IR 05000029/1988010
| ML20154J297 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 09/13/1988 |
| From: | Haverkamp D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20154J296 | List: |
| References | |
| 50-029-88-10, NUDOCS 8809220319 | |
| Download: ML20154J297 (25) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
! Report No: 50-29/88-10 ,
Docket No: 50-29
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Licensee No: DPR-3 l i
Licensee: Yankee Atomic Electric Company 1671 Worcester Road t Framingham, Massachusetts 01701 i Facility Name: Yankee Nuclear Power Station Inspection at: Rowe, Massachusetts [
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t Inspection Conducted: June 1, 1988 - August 1, 1988 '
Inspectors: Harold Eichenholz, Senior Resident Inspector [
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Cynthia A. Carpenter, Resident Inspector ,
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Approved By: Mb ** [M DonaldR.Haverkamp, Chief (/
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Date
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l Reactor Projects Section No. 3C l
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1 Inspection Summary: Inspection on June 1,1988 - August 1, _1988 -
4 (Report No. 50-29/-88-10) i
- Areas Inspected
- Routine onsite regular and backshift inspection by residen+ '
1 inspectors (295 hours0.00341 days <br />0.0819 hours <br />4.877645e-4 weeks <br />1.122475e-4 months <br />). Areas inspected included licensee action on previous I j inspecter findings, operational safety verification, radiological controls, !
! events requiring telephone notification to the NRC, plant events, maintenance !
- observations, surveillance observations, removal of loose resin and activated i-
] non-fuel components from the spent fuel pool, periodic and special reports, i
- on-site review committee activities, plant information raports, and Part 21 !
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j report In addition, an NRC Commissioner conducted a site visit during the
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! Results:_ No violations were identified by the inspector. Regarding an overall ;
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i facility a s ses smer.t for this inspection period, the NRC continues to note
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generally strong performance in plant operations, radiological controls, secur- [
! ity, and surveillance activitie This assessment is attributable to know-
! ledgeable Itcensee personnel and strong management oversight and involvement.
Areas that warrant increased licensee attention include: recognition of Tech-nical Specification inconsistencies and resolution of resulting issues in a !
pro-active manner (Section 4); providing a high level of quality maintenance in l a consistent manner (Section 8); and an atypical occurrence of developing ;
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insufficient written basis and conducting inadequate PORC review for a YNSD- !
developed safety evaluation (Section 12). l I
1 8809220319 880913 .
a PDR ADOCK 05000029 !
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TABLE OF CONTENTS
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Page Persons Contacted. . . . . . . . . . . . . . . . . . . . . . . . . I r Summary of Facility and NRC Activities . . . . . . . . . . . . . . l' Licensee Action on Previous Inspection Findings (IP 92702)*. . . . 2 Operational Safety Verification (IP 71707) . . . . . . . . . . . . 2 i Daily Inspection. . . . . . . . . . . . . . . . . . . . . . . 2 i System Alignment Inspection . . . . . . . . . . . ..... 4 ; Biweekly and Other Inspections. . . . . . . . . . . . . . . . 5 Backshift Inspection. . . .................. 6 !
, Radiological Control s (IP 71707) . . . . . . . . . . . . . . . . . 7 Events Requiring Telephone Notification to the NRC (IP 93702). . . 8 .
I Plant Events (IP 93702, 62703) . . . . . . . . . . . . ... . . . . 9 [
- Plant load reduction to perform condenser tube leak checks. . 9 ! Loss of Off-Site Communications . . . . . . . . . . . . . . . 9 ; Public Notification System Inoperability .......... 9 l Maintenance Observations (IP 62703). . . . . . . . . . . . . . . . 10 t Surveillance Observations (IP 61726, 62703). . . . . . . . . . . . 15 1 1 Removal of Loose Resin and Activated Non-fuel Components from the l Spent Fuel Pool (IP 71707) . . . . . . . . . . . . . . . . . . . 16 !
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1 Periodic and Special Reports (IP 90713) ............. 17 ?
