boiler feed pum Later the same day, May_10, 1988 the licensee commenced . an emergency load reduction to remove the plant from the grid due to the identification of a de, -iorated condition of a con-ductor on the C phase of the No. 3 station service voltage regulator. The plant was returned to full power operation on May 12, 1988 and remained at
[ ful.1 power until an automatic reactor-trip occurred on May 17, 1988. This reactor trip was initiated by a fault on one of the two off-siteitrans-l mission lines. The plant returned to full power operation on May 21,.1988 and rema,in at full power for the remainder of the perio During the week of April 11-15, 1988 an NRC Region I (NRC:RI) specialist
- inspector completed a review of eddy current inspections, wate ' chemistry l data and radiation exposure data (Inspection Report 50-29/83-07). During
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the week of April 24-28, NRC:RI emergency preparedness inspectors were
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on-site to observe, with the aid of the resident inspectors, the annual l' emergency preparedness exercise (Inspection Report 50-29/88-08). The j Deputy Director of the NRC:RI Division of Reactor Projects toured the l plant site on May 18, 1988 and held discussions with licensee management ;
-representatives concerning items of mutual interest. On May 23, 1988, the NRC Systematic Assessment of Licensee Performance (SALP) Board was held to review licensee activities for the period of October 7, 1986 to March 31, 1988 (SALP Report 50-29/86-99).
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3. Licensee Action on' Previous Inspection Findings .(0 pen) Violation 87-16-01: Failure to make a -timely notificatio This item concerned the licensee's failure, to make a timely 10 CFR
. _50.72 notification, in that the NRC was.not notified within one hour
, ~o f the initiation of plant shutdown as required by Technical Specifi-cation.(TS) 3. Licensee corrective actions in response to this violation have in-cluded, or is planned to include, operator re-training on reportable event determin.tions and notifications, review of the violation response in the operator training segment conducted -between March 22,'1988 and Agvil 22, 1988, and . implementation and training for a new procedure on potential - repcrtable events. In its letter FYR 88-33 to the NRC dated March 14, 1988, the licensee committed to have completed the above actions by May 27, 198 During a subsequent discussion with the resident inspector, the licensee requested and was granted a five-week extension to July 1,1988 for completion in implementing the procedure on poten-
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tial reportable events and training on this procedure. The extension was necessary to fully develop the procedure and properly train the operator The licensee is in the process of developing procedure AP-0008, Event Reportability Evaluation Process, _ to provide a sys-tematic evaluation process for determining the necessary immediate notification and written reports required as a result of any abnormal plant occurrenc The licensee has been responsive to prior NRC concerns involving the need to inform the NRC when a corrective action extension is required. This item remains open pending review of the licensee's implemented corrective action . Operational Safety Verification Daily Inspection During routine facility tours, the inspector checked the following items: shift manning, access control, adherence to procedures and limiting conditions for operations (LCOs), instrumentation, recorder traces, protective systems, control rod positions, containment tem-perature and pressure, control room annunciators, radiation monitors, radiation monitoring, emergency power source operability, control room and shif t supervisor log, tagout log and operating order No inadequacies were identified except as noted belo ~
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(i)' On April 1, 1988 the l'icensee identified a procedural inadequacy L involving a steam generator tube rupture event, in that the No and 3 steam generators would become cross connected through their respective emergency steam supplies to the steam driven emergency feed water pump. The procedure OP-3107, Rev. 17, Steam Generator Tube Rupture, failed _ to provide. adequate
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instruction for proper isolation of the affected Nos. 2 or 3 steam generators following a tube rupture. Specifically, the procedure did not provide for shutting the respective isolation valves on 1-1/2 inch steam supply lines for the steam driven emergency boiler feedwater pump (EBFP).
During a plant shutdown on March 22, 1988 the plant was being maintained in hot standby and the licensee was utilizing the emergency atmospheric steam dump (EAS0) valves to maintain main coolant system temperature and pressur At that time, the plant operators noticed that venting of either the Nos. 2 or 3 main steam lines. affected the other steam generato In the event of a steam generator tube rupture, licensee procedure OP-3107 directed the operators that once the ' faulted steam generator has been identified, to close the NRV and stop all feedwater flow on the faulted loo A rapid cooldown and depressurization of the main coolant system is performed by either 1) using steam dump to main condenser or 2) atmospheric steam dumps from the unfaulted loop Consequently, on a tube rupture in the Nos. 2 or 3 steam gener-ators, they would be essentially cross-connected through the common cmergency steam supply to the steam driven EBFP. Wnen the EASD valve was opened to help initiate main coolant system cooldown on the unfaulted steam line (either Nos. 2 or 3), main steam from the faulted steam generator could backfeed through the penetration provided for the EASD and EBFP common header and be vented to the atmosphere via the unfaulted steam generator EASD valv Immediate corrective action by the licensee consisted of issu-ance of a proposed change notice (PCN) to procedure OP-310 The PCN added steps that if the Nos. 2 or 3 steam generator is faulted, the appropriate steam supply valve (MS-V-693 or MS-V-694) to M.a steam driven EBFP must be closed to prevent spread of .adioactivity to the environmen The licensee reviewed t',e system design basis and determined that the revised
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procedural steps were appropriate and consistent with the system design. The licensee has issued LER 50-29/88-04 as a result of this even The inspector reviewed the licensee's procedure revision to OP-3107 and had no further question . _
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(2) _0n April 8,1988 the inspector observed control room operations during removal of the generator from the grid and subsequent operations later that day to phase the generator to-the grid and return the plant to full power operation During performance of both of these plant evolutions, as well as maintaining the plant in Mode 2 (startup mode), good communi-
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cations were evident. between the primary and secondary reactor operators. Additionally, the operators and licensee staff were very knowledgeable of plant conditions and systems, plant per-formance and licensed duties. The inspector noted that appli-cable operating procedures were in use by the operators. Remov-
~ing the main turbine and generator from _ service and the subse-quent return of the plant to power operation was observed to be exceptionally smooth and uneventful. Good oversight of plant operations and conditions by the shift supervisor and senior control room operator was clearly eviden Although returning the piant online occurred on the third shift (3:49 p.m.) the second shift (day shift) held over to accomplish this evolution in order to provide continuity, illustrating the same high level of performance demonstrated by the operating staff throughout the day. Strong interdepartmental cooperation was noted between operations and reactor engineering. Given the nature of the plant design, and the required complex set of manual operations involving both the primary and secondary side
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of the plant, the licensee's operating staff demonstrated a high level of effectivenes (3) Inspection . Report 50-29/88-05, Section 7h discussed the cause of inoperable primary vent stack (PVS) process radiation mon-itoring equipment on March 23, 1988, including the high range -
monitor. Investigation revealed that the heat tracing on the system piping had been de-energized when plant personnel turned off the controller due to unfamiliarity with its operation The licensee prepared a PCN to procedure OP-8042, Radiation Protection Shift Personnel Duties and Surveillances. This PCN added a step to the procedure for personnel to observe the PVS heat trace controllers, defined the normal condition of the lamps on the controllers and action to be taken if any other lamp configuration existe Licensee corrective actions to define cff-normal conditions and ensure timely response to equipment problems, especially when equipment failure results, was noted to be prompt and indicative of a conservative attitude toward ensuring the operability of equipment that provides emergency assessment capabilit _ ._, _- _ _ ,. _ _ _ _ _ _ - ._ _ _ -. _ . .