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12. On-site Review Committee Activities (IP 40700) . . . . . . . . . . 19 l
1 Plant Information Reports (IP 90712) . . . . . . . . . . . . . . . 20 i 1 Pa rt 21 Repo rt ( I P 36100) . . . . . . . . . . . . . . . . . . . . . 21 f
1 Site Visit by NRC Commissioner and Staff . . . . . . . . . . . . . 22 i i
16. Management Meetings (IP 30703) . . . . . . . . . . . . . . . . . . 22 ,
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The NRC Inspection Manual inspection procedure (IP) that was used as i inspection guidance is listed for each applicable report sectio L i
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DETAILS Persons Contacted Yankee Nuclear Power Station N. St. Laurent, Plant Superintendent T. Henderson, Assistant Plant Superintendent R. Mellor, Technical Director Yankee Atomic Electric Company (YAEC)
B. Drawbridge, Vice President and Manager of Operations J. DeVincentis, Vice President, Projects J. Haseltine, Project Manager D. Maidrand, Assistant Project Manager-D. Edwards, Director of Industry Affairs G. Papanic, Senior Project Engineer, Licensing
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The inspector also interviewed other licensee employees during the inspec-tion, including members of the operations, radiation protection, chemis-try, instrument and control, maintenance, reactor engineering, security, training, technica* services and general office staff . Summary of Facility and NRC Activities At the start of the inspection period on June 1,1988 the plant was at 100% of rated power. Plant conditions remained stable until June 17, 1988 when a planned load reduction was initiated to approximately 47% of rated power to perform condenser tube leak checks. The plant was returned to full power on June 19, 1988 and except as noted below was maintained at full power until the end of the inspection perio During the latter part of this inspection period, the plant load was re-duced several times due to high circulating water discharge temperatur The licensee is required to maintain the discharge water temperature to Sherman Pond at less than 88 degrees F in accordance with its EPA dis-charge permit. On several occasions, the outlet temperature had risen to between 36 and 87 degrees F. In order to prevent exceeding to EPA dis-charge temperature, plant load was reduce During the period of June 20-24, 1988, an NRC Re9 1 on I (NRC:RI) specialist inspector completed an inspection of the licensee's solid radwaste pre-paration, packaging and shipping program (Inspection Report 50-29/88-12).
Also during the period of June 20-24, 1988, an NRC:RI operator licensing examiner completed Operator Licensing Examination 50-29/88-11. During the week of July 11-15, 1988, an NRC:RI security specialist completed a review of the security organization and security concerns (Inspection Report 50-29/88-13). During the week of July 18-22, an NRC:RI specialist inspec-tor completed an inspection of the licensee's in-service testing program (Inspection Report 50-29/88-14).
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- On June 29, 1988, licensee representatives held a meeting with NRC
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! hance the security progra On July 8, 1988, NRC Commissioner i
- Kenneth Carr visited the site and discussed items of mutual concern ,
between the NRC and the licensee. A meeting was held at the NRC:RI Office
, on July 14, 1988 between licensee and NRC representatives to discuss the !
i Systematic Assessment of Licensee Performance Board Assessment of licensee i performance for the period October 7,1986 through March 31, 198 !
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3. Licensee Action on Previous Inspection Findings
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(Closed) Violation 50-29/88-05-02: Review plant information report (PIR)
88-3 describing an event involving the removal of a radiation area / radio- i active material barrier and posting. The inspector's review of this PIR
- is contained in Section 12 of this report.
j This item is close . Operational Safety Verification i Daily Inspection
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I 1 During routine facility tours, the inspector checked the following
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items: shift manning, access control, adherence to procedures and limiting conditions for operations (LCO's), instrumentation, recorder
- traces, protective systems, control rod positions, containment tem-j perature and pressure, control room arnunciators, radiation monitors, j radiation monitoring, emergency power source operability, control
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room and shift supervisor log, tagout log and operating orders. No
! inadequacies were identified except as noted belo I j
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Ouring a review of the operations log book on July 7,1988, the
- inspector noted that on July 7,1988 at 5
- 15 a.m. the licensee
received a fire alarm for the vapor container (VC) cable tray
! (Zone 19). The licensee immediately performed OP-3017, Fire Emergency and entered the VC to investigate. No fire was dis-i covere The licensee utilizes linear fire detectors to provide j early warning for insitu combustibles in the V The linear j fire detectors are called "protectowire"; this consists of two j wires inside a protective jacke The licensee has two fire j zones inside the VC. Zone 18 is located in the grating at the
- personnel hatch entry way and is utilized to indicate to per-l sonnel who enter the VC the presence of a fire. Zone 19 con-l tains the linear fire detectors in the control rod drive cable j trays from the top of the reactor vessel to the various junction i boxes. These linear fire detectors annunciate and alarm in the l control room.
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Subsequent licensee investigation revealed that the protectowire in the cable tray had sustained a shor Technical Specifica-tion (TS) 3.3.3.4 requires a minimum of one fire detection instrument in the VC to be operable. With the number of oper-able fire detection instruments less than required, the TS requires a fire watch patrol to be established within one hour to inspect the zone at least once per hou The fire protection coordinator (FPC) discussed the meaning of the TS requirements with the shift supervisor. According to the FPC, the only linear fire detection system in the VC is for the insitu combustibles in the cable tray; the other fire detection string near the VC personnel hatch is for information only and no combustibles exist in that are Consequently, the TS requirements apply to that fire detection system in the cable tray only, which at that time was inoperabl Therefore, tech-nically the licensee did not appear to have the required minimum one channel of fire protection in the VC operable per TS 3.3. requirement However, following shift supervisor and plant management review, the position taken was that because two zones of fire detection existed in the VC, the licensee still met the minimum number of required instrument To restore appropriate fire detection capability in the cable tray area, the licensee connected an ohm meter to each end of the protector wire with the ohm meter having a read out in the control room. Instructions were provided to the control room operators to record the lead resistance readings from both sides of the loop to the shorted section every hour; if a drop in the reading of > 0.002 kohm from the base line values is read on either meter, then a VC entry was to be made to investigate the possibility of a cable tray fire or overheating cable No alarm function has been installed, although the licensee is investigating this possibility. This system provides for rapid detection of a potentially degraded condition since there are insitu combustibles located in the rod drive cable tra The inspector considered that the licensee failed to recognize the significance of the loss of the cable tray fire protection system with respect to its applicability to TS requirement The inoperability of this fire detection system actually placed the number of fire detection instruments in the VC at less than the minimum required per TS 3.3.3, Although the licensee compensated for this loss, the license * failed to recognize the applicability and failed to enter the TS action statemen __ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ . _ _ _ _ _ _ _ _ - __-_ _________________ __ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _
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The inspector discussed this issue with licensee representa-tives, who acknowledged the inspector's comments and concerns.
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The licensee then agreed that the inoperability of the cable tray linear fire detector constituted less than the minimum required channels per TS, which then resulted in the licensee considering that.they were in the action statemen Although the licensee considered that TS 3.3.3.4 was now appli-
, cable, they believed that the action statement was not appropri-ate for the remote location of the VC. Their concerns centered
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result in degradation of personnel hatch integrity. The inspec--
tor agreed with the licensee's concerns as for the inappropriate
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nature of the applicable action statemen ,
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the NRR Project Manager the licensee's concerns, their use of ,
the non-alarming ohm meter, and logging the resistance readings !'
hourly as a substitute for an hourly fire watch patrol. The NRC i considered the use of the remote readout to be an appropriate
equivalent compensato.y measure to the loss of normal fire detection instrumentatio *
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The licensee has committed as a longer term corrective action to propose an amendment to the TS for more appropriate action
- statements for inoperable fire prctection instrumentation in i restrictive areas.
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. The failure of the licensee to recognize the applicability of the TS action statement was to some degree the result of an
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i inappropriate and apparently ambiguous action statement. In the
most recent SALP report, 50-29/86-99, NRC concerns in this !
j regard were documented. This example further illustrates the '
i need for a coordinated effort to resolve TS inconsistencies and
- ambiguities in a pro-active fashio Because the licensee took l prompt compensatory measures to assess the degraded but usable ,
instrumentation system on an ongoing basis, and the fact that
the inability to complete the required action was of low safety
! significance and atypical licensee performance, a violation will !