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., System Alignment Inspection-Operating confirmation was made of selected piping system train Accessible valve positions and- status were examined. _ Power supply and breaker alignments 'were checked. Visual inspections of major
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components were- performed. Operability of instruments essential to system performance was assesse The following' systems were checked during plant tours and control room panel status observations:
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Low pressure and high pressure injection systems
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Non-return valves
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Charging system
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Emergency diesel generator units No inadequacies were identifie Biweekly and Other Inspections (1) General Facility Observation During plant tours, the inspector observed shift turnovers, com-pared boric acid tank sample analyses and tank levels to Tech-nical Specif_ications requirements, and reviewed the use of radiation work permits and radiation protection procedure Area radiation levels and air monitor use and operational status were reviewed. Verification of tagouts indicated that actions were properly conducte No inadequacies were identifie (2) - Fire protection and Housekeeping No inadequacies were noted regarding licensee housekeeping or fire protection practices. A strong commitment to proper house-keeping conditions and practices by the plant staff is routinely observed by the inspector. Performance in this area continues
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to be viewed as a licensee strengt On April 1, 1988, an evaluation of six penetration fire barriers protecting safety-related areas resulted in the doors being declared degraded. The fire rating degradation resulted when the doors were modifieo to provide access control; this modification allowed the doors to unlatch on a loss of electrical power to the doors' latch mechanism This condition compromised the doors' Underwriters Labora-tory fire rating. The licensee declared the six penetra-tion fire barriers to be non-functional and established a continuous fire watch on at least one side of the affected penetrations within one sour as required by TS 3. 7.11.
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The door hardware was changed the following day, the doors were satisfactorily tested. under a loss of power condition
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and the continuous fire watches secured. The licensee ~
attributed the event to an inadequate review of the possi-ble door failure modes during their preparation of"the fire hazards apolysi.s. The loss of power failure was not spec-ifically identifie The operations department issued Special Order No. 88-38 on April 2,1988, which informed
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the operators of the. manner in which they are to ' override the secured condition of the doors during a loss of power even The inspector had no further questions of the licensee on this ite The licensee issued Licensee Event Report (LER 50-29/88-05)
for this event on April 30, 1988, which accurately de-scribed the event, specified the root cause, and documented their corrective actions. Based upon the licensee's self-identification of the condition, the responsive development of corrective measures, and the determination that the event was reportable will enable the inspector to treat this item as a licensee-identified violation. In accord-ance with the provisions of 10 CFR 2, Appendix C a notice of violation will not be issue \
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As a result of surveillance testing activities on April 6,1988, the operations department became aware of a condition involving one of the halon system's storage bottles being 'two pounds lo The action statement of TS 3.7.10.5 was immediately entered, which resulted in the .
establishment of an hourly fire watch patro In addition, Special Order No. 88-39 was issued which stipulated what backup fire suppression equipment was to be used by the fire watc The inspector noted that the special order also contained a fire fighting strategy document for the switch gear reom. The 'iicensee's actions are indicative of the conservative manner in which they approach fire pro-tection issues and reflects positively on their commitment to maintaining fire protection and prevention programs at a high leve (3) Observations of Physical Security
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On April 7, 1988, the licensee received safeguards mate-rials that did not appear to be properly package Upon arrival at the local post office and while sorting the mail the postal clerk discovered that a safeguards document and the inner envelope used to contain the safeguards document
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had come out of its larger, outer envelope, _The postal- '
clerk placed the safeguards document and inner, marked envelope back inside the outer envelope and placed a Postal Service sticker on the; outer flap and initialed the sticker to indicate that the envelope had been found open with the contents not insid The licensee conducted an immediate investigation to deter-mine the cause of the event. Interviews were conducted with the postal clerk and involved licensee personnel. The safeguards document was sent from corporate headquarters to the plant; at corporate headquarters, 'the document was-placed in an envelope, properly marked and sealed with two pieces of . tap The licensee also verified that all original'pages of the document were accounted fo A one-hour ENS call was made to the NRC in accordance with procedure AP-0009, Reporting of Safeguards Events. This procedure requires a one-hour notification be made in the event of loss, thef t or compromise of safeguards informa-tio The licensee will continue their investigation of this event to include follow-up with the U.S. Postal Ser-vice to. determine when the package became unsealed and will document their long-term corrective actions in their 30-day report to the NR During a routine - tour of the turbine building on May 19, 1988, an auxiliary operator discovered what he believed to be an off-normal position for the control switch of a non safety-related secondary side No. 2 con-denser vacuum pump. .Until the licensee's staff could con-duct a thorough investigation, the event was considered to be susoicious in nature and treated as a potential tamper-ing even Within one hour of discovery of the condition, the licensee made a one-hour ENS notificatio Following a lengthy investigation, the licensee's staff was able to conclude that tampering with the switch had not occurred, but was the result of a misunderstanding by the operator as to the l as-found condition. The event classification was rescinded
- later the same da Upon discovery of the possible of f-normal position of the switch, the licensee initiated its compensatory measures, l as described in security-related document Security per-
! sonnel were involved in all aspects of the investigation l and management oversight was aggressiv Appropriate
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compensatory measures were initiated.
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The.-licensee has two documents that provide guidance to its personnel in dealing with suspected tampering against plant equipment. The first document is OP-Memo 2T-5, Rev. O, Response to Events Concerning Deliberate Acts .or Suspected
. Tampering Against Plant Equipment, and th'e second is pro-cedure AP-0404, Rev. 11, , Emergency Security Procedure The inspector noted that.both documents provide examples of deliberate tampering that include, but are not limited to, events that involve equipment important to safet The discovery on May 19, 1988, of what was believed' to be an of f-normal position of a switch, involved plant equipment that had no relationship to safety and did. not really . fit the criteria of either documen However, other recent events and an appropriate awareness of the importance of physical security at the plant had resulted in the opera-tions department personnel classifying the event in a con-servative manner. The licensee is reviewing its existing guidance, and adjusting its response accordingly, should there be a need in the future to respond to suspected tampering event No unacceptable conditions were identified by the inspector as a result of reviewing the licensee's response to this even In fact, the plant personnel demonstrated a healthy questioning attitude about a potential off-normal equipment condition and a proper level of concern for physical security aspects of plant operatio .