4 not be issued for failure to adhere to TS requirement I System Alignment Inspection ,
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j Operating confirmation was made of selected piping system trains.
j Accessible 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 assesse The following systems were checked ;
! during plant tours and control room panel status observations: ;
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Low pressure and high pressure in,iection systems
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emergency diesel generator unit
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spent fuel cooling system
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charging system c. 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 Specifications 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 the action was properly conducted. No inadequacies were identifie (2) Fire Prntection 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 inspecto Performance in this area continues to be viewed as a licensee strergt No violations or deviations were identified in the review of this program are (3) Observations of Physical Security
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On June 16, 1988 at 2:48 a.m., the Central Alarm Station (CAS) operator was found to be asleep at his post by the security shif t superviso This item is discussed further in Section 6 of this report. The inspector had no further
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questions of the licensee on this item.
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Improvements in station management oversight of security activities continues to be observed by the inspecto The following actions have been taken by the licensee during this inspection period to enhance the existing security progra l
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. 6 (a) Procedure AP-0478, Rev. O, Security Preventative Main-tenance and Surveillance Matrix was issued. The issu-ance of this procedure completes a commitment made by the licensee at the June 23, 1987 Enforcement Confer-ence(50-29/87-08).
, (b) On July 24, 1988 the licensee hired five security l
shif t supervisors (SSS). These SSS were formerly con-tractor employees. Contracted lead security officers have been assigne In addition, a physical security specialist has been hire (c) A total of nine plant security doors, including three control room doors, were replace This item ad-
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dressed prior NRC concerns relating to door closure inadequacies and high false alarm rates.
l (d) Eight new security officers were hired to address staffing adequacy issue (e) The internal audit program has been enhanced.
l (f) New security equipment and weapons were obtained or
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approval for purchasing received from licensee manage-4 men Personnel / retraining in the use of the PR-24 baton has been accomplishe (g) RER findings continue to receive a high level of
- Itcensee attention. An alternate storage location
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for contingency weapons was implemented. Assassment
- perimeter modifications and system improvements are
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continuing. This last item was discussed in der. ail at a management meeting held on June 29, 1988 at the NRC:RI Office.
] d.- Backshift inspection i
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The inspector conducted backshif t, weekend or holiday inspections on
! June 1, 8, 16, 19, 20, 21, 28, and July 3, and 6. Operators and j shift supervisors were attentive and responded appropriately to j annunciators and plant condition No violations were identified
during backshift inspections.
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. 7 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 tne t inspecto ;
During the course of this inspection period, the licensee has been active-ly performing work in the spent fuel pit area to remove loose resin and activated nonfuel components from the spent fuel pool (See Section 10).
During the resin clean-up job in the spent fuel pit, the inspector ob-served that the licensee had implemented AP-0811, Hot Particle Contamina-
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tion Control. The licensee evaluated the job prior to performance and i established hot particle zones (Zones 1, 2, or 3) based on the probability ,
of the area becoming contaminated. Specific guidelines were then estab- i lished for each hot particle zone, including requirements for barriers, !
protective clothing, radiation protection coverage frisking and posting Additionally, ALARA reviews and briefings were perforaed and discussed prior to the job. Additional radiation protection personnel contractors ,
were brought in to assist in following the jo Radiation protection personnel were available in the area and were very knowledgeable of the j job including radiological and hot particle concerns. Also, RP engineers followed the job throughout, and an additional decontamination person was (
brought in to support the program by performing materials control, setting l up the work program and decontaminating equipment as it was removed form '
the spent fuel poo The radiological aspects of the spent fuel pool cleanup appeared to have I been well planned, adequately supervised and staffed so that an effective radiation protection program was implemented to ensure good control over t the work. The implementation of the licensee's new hot particle contamin-ation control program was effective and noteworth On June 1,1988, the inspector reviewed repair activities associated with isolation valves WD-V-722 and WD-V-723 valves on the activity decay and dilution tank in the radwaste system. The radiation protection department
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(RPD) issued an ALARA review (No.88-019) in accordance with plant proced-ures. The inspector verified that a pre-job checklist was developed, that ALARA controls were implemented, and that a pre-job briefing was conduc-ted. A post job debriefing was held on June 3, 1988 which included repre-sentatives from the various plant departments involved in the repair activitie A high level of RPD supervision and job oversight was observed by the inspecto No unacceptable conditions were identifie . _ _ __ __ _
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The licensee issued a Procedure Change Notice to OP-8020, Rev. 3, Imple-mentation and Documentation of ALARA Job Reviews on July 14, 1988. This revision to the procedure incorporates an ALARA review audit log to record l'
planning and review information while work activity is ongoin This-enhancement to the ALARA program allows pertinent information to be trans-ferred to the post job review at the completion of the work activities.
i Review of 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 2:48 a.m., June 16, 1988, the central alarm station (CAS) operator ;
, was found to be inattentive at his post by the Security Shift Super-
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visor. When last checked at 2:?1 a.m., the CAS operator was observed i
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to be attentiv The inattentive CAS operator was relieved of his !
i duties, but because of manning requirements at the time, was reas- '
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signed to duties in the gatehouse. Upon the arrival of a replacement security guard at the plant at 3:25 a.m., the previously inattentive
- guard was released from duty and his site access cancelle An assessment performed by the licensee determined that no actual de- l
, gradation of security functions occurred during the 27 minutes that l l the CAS operator could have been inattentive to his dutie The NRC i was notified of this event at 3:35 a.m. The inattentive security I l guard's employment with the security contractor was terminated. The I
- licensee's actions as a result of this event were timely and ,
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At 9:46 p.m. on June 16, 1988 the NRC was notified in accordance with 4 10 CFR 73.71(b)(1) of a potential bomb threat. The licensee became aware of the threat from the Massachusetts State Police (MSP). The !
MSP received an allegation that the threat involved a plant and con- I 1 tractor employee. A thorough licensee investigation was conducted j concerning this event. Compensating protective security actions were
- placed in ef fect at the plant until the licensee's investigation ;
could determine that the incident was a hoax and the event terminated '
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at 9:57 a.m. on June 17, 1988. Security management oversight was i
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evident throughout the investigation, with the inspector noting that
the entire event and development of licensee actions were treated L from a conservative and concerned viewpoint by the license ;
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At 5:55 p.m. on June 23, 1988 the NRC was notified in accordance with ,
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10 CFR 50.72(b)(I)(v) that the radio (KN8Y-400) in the plant radio j page system was inoperable resulting in a major loss of off-site communications capabilit This event is discussed further in i Section 7.b of this report, i l
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At 4:33 p.m. on July 26, 1988 the NRC was notified in accordance with 10 CFR 50.72 (b)(1)(v) of the loss of the Ames Hill (Marlboro, Vermont) section of the Public Notification Syste The Ames Hill portion of the public notification system was determined to have been out-of-service at 3:00 p.m. on July 26, 1988 and was returned to service at 6:10 p.m. the same da The Massachusetts and Vermont State police were appropriately notified. This event is discussed further in Section 7.c of this repor . Plant Events Plant Load Reduction to Perform Condenser Tube Leak Checks On June 17, 1988 the licensee commenced a planned load reduction to approximately 47% of rated power in order to perform condenser tube leak checks and other maintenance. Nine leaks were identified and plugged. The licensee also conducted non-return valve 10% closure test and turbine throttle valve exercise in accordance with procedure OP-4225. The plant was returned to full power operation at 2:00 on June 19, 198 No unacceptable conditions were identifi2d by the inspector, Loss of Off-Site Communications During the licensee's daily radio check, at 5:36 on June 23, 1988, the plant radio page system (KNBY-400) was found to be inoper-able. The plant radio page system activates the plant pagers used for emergency plan augmentation. The transceiver is located on M Olga in Wilmington, Vermont and the licensee suspected that the severe thunderstorms on June 22-23, 1988 may have caused the trans-ceiver to be inoperabl The transceiver was repaired on June 24 1938. The licensee considered this event to be a major loss of off-site communications capability and made a one-hour non-emergency ENS cal Licensee back-up for this system is the commercial telephone networ No violations were identified as a result of the inspector's review cf the event details, Pyblic Notification System Inoperabiliy On July 26, 1988 at 4:00 p.m. , the licensee was notified by their contractor of a loss of the Ames Hill (Marlboro, Vermont) section of the public notification system (PNS). The system had been out of service at 3:00 that day and was returned to service at 6:10 p.m. later that da . .