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The security off4cers of the licensee's contracted security service, Green Mountain Security Service (GMSS), are repre-
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scnted by the Independent Security Union, Local The union and GMSS ratified a new labor agreement that extends to the remainder of the GMSS contract with the license However, in a May 10, 1988 letter, GMSS gave the licensee notice that as of September 17, 1988, they will terminate their contract to provide security services for the licen-se The licensee is actively pursuing obtaining a replacement security contracto Improvements in station management oversight of security activities continues to be observed by the inspector. Items that reflected an improving trend in their concern for security were:
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(a) Two security shift supervisors wil1~be sent.in June, 1988.to an NRC/00E two week tactical firearms training program at the DOE Central Training Academy~ in Alburquerque, .(b) Participation in the NRC Region I Security Association (c) 'The initiation of .a weekend -contingency drill program for security and operations personnel (d) All security force personnel received bomb threat /
search training at a seminar conducted onsite by per-sonnel of the Explosives Ordinance Department located at Ft. Devins, M (e) The' Security Manager was . weapons qualified in accord-ance with the licensee's training and qualification
. pla ( f)' 'As part of the on going efforts involved with the security department's reorganization, a technical
, assistant for the department was hired, and licensee and security contractor offices were relocated tc more appropriate location (g) A recru. ment campaign was initiated to address NRC and licensee concerns- associated with a high level of -
security force attrition and maintaining the security force in a fully staffed capacit Backshift Inspection
,. The inspector coe ducted backshift, weekend or holiday inspections on April 13, 26, 30, and May 5, 7, 8, 15 and 1 Operators and shift supervisors were attentive and responded appropriately to annuncia-tors and plant condition No violations were identified during backshift inspectio . Bimonthly Safety System Verification The inspector independently verified the operability of a selected engi-neered safety features (ESF) system by performing a complete walkdown of the accessible portions of the system to:
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confirm that the licensee's system lineup procedures match plant drawings and the as-built configurations;
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identify equipment conditions and items that might degrade performance;
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. verify appropriate levels of cleanliness were being maintained;
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verify technical specification requirements are adhered to;
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verify instrumentation lineup and calibration; and
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verify proper valve position, availability for function and position-indicatio .
The steam-driven emergency feedwater system was reviewe The inspector had the following comment During review of Procedure OP-4211, Rev. 22, Emergency Feedwater System Operability Test, the~ following discrepancies were 'note (1) Attachment A, Step ~ 38 a(3) checked open AS-TV-669, nuclear auxiliary steam cross connectio The next step, step 38a(4)
checked closed AS-V-669, nuclear auxiliary steam cross connec .(2) Attachment C, Step 37 a(6) checks open HC-V-813. The next step, step 37 a(7) checks closed HC-V-81 (3) Attachment C, Step 37 c(3)(b) closes AS-V-760
.(4) The procedure closes valves HC-V-827, ,HC-V-813 and HC-V-828 In the first two examples, incorrect valves were.' specified to be checked either open or closed. In example (1), the open valve should be AS-V-669 (the "T" is a typo) and the closed valve should be AS-V-75 The licensee was aware of this discrepancy and had
~ initiated a procedure -change notice (PCN). In -example (2), step 37
- a (6).should have specified valve HC-V-812. In example (3), both the P drawing (YAEC 9699-FM-3A) and attachment A state t. hat the valve is to be "locked" closed. Attachment C should be changed to maintain con-sistency and to ensure the valve is properly locked for operation The operations department was unaware of these two items and will add them to the PC The next example, number (4) above illustrated procedural inconsis-tencies with plant drawing In these three cases, the procedure stated to close these valves; however, plant drawings (YAEC 9699-FM-3A and 9699-FB-7A) showed these valves to be ope Plant system walkdown showed the valves to be in the correct position per the procedure but incorrectly positioned in accordance with plant draw-ings. This is also true of valve DW-V-793. In order to alleviate possible future confusion, plant drawings should accurately reflect normal at power system line ups. This type of condition has been the j
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subject of ongoing NRC concern Inspection Report 50-29/84-20 docu-mented in Open Item 50-29/84-20-08 the need for the licensee to cor-rectly translate the applicable design basis into drawings, and indi-cated that the licensee will review its practices accordingly. Addi-tional management attention is warranted to resolve this issue .in a timely manner. Finally, although the licensee has already begun a program to re-tag ' valves, .several valves, particularly in the Auxiliary Boiler Room, were found with-no tag No violations were note . Radiological Controls Radiological controls were observed on a routine basis during the report-ing period. Standard industry radiological work practices and conformance to radiological control procedures and 10 CFR Part 20 requirements were observed. _ Independent surveys of radiological boundaries and random sur-veys of nonradiological areas throughout the facility were taken by the inspecto During tours of the Primary Auxiliary Building, the inspector noted that, in the cubicles that contain the various pumps (purification pumps, safe shutdown pump, etc.) signs have been erected in a low radiation dose area that read: "Think, work ALARA; LOW dose rate area here" with arrows pointing down. These are large white signs with red lettering, posted' on one of the four walls, directing the workers to think about keeping radia-tion exposure as low as reasonably achievable and illustrating clearly the lowest radiation dose area in the cubicl These signs illustrate a positive trend toward improving radiological controls and a meaningful licensee effort to keep worker radiation exposure as low as possibl . Events Requiring Telephone Notification to the NRC The circumstances surrounding the following events, which required NRC notification via the dedicated ENS-line, were reviewed. A summary of the inspector's review findings follows or is documented elsewhere as noted below:
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At 10:26 a.m. on April 1,1988, the NRC was notified in accordance with 10 CFR 50.72 (b)(2)(iii)(C) that an inadequacy existed in the emergency response procedure for a steam generator (SG) tube rupture even This condition is discussed in Section 4a(1) of this repor At 6:25 p.m. on April 1, 1988, the NRC was notified in accordance with 10 CFR 50.72 (b)(2)(vi) that approximately 100 gallons of diesel fuel oil was spilled into adjacent stream This event is discussed in Section 8a of this repor , , .- . - -. - . _ . - . - - . , - _ _ - , .
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1 At '12:37 p.m. on April- 7,1988, the 'NRC ~was ' notified in accordance
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with 10 CFR 73.21' (b)(1)(viii) 'that a possible compromise of safe-
. guards information may have occurred. -This event is discussed in Section 4c(3) of this repor At 6:30 on April 30, 1988, the NRC was notified in accordance with 10 CFR '. 50.72 (b)(2)(ii) of an inadvertent actuation of the pressurizer power operated relief valve for two seconds while per-forming - a' surveillance' of main coolant flow channel The block valve, however, was close This event is . discussed further in Section 10 of this repor At 1:20 a.m. on May 18, 1988, the NRC was notified in accordance with 10 CFR 50.72(b)(2)(fi) of a reactor scram caused, in part, by an apparent fault on one of the two off-site transmission line This event is discussed further in Section 8e of this repor At 6:35 a.m. on May 19, 1988, the NRC was notified in accordance with 10 CFR 73.71(b)(1) of a possible tampering event. After thorough investigation by the licensee, this ENS call was rescinde The inspector's review of licensee actions and event details are contained in Section 4.c(3) of this repor . Plant Events
- (a) Diesel Fuel Oil Spill On ' April 1,1988, the licensee discovered that approximately 100 gallons of diesel fuel oil had been inadvertently spilled from a small fuel oil tank used for fire training in the upper parking lo The licensee considers that the probable cause of' the spill was due to a vehicle backing into the tank, damaging the tank transfer valve and piping assembl The oil was confined to the grouid around the tank and the ground in contact with run off water but did not get into the river downstream of the plan Water samples were also take The Massachusetts Department of Environmental Quality Engineering (DEQE) and- the Environmental Protection Agency (EPA) were notified. The NRC was notified in accordance with 10 CFR 50.72 (b)(2)(vi) because of the licensee's notification to other government agencies. The inspector notes that the ENS call to the NRC due to the licensee's notification of other government agencies was done promptly, indicating a proper assessment by the operators of NRC notification requirement Plant Information Report 88-4 was issued concerning this incident.
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(b) Planned Load Reduction for Main Coolant System Inspection
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On April 7,1988, the licensee commenced a planned load reduction to take the plant off line in order to perform a thorough hot leak inspection of the main coolant system (MCS).
At the close of the last. inspection pericd, the plant had tripped and was in hot standby twice in- the period of March 22-26, 1988. .When the plant- returned to full power operation, the licensee noted increased unidentified leakage in the vapor container (VC) as a result of increased levels in the vapor container drain tank (VCDT),
increased VC air particulate activity as seen on the main coolant system leakage air particulate monitors (MCSLAPMs) and increased unidentified leakage as determined from' the daily water balance The licensee continued to monitor the above parameters . closely and trended the number of gallons of water dumped from the_ VCDT, the highest MCSLAPM reading and the amount of unidentified leakage on a daily basis. The MCSLAPM's are designed to detect a one gallon per minute (gpm) leak rate from the MCS, Most of the particulate radio-activity detected in the ' event of MCS leakage will be due to rubidiurn-88, the daughter of krypton-8 TS 3.4. limits the unidersified MCS leakage to 1 gpm. Trending of the above data showed that with the plant at full power, the highest air particulate activ-ity remained at - approximately 180,000-200,000 cpm, then began to decrease to approximately 150,000 com. Unidentified leakage was approximately .20 gp Chemical analysis of the leakage collected in the VCDT showed it to
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be approximately two-thirds main coolant leakage and one-third secondary leakag The licensee also performed several VC inspec-tions while at full power to determine the possible causes of the unidentified leakage. The licensee was aware of two valve leaks in
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containment: a hot leg main coolant isolation valve (MCIV) body-to-bonnet leak in loop No. 2 and the charging line check valve (CH-V-611A) located in the No. 4 loop, which had a cap lea Continued monitoring and trending by the licensee demonstrated that the unidentified leakage remained well below TS limits and that the leakage and air particulate levels were not becoming worse but rather showed improvement. The licensee considered the increases to be due to plant thermal cycling; the above air particulate limits increased significantly following the plant shutdowns the previous wee .
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However, on April 8, the licensee decided, as' a precaution, to per-
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form a thorough hot leak inspection to ensure that- the leaks on the Nos. 2 and 4 loops were as they ~ expected, and to ensure no other problems / leaks were present. The plant was removed from the grid and
,_ maintained in Mode 2 while a thorough inspection of the VC and piping was performed. The inspection of the VC and loop piping showed no additional leakages other than those previously identifie During the hot leak inspection, the licensee also inspected the bolts on the hot leg MCIV in loop No. 2 due to concern over boric acid corrosion of components. The licensee received Generic Letter 88-05, Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Com-ponents in PWR Plants, and due to the known leakages in containment, were concerned about possible boric acid corrosion of the bolting on the MCIV's. The material composition of the MCIV bolts was unknown; the valves were manufactured by Westinghouse and the bolting material was considered proprietar Therefore, the licensee does not have drawings identifying the material of these bolts. Due to known leak-age occurring in containment and the issuance of the generic letter, the licensee inspected the MCIV bolting to confirm that no boric acid corrosion of the bolts was occurring; bolts were subsequently deter-mined to be 316 stainless steel, which is neither a low alloy steel or carbon steel and is therefore not susceptible to the generic letter concern The inspector notes the conservative precautionary approach taken by the licensee- with respect to the increased unidentified leakage, increased air particulate readings and inspections of the MCIV's over concern for boric acid corrosion. Licensee philosophy is to operate with no system leaks if possibl (c) Load Reduction for No. 3 Station Battery Replacement On April 29, 1988, the licensee commenced a planned load reduction to hot standby in order to install the new station battery and perforc other maintenanc While the plant was maintained in hot sti.ndby, the licensee performed another thorough hot leak inspection of the VC and loop piping. Inspection revealed several hot stem leakoffs on the Nos. 2 and 3 main coolant loop bypass valves. The hot leg MCIV in loop No. 2 and charging line check valve CH-V-611A located in loop No. 4, which were identified as leaking previously and dis-cussed in 8b above were also inspecte The inspection indicated that there was no visible degradation apparent. Trending of uniden-tified leakage has shown the unidentified leakage to have decreased significantly, as well as the air particulate levels since the problem was notice , -
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,, Finally, a small leak was identified in the No. 2 VC air cooler at a welded connection. The air cooler is part of the vapor containe atmosphere control system, which maintains the average VC temperature and is used to cool containment air during operation in the warmer month The cooling coils are supplied with service ' wate The licensee reviewed IE Bulletin No. 80-24: Prevention of Damage Due to Water Leakage Inside Containment (October 17, 1980, Indian Point 2 Event) and their response to the NRC with respect to this bulleti The licensee committed in letter FYR 81-7 dated January 5,1981, to notify the NRC of any significant service water system leaks within containment via a special LER (24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with written report in 14 days). The licensee also noted that a 'significant leak would be considered any leak in excess of~one gallon per minute. The inspec-tor reviewed the licensee's response to IE Bulletin No. - 80-24 and concurred with the licensee that a special LER was not warranted in this case since the VC air cooler leak was less than 1 gallon per minut The reactor was brought critical on April 30, 1987. However, during the turbine startup, problems were encountered with the No. 3 and 4 turbine control valves. Troubleshooting by a Westinghouse service representative revealed a blown gasket in the turbine overspeed trip oil line, decreasing the oil pressure in the line to the No. 3 and 4 servomotors. Approximately 50 psig is required to open these valves, but due to the blown gasket in the overspeed trip line, only approxi-mately 27 psig was availabl This blown gasket acted as an over-speed tri The problem was fixed and plant startup continue Return of the plant to full power was delayed a week due to the shaft on the No. 1 BFP motor being out of round. The pump was sent to a manufacturer off site for repair. The pump was re-installed on May 9 and the plant was back at full power on May 10, 198 Licensee actions during this shutdown were noted to be conservative and aggressive in their approach to safety-related component d. Load Reduction to Remove No. 3 Station Service Voltage Regulator ,
From Service On May 10, 1988, the licensee noted that the bus conductor to the "C" phase, No. 3 2400 Vac station service voltage regulator was in a degraded condition and initiated an emergency load reduction from 100*; power. At approximately 60*4 of rated power, the emergency load reduction was terminated and the licensee continued with a normal load reduction until load was reduced to < 30 MWe. TS 3.8.2.1 allows two 2400 Vac buses to be cross tied and two 480 Vac buses to be cross tied when the plant is < 30MW . .-