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. 10 Upon questioning the control room operator about the condition during the ENS call, the NRC duty officer asked if the event involved inoperability of the strens. On two occasions during the ENS call, the control room operator indicated that the sirens were affected by the condition. Although sirens constitute a portion of the PNS, the Ames Hill section consists of transmitters and receivers that are utilized to actuate the tone alert radios that are in use in portions of the Emergency Planning Zon Following discussions with addi-tional control room operators, the inspector concluded that a weak-ness existed in the knowledge level exhibited by the operators per-taining to the PNS and the siren It was clear that the lack of understanding about which components comprise the parts of the PNS resulted in misinformation being transmitted to and recorded by the NRC. The issue of not understanding the components that comprise the PNS was previously addressed in Inspection Report 50-29/85-0 The inspector discussed the issue with cognizant licensee personnel, who acknowledged the NRC's comments and concer As a result of a prior commitment to NRC:RI emergency preparedness specialists, the licensee intends to revise their emergency planning couunications procedures to clarify what constitutes a major loss of communications capability. At that time, the Operations Memo, OP-Memo 2U-5, Public Notification System, will be revised to provide clarification to the operators as to the various components of the PNS. The inspector informed the licensee that their training programs for developing and maintaining operator licensees warrant upgrading to provide for effective training in the operation and use of all aspects of the PN The inspector had no further questions in this are . Maintenance Observations The inspector observed and reviewed maintenance and problem investigation activities to verify compliance with regulations, administrative and main-tenance procedures, codes and standards, proper QA/QC involvement, safety tag use, equipment alignment, jumper use, personnel qualification, radio-logical controls for worker protection, fire protection, ratest require-ments and reportability per Technical Specification The following activities were reviewed:
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MR 88-867; Repair CP-30 fire system control panel module
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MR 88-893; Discrepancy on No. 3 hot leg wide & narrow range indica-tors on main control board
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MR 88-921; Add to transfer pump valve WD-V-722 hand wheel discon-nected from valve
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MR 88-928; Add drain line valve - change diaphragm anc bolting
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MR 88-971; Comsip hydrogen analyzer flow meter indicates flow without system flow
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MR 88-973; CS-MOV-529 BAMT to charging pump suction valve motor trips on overload when fully open
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MR 88-1084; MS pressure switch MS-PS-21 replaced
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MR 88-1085; MS pressure switch MS-PS-22 replaced
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MR 88-1086; MS pressure switch MS-PS-23 replaced
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MR 88-1093; CS-MOV-529 BAMT to charging pump suction valve motor trips wr.:n attempting to cycle open
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MR 88-1259; Primary rod position indication (rod No. 24) lights not -
fully functioning
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OP 5000.227; Rev. O, Specific gravity correction of batteries N and 2 Based upon a review of licensee activities in this area, the inspector noted the following:
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Regarding MR's 88-1084, 88-1085 and 88-1086, the licensee has re-placed the remaining three main steam line pressure switches that they had committed to replac At this time, all main steam line pressure switches have been replaced since recent problems were encountered on December 3, 1987. The licensee is continuing their investigation of the problems with excessive drift of the pressure switche The licensee is continuing their surveillance frequency at twice a month and continues to trend the as-found condition of the pressure switche During routine surveillance testing of the station batteries No and 2, the licensee noted that the specific gravities were dropping and approaching the TS 4.8.2.3.2. surveillance limits of 1.200. These batteries are 59 cell type LC-19 C&O batteries. On May 24, 1988 this condition was discussed at PORC meeting 88-54. It was noted that recently procedures have been received from the C&O Company that can be used to correct the low specific gravity condition in the batter-ie The licensee informed the inspector that based upon discussions with the manufacturer and installation personnel, it appears that following the batteries being installeu in 1982 (batteries were shipped with electrolyte at the low level mark) they were probably topped of f with denineralized water in lieu of electrolyte, and that ;
is the reason for the existing electrolyte low specific gravity condition.