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In order to commence repair to the- bus conductor on the No. ,3 2400
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Vac bus voltage regulator, the voltage regulator was required to be isolated from all power supplies. By isolating the 'No. 3 voltage-regulator, power to the No. 3 2400 Vac bus and 480.Vac bus 6-3 would be interrupted unless the buses were cross tied with the No. 1 2400 Vac . bus and 480 Vac bus 4-1, respectively. As stated above, The TS allows-this condition only when the plant is < 30 MWe. The resulting
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a condition was that the 2400 Vac buses 1 and 3 were cross tied and 480 Vac buses 6-3 and 4-1 cross tied. This would rescit in the config- '
uration where:480 Vac buses 6-3 and 4-1 and emergency buses 1 and 2 all are fed through the No. 4 station service transformer (SST). In LER 50-29/86-07, the licensee identified a concern in meeting the minimum starting voltage requirements for the high pressure safety in,iection (HPSI) pump motors when 480 vac buses No. 6-3 and 4-1 were cross tied. The licensee indicated that the SST which feeds the two tied buses could exceed its rated capability if a safety injection actuation signal (S1AS) occurred when the plant was in 'this configuratio '
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Specifically, if 480 Vac buses 4-1 and 6-3 are tied together and fed from one SST and if an SIAS would occur with these heavily loaded buses, low pressure safety injection pumps Nos.1 and 2 motors will successfully start and operat However, due to the initial heavy loading and the addition of LPSI pump loads, a _ low voltage level at 480 Vac emergency buses 1 and 2 will not allow starting of the high pressure safety injection pumps Nos. 1 and 2. The circuit breakers which tie the 480 Vac normal buses to their associated emergency buses would have. opened on degraded voltage- and the safety loads
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would be restarted by the diesel generator Although this would result in the successful . operation of the safety loads, the sequen-cing of the loads would be delayed due to the starting and loading of the diesel As s :ated above, in order to isolate and repair the No. 3 voltage regulator, 480 Vac buses 4-1, 6-3 and emergency buses 1 and 2 all would be fed through SST No. 4, raising the same concern as that in the LE The licensee initiated the same action as originally pro-posed as an interim measure in the LER; that is, a dedicated licensed operator was stationed in the control room near the emergency diesel generator (EDG) board. His sole responsibility was to open breakers BT1A and BT1B (dropping off emergency bus No. 1) if an SIAS was received. This would prevent exceeding the capability of SST No. 4 and allow the loads on emergency bus 1 to be supplied solely by its respective EDG. To perform the planned action, the Manager of Oper-ations (M00) issued Directive 88-2 to the Plant Superintendent de-scribing the directed sequence in plant conditions and electrical lineups, and compensatory operator actions. These instructions re-flected input from YNSD engineering. The existence and use of the M00 Directive is representative of the active involvement of off-site management in site activities, and ensures that decision making has adequate management revie _ - - -- -
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The inspector followed licensee actions with regard to its proposed
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actions', ensured that ..TS requirements were being met, and verified that appropriate procedures were developed and in place for' operator use during the off-normal electrical -lineup. Tne inspector reviewed the licensee's intended corrective actions and bus cross tying plans with the NRC:NRR project manager and NRC:RI management. No concerns
'were identified by the NRC with regard to the licensee's response to t.hi s event. This event, although reactive in nature, demonstrated the licensee's ability to properly assign priorities and develop well-stated procedures to control important activitie Reactor Scram Due to Fault on Transmission Line At '11:24 p.m. on May 17,1988, a fault on one of the two off-site transmission lines occurred (the Z-126 Harriman line). The center 2400 Vac bus (bus #1) was also observed to have been lost. An auto-matic reactor trip subsequently occurre The plant .was maintained in hot standby (Mode 3) while the licensee conducted troubleshooting-to determine the exact cause of the reactor trip and the nature of the fault on the transmission lin The licensee determined that a fault occurred on the' transmission line, but the exact cause of th9 fault could not be determine The transmission line was returned to service at 11:30 p.m. the same da Preliminary investigation by the licensee indicated that the fault on the transmission line may have caused the 2400 Vac supply breaker for the generator static exciter system to trip open, in response to the fault condition, resulting in a turbine overspeed condition and opening of the reactor trio breakers in response to closure of the turbine control valves. Following the licensee's preliminary inves-
-tigation, the licensee commenced reactor startup, phased the gener-ator to the grid at 3:36 a.m. on May 19 and maintained minimum power level to facilitate further troubleshooting on the static exciter. A vendor field engineer was . called to perform this additional trouble-shooting; no equipment problems with the : static exciter were foun The licensee removed the plant from the grid temporarily to remove the static exciter from service and place the rotating exciter in service. Also, the 2400 Vac feeder breaker to the static exciter, which was also suspect was removed from servic The licensee sus-pects that this breaker did not function properly by opening in response to the transmission line faul The plant was re phased to the grid and reached full power operation on May 21, 198 Addi-tional concerns were raised by the licensee due to the manner in which the static exciter responded to the faulted line conditio Because a loss of field trip condition was not sensed within the time l
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f rame . o f initiating conditions, the automatic crost tie feature-between two .of the three 2400 Vac buses did not occur prior to th reactor scram. Following the plant trip and loss of a transmission line, without the automatic cross tie : operation, only one main cool-ant _ pump- (MCP) was in operation. The plant operators secured the re.naining running.MCP and used natural circulation cooldown until the
. electrical' buses could be normalized and the MCP's restarte The reactor plant responded as expected and no deficiencies were identified. On May 20, 1988, the YNSD issued an evaluation that con-cluded that the event did not represent an unanalyzed condition
.(i.e., reactor trip followed by three of fo'ur pump loss of forced main coolant flow) since it is bounded by the complete loss of flow analysis. Management involvement was evident and the licensee main-tained a conservative approach throughout this event in that they did not want to resume. full power operation until the cause of the plant trip had been determined, -in order to preclude another plant tri The licasee has identified' the need to submit an LER (50-29/88-08)
for this event. The inspector will conduct further event fo! low up and review of licensee corrective actions as part of the LER revie . Maintenance Observations The inspector. cbserved and reviewed maintenance and problem investigation activities to verify compliance with regulations, administrative ano main-tenance pocedures, codes and standards, proper QA/QC involvement, safety tag use, equipment alignment, jumper use, personnel qualification, radio-logical controls for worker protection, fire protection, retest require-ments and reportability per Tecnnical Specification The following activities were included:
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'MR 88-765; #2 VC Cooler leak at welded connection
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MR 88-610; Security /firedoor 213 change door hardware
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MR 88-611; Security /firedoor 405 change door hardware
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MR 88-612; Security /firedoor 408 change door hardware r
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MR 88-613; Security /firedoor 409 change door hardware
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MR 88-828; #3 Station Service Voltage Regulator, Phase C-Hot conduc-tor off of regulator termir,al. Discoloration bird caging and broken strands on conductor
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MR 88-745; MS-PS-11 main steam line pressure switch replacement ,
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MR 88-746; MS-PS-12 main steam line pressure switch replacement
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MR 88-747; MS-PS-14 main steam line pressure switch replacement
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MR 88-766; No. 4 loor hot leg drain packing leak
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MR 88-767; No. 3 loop flow detector high side isolation packing leak
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MR 88-768; No. 3 loop flow detector inlet packing leak
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MR 88-771; Bleed line stop valve hot stem leakoff
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> , 19 Based upon a review of licensee activities in ' this area, the inspector noted the following:- , ._
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Regarding MR 88-765, the inspector's review of the VC cooler leak is
' documented in-Section 8c of this repor The inspector's review of' activities associated with MR's88-610, 88-611,88-612 and 88-613 are contained in Section 4c(3) of this repor .
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The : inspector reviewed the activ'ities associated with MR 88-828 in Section 8d of this repor The events associated with MR's88-745, 88-746 and 38-747 are ;
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detailed in Section 10 of this repor The inspector reviewed MR's88-766, 88-767,88-768 and 88-771 during the inspection reviewed in Section 8e of this repor . Surveillance Observations The inspector observed tests or parts of tests to assess performance in accordance with approved procedures and LCOs, test results (if completed, '
removal and restoration of equipment, and deficiency review and resolution. The following tests were reviewed:
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OP-2677, Re , Attachment 8, Routine venting of Safety Injection (SI) Header Pressure after SI Pump Operations
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OP-4204, Rev. 35. Test or Special Operation of the Safety Injection Pumps and determination of ECCS Subsystem Leakage
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OP-4244, Rev.10, Fire Suppression System Flush
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OP-4611, Rev. 19, Nuclear Instrumentation and Reactor Protection !
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OP-4656, Rev. 6, NRV Pressure Switch Functional Test of the NRV Main ,
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Based upon a review of licensee activities in this area, the inspector noted the following:
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On April 5,1988, Automatic Switch Company -(ASCO) _ disassembled four
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- 87-16, 50-29/88-02 and 50-29/88-05 for further details) to determine the cause(s) of the pressure switches failing to actuate and/or the
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excessive setpoint drif These pressure swithes had been installed approximately twenty-five months prior to being replace The pressure switches contain a cast polyurethane disc inside the pressure switch cylinder but underneath the pisto This poly-urethane disc acts as a buffer to protect the process sensing viton diaphragm from adverse action of the piston. Visual-inspection of the pressure switches revealed that the polyurethane disc had extruded into the area between the piston and the cylinder walls, severely inhibiting movement of the piston. Indications are that this extrusion of the disc has caused the' increasing and excessive dead-band being experienced. Also, on the two pressure switches that failed to actuate in December, 1987 (Inspection Report 50-29/87-16)
the polyurethane disc was found extruded around the piston and cylinder wall The degree of extrusion varied from switch to switc Further testing is to_ be performed by ASCO. The condition of the polyurethane disc being found. extruded around and on top of the piston is the same problem as that found during visual inspec-tions of previous pressure switches that failed in 198 .
The marufacturer is of the opinion that the cause could be due to the pressure swi tch sensing a constant pressure continuousl It was :
p recommended that the pressure switch be cycled more often by com-l~ pletely removing the pressure. Currently, surveillance of the switch i is performed by relaxing the pressure only until a trip condition occur The licensee plans the following corrective actions:
1) replace the remaining six main steam line pressure switches as soon as practicable; 2) have the manufacturer evaluate the poly-urethane material for suitability and perform dimensional analysis on r the piston and cylinder; 3) evaluate completely removing pressure to the switch during surveillance testing; 4) increase surveillance intervals to biweekly until further evaluations indicate proper actuation of the pressure switches; 5) cycle each pressure switch three times during each surveillant.e and 6) inquire as to the type of pressure switches used by other licensee The licensee had issued on April 26, 1988 a Notification of Potential Existence of a Design Defect ir. accordance with 10 CFR 21.21 due to this problem. During the plant shutdown on April 30, 1988 three of the remaining six pressure switches committed to be replaced by the licensee in the Part 21 notification were replaced. The last three pressure switches will be replaced when aoditional switches are received from the manufacturer. The inspector will continue to follow licensee corrective actions with respect to the pressure switches.
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The inspector observed the performance of procedure 0P-4204., . Test or Special Operation of the Safety Injection Pumps and Determination of ECCS Subsystem Leakage, as well as procedure OP-2677, Routine Venting of Safety Injection (SI) Header Pressure after SI pump operation Discussions with the auxiliary operator illustrated that he was know-ledgeable of the system and the requirements of the procedure. Com-munications' between the auxiliary operator performing the job and the operator in the control room were good. The inspector observed that the HPSI and LPSI surveillance tests were being performed as require After the HPSI pumps are run, the HPSI header pressure is vented requiring the operator to enter the TS action statement to restore containment integrity within one hou Procedure OP-2677 clearly recognizes that, prior to opening the designated containment isola-tion valve, the control room is to be notified that this action will cause them to enter the action statement and that an appropriate entry should be made in the control room log. The inspector verified that an appropriate entry in the control room log was made when the licensee entered and exited the TS Action Statement. No deficiencies were noted during the performance of this surveillanc During the performance of procedure OP-4244, Fire Suppression System Flush, the inspector noted that four licensee personnel had diffi-culties holding on to the discharge end of the fire hose; two of the personnel were observed to have fallen due to the excessive force of the water coming out of the hose. Precaution No. 1 of OP-4244 states that to prevent accidental injury from hose whip, the fire hose will be securely fastened at its discharge end or held by two men when flushing fire hydrants. The inspector observed four men holding the hose end (two were recommended by the procedure) and that when the fire pump was started, due to excessive pressure, the fire hose still whipped around and "got away" from the licensee personne One man sustained slight injury to his hand. The fire protection coordinator recommended earlier in the day and understood that the discharge end of the fire hose would be tied off to a pickup truck; this was not done and the result was that the four men holding the hose could not control its movement. Although the procedure was inadequate to pro-vide for proper personnel safety, this event would not have occurred had the FPC's guidance been implemented. Discussions with the FPC indicated that the next time the procedure was performed, the fire hose discharge end would be securely tied to an adequate suppor _ - _ _ _ _ _ - _ _ _ _ _ .