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The licensee developed a one-time maintenance procr jure OP-5000.227 to perform the specific Jravity correction on stati a batteries No and 1. The procedure involves removing electrolyte from one cell
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at a time and replacing it with battery ecid. When the electrolyte is removed from the cell, the electrolyte le,;l will drop below the minimum mark, but would remain above the top of the plates. Since TS 4.8.2.3.2. surveillances requires that the electrolyte level be maintained between the minimum and maximum level marks, the mainten-ance manager discussed this cordition with the inspecto The licensee believed that the envisioned process, as supported by engi-neering evaluations and the manufacturer, would not result in the batteries being in an inoperable condition while each cell is being corrected. A determination, that a single cell which has an elec-tvolyte level below the low level mark creates an inoperable battery, would necessitate entering action statement b. of TS 3.8. The inspector discussed this Tsue with an NRC:RI specialist inspec-tor, whu concluded that tne battery was not inoperable given the specified conditions and t'nat entry into the TS action statement (i.e., restore the battery to operable status within two hours or initiate a plant shutdown) was unwarranted. Additionally, it was noted that the practice of specific gravity correction in the field is fairly common and has not resulted in any regulatory concerns when performed correctly. The inspector informed a licensee representa-tive that their planned maintenance process would not be considered to create a battery inoperability solely due to removing electrolyte level belnw the low level mark as part of the correction prcces The average specific gravity prior to the correction process, as l determined by th0 OP-4501 quarterly battery surveill .nce test was 1.202 and 1.203, for batteries 2 and 1 respectivel Following the correction for battery No. 2, as performed on June 27-30,1988, the average specific gravity was 1.21 Similarly, battery No. I was
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corrected on July 5-6, 1988 with the result of an average specific ,
j gravity of 1.214.
The inspector noted that the entire evolution represented a well planned maintenance process. Active involverant and oversight by the maintenance support department was evident, with corporate engineer-ing support provided when conditions required their review and [
involvemenc. Appropriate procedures were developed and in.plemented to control the activitie Documentation of the activities wa; readily available and site QC personnel were invoived in the imple-nentation of the maintenance activity. No unacceptable conditions were identified by the inspector as a result of reviewing the licen-see's activities r9 this ite '
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t On June 1, 1988, the inspector reviewed maintenance activities asso-
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ciated with MR 88-921. This MR resulted from an operator on May 31, l 1988 determining that the hand wheel continued to turn (valve hand l wheel utilizes a reach rod through a concrete wall) on valve
- WD-V-72 Subsequent investigation by maintenance personnel deter-
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mined that the hand wheel had disconnected due to a cracked valve j bonnet. Because this was an outlet ' olation valve on Tank Tk-32, -
activity dilution and decay tank, and a possible failure of the valve
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bonnet could release the contents of the tank (52,000 gallons of I radioactive liquids) into the moat, further maintenance activities were suspended. Although the moat was designed to hold the contents j of the tank, the primary concern was the release of the cover gas and
, the release of gas dissolved in the water. According to a detailed i analysis of the worst case scenario for tank contents release con-i tained in PORC Meeting Minutes No. 88-56, dated May 31, 1988, the estimated doses were 2,79 mrem /yr, skin; 2.7 mrem /yr total body.