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On April 27, 1988, procedure OP-4244 'was again performed, with the fire hose being securely tied to the end of a forklif The health and safety supervisor was present during the reperformance of the fire suppression system flush. to observe and evaluate more suitable methods to do the system flush. The involvement of the health and safety supervisor to evaluate the current method of system flush and la more suitable alternative is a positive illustration of the licen-see's commitment to personnel safety. The procedure should be changed to reflect the guidance provided by the FPC to securely tie off the hose to an adequate suppor At 3:10 on April 30, 1988, an inadvertent actuation of the pressurizer power operated relief valve (PORV) occurred while the licensee was performing a surveillance of main coolant flow channels
>- for reactor startup. The plant was being maintained in hot standby and the licensee was performing procedure OP-4611, Nuclear Instru-mentation and Reactor Protection System Precritical Check prior to beginning reactor startu During the performance of main coolant flow channel checks, the I&C technician accidentally placed the test device in the incorrect Jack, inadvertently cycling open the PORV-(PR-S0V-90) for approximately two second When the error was noticed, the test device was removed and the valve was immediately
, close The PORV block valve, PR-MOV-512, upstream of the PORV was closed at the time and no radiological release occurred. The block valve is closed and tagged out of service for personnel safety whenever personnel are in the VC during operational modes 1-4, tem-perature ? 380 degrees F and general access is permitted. Personnel were inside containment at the time the PORV block valve was iso-i lated. No radiological or technical specification concerns were identified. The licensee reported this event via ENS at 6:30 p.m. on
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April 30, 1988; see Section 7, since the plant staff classified the event as an acttation of an Engineered Safety Feature (ESF). In addition, the licensee issued LER 50-29/88-07 on Mr.y 27, 1988 to describe the event and the corrective actions. The inspector ques-tioned the licensee as to how it arrived at a determination that the inadvertent actuation of the PORV constituted an ESF actuation. A document was developed by the technical services department that provided a rationale for the PORV and other systems to be considered ESF' The inspector will review the licensee's classification
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process during the subsequent review of the LER issued for this even . Onsite Review Committee Activities The inspectors attended regularly scheduled meetings of the Yankee NPS on-site review committee (PORC) on April 5,14 and May 10,17, and 18 to ascertain that the provisions of T.S. 6.5.1 were met.
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On May.4,,1988, the inspector became aware that PORC meeting minutes hac not been issued since October 14, 1987, a time span of almost seve, month TS 6.5.1.8 requires that the PORC maintain written minutes ef -
.each meeting and copies shall be provided to the M00 and chairman of t Nuclear Safety Audit and Review Committee (NSARC). T.S. 6.5.1.7.a states that the PORC shall recommend to the Plant Superintendent written-approval or disapproval of items considered under 6.5.1.6(a) through (d). These items. 6.5.1.6(a)-(d) include review of all procedures required by Spec-ification 6.8 and changes thereto, (i.e. , the generation, implementation and changes of required procedures), review of all proposed tests and experiments that affect nuclear safety and review of all proposed changes and modifications to plant systems or equipment that affect nuclear safet TS 6. 5.1. 7. b. states that PORC shall render determinations in writing with regard to whether or not each item in 6.5.1.6 (a)-(e) con-stitutes an unreviewed safety questio Finally, TS 6.5.2.2h requires that the Nuclear Safety Audit and Review Committee review reports and meeting minutes of the PORC. Licensee failure to maintain written minutes of PORC meetings and written determinations in accordance with TS 6.5. and TS 6.5.1.7.a and b is considered a violation (50-29/88-09-01).
As stated above, the licensee had failed to issue PORC meeting minutes since October 14, 1987. The inspector had several concerns regarding the licensee's failure to issue PORC meeting minutes in a timely manner. With no issued PORC meeting minutes within the time period, the licensee was not fulfilling TS requirements with respect to the responsibilities of POP.C . Also, the failure of the licensee to maintain written mf nutes of each PORC meeting and provide copies to the M00 and chairman of the NSARC in a timely manner brings into question how the licensee is fulfilling, officially, the requirements of TS 6.5.1.7a and 6.5.1.7b and also how the licensee officially documents PORC performance of TS 6.5.1.6 requirement The licensee is currently in non-conformance with the TS recuirements that determination, as to whether an item constitutes an unreviewed safety question or not, be in writing and written approval or disapproval of TS 6.5.1.6 (a)-(d) requirements. The inspector also questioned the licensee as to how the licensee would fulfill the intent of the requirements of TS 6.5.2.8.h, wnich requires the NSAR Committee to review reports and meeting minutes of the Plant Operation Review Committee. The next scheduled meet-ing was to be held on May 18, 1988. In order for NSARC to fulfill its intended function as an off-site review committee, they should review recent events at the plant rather than those which occurred quite sometime ag Consequently, the NRC was also concerned about the timeliness of NSARC review of TS requirement The NRC expressed concern to the chairman of the NSAR committee as to how timely review of PORC meeting minutes would be conducted in accordance with TS requirements if no PORC meeting minutes had been issued in the previous seven month ~
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n Discussions with the chairman of the NSARC indicated that he was aware of the requirement for NSARC to review PORC . meeting minutes and was also aware that he had not received copies of. the minutes since October 198 Additionally, he indicated that prior to the NRC involvement in this issue the plant was informally requested, below the management level to submit the . rest of the PORC minutes. The - chairman intended to discuss this timeliness issue at ,he NSARC meeting in May and also expressed that this was symptomatic of a larger problem involving timely corrective action Although annual audits are conducted of TS requirements, this lack of timeliness was not picked up on the annual TS audit just completed in March, 1988. This was because the prior audit had looked in this area and no problems were note Licensee response to NRC initiatives and concerns in this area have been aggressive and included the following immediate corrective actions: 1) all of the remaining 1987 and the majority of 1988 PORC meeting minutes hald as of May 4, 1988 were typed with special PORC meetings held to review and
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approve the meeting minutes and 2) the majority of these PORC meeting minutes would be issued and provided to the NSARC for their review in accordance with applicable TS requirements. Long term corrective actions to prevent recurrence of this problem includes: 1) revising procedure AP-0003, Plant Operation Review Committee to reflect a timeline for the timely issuance of PORC minutes (i.e., 15 to 20 working days to have the minutes revicwed by PORC and 30 working days to have the minutes signed),
2) ensuring personnel are aware that the PORC neeting minutes are a part of TS requirements, 3) tracking of PORC meeting minutes, which will be added and controlled as a part of the job responsibilities of a new full-time technical assistant in the technical services department, whose pri-mary function will be the administrative control of the- plant commitment tracking system, and 4) adding PORC minutes, by date, to the monthly status report provided by the technical assistant and issued to plant managemen The licensee was very responsive to NRC concerns in this area, took appro-priate and aggressive corrective actions and self-identified that this issue was symptomatic of larger licensee concerns in the area of correc-tive actions / commitment tracking system. The inspector had no further questions or concerns in this are . Plant Information Reports Plant information reports (PIRs) prepared by the licensee per AP-0004 were reviewed. The inspector determined whether the conditions were reportable as defined in the TS's and whether the licensee's system of problem identification and corrective action is being effectively utilize The following PIRs were reviewed:
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PIR N Occurence Date' Report Dat ' Subject 88-01 02/20/88 3/22/88- Nos. 1,2 & 4 NRV, Train A, 10% Test Circuit s
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failed to' operate-88-04 04/01/88 5/11/88 Fire Training 011. Stor-age Tank Spill
.PIR 88-01: This event was reviewed in Inspection Report 50-29/88-01 (Section 9). The solenoids (both test and dump) are scheduled to be-replaced with a different type during the 1988 refueling period by Rock-well.(NRV Vendor). The inspector identified no violations regarding _the licensee'.s. actions associated with these events and has no .further question .
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PIR 88-04: -This event is reviewed in Section 8a of this repor . Review of Licensee Events Reports !
Licensee _ Event Reports (LERs) submitted to NRC:RI were reviewed to verify that the details were clearly reported, including accuracy of the descrip-tion-of cause and adequacy of corrective action. The inspector c'etermined i whether further information was require 41 #orm the licensee, whether generic implications were indicated, ard whether the event _ warranted on-site followup. The following LERs were reviewed.
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Event Report LER N Date Date S_ubject 50-29/87-03 02/18/87 03/20/87 Reactor Scram -
High Main Coolant Pressure 50-29/88-01 01/26/88 04/15/88 Procedure inadequacy That Could Allow Insufficient Load Testing of Emergency Generators
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. 50-29/88-02 03/22/88 04/21/88 Reactor Scram -
Loss of Power to Nuclear Instrument Cabinet "A" 50-29/88-05 04/01/88 04/30/88 Degraded Fire Doors , LER 50-29/88-02: The details of this event are discussed in Section 79 of Inspection Report 50-29/88-05. With respect to the adequacy of the LER, the inspector noted the following:
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the licensee did not reference a possible similar past event
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that occurred with respect to a regulating transformer failure f
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the LER states that the capacitor was replaced in kind; there -
was no explanation of the corrective action taken to ensure the correct capacitor was installed and properly dedicated
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the LER did not include the indications lost due to loss of power to the Nuclear Instrumentation Cabinet
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it -is unclear what is meant by the cabinet having been satis-factorily teste ,
The inspector is concerned about the adequacy of this LER, the inspector discussed ~ the issue of LER adequacy with the technical services manager, who acknowledged the inspectors comments and con '
corns. The technical services manager indicated that following a review of the inspection report details, the licensee will evaluate the necessity of issuing a revised LE This LER remains open pending subsequent licensee review of the aforementioned concern i l LER 50-29/88-01: The details of this event are discussed in Section 7e of Inspection Report 50-29/88-05. The LER notes that the delay l in submission of.this report following the request for an evaluation on January 26, 1988, was due to an untimely review, which was not completed until March 14, 1988. It further states that to prevent a recurrence the importance of compliance with the requirement to sub-ait a reoort within 30 days of the discovery of an event has been emphasized to the appropriate plant stcf The emphasis appears to be on the need for a timely 30-day repcrt fcr compliance, rather than on the importance of a timely detailed evalu-atio This event could have been a safety significant event if the emergency diesel generators had been found to be incapable of per-forming their intended functio The inspector considered that emphasis should be on a tirrely evaluation of the event; secondary should be the requiremant for complianc The inspector had no ;
further comments and this LER is close LER 50-29/87-03: The details of this event are discussed in Section ;
8 of inspection report 50-29/87-0 The LER attributed the root cause of this event to be an open circuit in the stationary gripper coil for No. 2 control rod. However, the LER should address the root cause of the reactor scram since this is the event being reporte Additionally, the LER notes that the operators immed:ctely began to reduce ;M erator load to restore the main coolant system average tempe acre in accordance with a plant procedur A subsequent l
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engine * analysis by the cognizant engineering group of the Yankee Nuclear ..rvices Division, as documented in memorandum TAG 87-95 dated May 19, 1987, states that a best estimate analysis of the ~
February 18, 1987 rod drop with no load reduction indicates that a j trip on low pressure would not have occurre Therefore: they !
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,believe that' the best course of action followirg a rod drop is to allow.the transient to progress without immediate operator action to reduce loa Procedure OP-3118, Mispositioned or Dropped - Control rod (s) was later revised to reflect the engineering analysis; i.e., j if a reactor scram has not occurred allow turbine / reactor load an i main coolant system. average temperature to stabilize. The LER when 1 written noted that no other corrective actions were deemed necessar J
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The LER should be updated to reflect the subsequent analysis per-formed, the additional corrective actions taken and address the appropriate root cause determinatio This LER remains open, LER 50-29/88-06: The details of this event are discussed in Section 4.c(P) of this repor This LER is close . Emergency Exercise The inspectors participated in the review of the licensee's emergency exercise which took place on April 26, 1988. This review included three major areas:
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exercise preparation / review of scenario
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exercise observance
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review of licensee critique / presentation of NRC findings The details of the inspector's comments and findings were preser.ted to the NRC:RI team leader and is described in NRC Ir,spection Repert No. 50-29/
88-08 ,
15. Management Meetings ,
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Ouring the inspection period, the following management meetings were con-
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ducted or attended by the inspectors as noted below:
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The inspector attended an exit meeting held April 15, 1988, with an NRC:RI specialist inspector at the conclusion of Inspection 50-29/
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88-07, which consisted of a review of the licensee's eddy current inspections, water chemistry data and radiation exposure dat On April 20, 1988, the inspector attended at the NRC Region I Office )
a licensee presentation on the current status of the licensed operator training program.
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The inspector participated in management meetings associated with the i.- team inspection 50-29/88-07 of the licensee's Annual Emergency Plan Exercise conducted during the period April 26-27. 198 At periodic intervals during the course of the inspection period, ,
meetings were held with senior facility management to discuss the inspection scope and preliminary findings of the resident inspector,
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