l The PORC comaittee reviewed the situation, evaluated various types of I corrective actions as methods to mitigate the consequences a failure i of valve WD-V-722, and recommended a course of action. The inspector i determined that the PORC's ratienale was appropriate, and from a
! tafety perspective was conservativ Additionally, PORC determined j that appropriate guidance was needed for plant operators to make an
- Emergency Classification decision in the event that a ground level l release occurs. Subsequently, the operations department issued 1 Special Order No. 88-54 which provided the necessary guidance, iden-
! tified corrective actions that have been initiated and items for the
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e, wators to be cognizant o On June 1,1988, the diaphragm valve bonnet and bolting were re-placed. Freeze sealing was utilized by the maintenance personnel to isolate the valve from the tank bouadar The inspector verified
- t.hst appropriate maintenance procedures were developed and utilized i to control the activities. The entire evolution was well coordinated 1 by tne maintenance staff, and involved personnel from operations, J maintenanc.e. QA, stores and radiation protection department In-spector observations pertaining to the licensee's radiological con-
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trols established for control of activities associated with this l maintenance evolution are disci. sed in Section 5 of this repor No 4 release of radioactive materiai, to the envi-onment occurred.
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l The inspector identified no concerns as a result of reviewing licen-I see activities on this item. On July 13, 1988, the licensee issued j Plant Information Report 88-7 to describe the event and corrective
, actions, including the need for installing additional piping j supports.
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Regarding MRs88-973 and 88-1093, the licensee experienced repeated i problems with the boric a;id mix tank (OAMT) isolation valve CS-MOV-52 MR 88-973 was issued on June S. 1988, when the plant operators identified that the valve motor power was tripping out on overload. This occurred during the performance of surveillance pro- ;
cedure OP-4214, Rev.13, Chemical Shutdown System Operability Chec As part of the corrective maintenance, the limit switches were thought to be out of adjustment and were subsequently adjusted. Pro-cedures OP-4517 and OP-4214 were used to cycle the valve and check t its open/ closing time and verify the free flow past the seat, respec- ;
tively. The valve is part of the Inservice Testing (IST) Program, so ,
that the decreased stroke time of 16.59 seconds was compared to the i acceptable range of 16 20 seconds. The decrease in stroke time was !
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acceptable to ASME Section XI-IWV and plant procedure OP-4518, Travel Time of Motor Operated Valves - Master Lis The IST coordinated ;
entered the 16 second stroke time into the IST data base when he }
verified that the limit switches for the valve were adjuste The maintenance activities for the CS-MOV-529 valve were reviewed by !
a QC inspector on June 8,1988. This review used QC implementing '
instructions and a checklist. According to QC Inspec*, ion Report N , the QC inspector identified a concern about the possible l butidup of boron in the back seat of the valve, which could be the l cause t'or it not fully openin Additionally, no verification of i flow quantity had been performed. As a result of the QC concern on ,
the valve's ability to pass the required flow rate, a Deficiency / !
Observation report form, OQA-X-5.4, was issued on June 10, 1988 by l the QC inspecto The cognizant maintenance support department :
angineer, who is also the IST program coord'1ator became involved in [
the re-review of the operation of the CS-M05-529 valv On June 21, '
1988, the valve was again cycled and actual valve stem travel was i recorded. The dimensional check was compared to the valve manufac- t ture 's drawing, and it was determined that the valve was not fully !
back seate The valve was determined to be 70.9% open. The main-tenance department supervisor was informed on June 23,1988, which !
resulted in a determination to perform a flow calculation on the l present valve position and ensure compliance with the Technical -
Specifications. Subsequently, the flow calculations were performed by [
the cognizant maintenance support department engineer and Yankee '
Nuclear Service Division (YNSD). The formal YNSD calculations were !
transmitted to the plant on July 1, 198 The calculations were I attached to a memorandum dated July 6,1988 by the IST coordinator f that described the problems experienced with the valv ,
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On June 29, 1988, while again performing the valve stroking required
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by surveillance procedure OP-4214, the CS-MOV-529 valve motor power tripped out on overload prior to completing its full trave MR
88-1093 was submitted, the valve was declared inoperable, and the j plant operators entered the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of allowed out-of-service time provided by the action statement of TS 3.1. This time the sus-i pected boron buildup was considered the primary reason for the in-
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operability of the valve, and resulted in a detailed investigation by the maintenanca departmen Heating the bonnet of the valve,
while cycling it, was successful in removing the boron buildup. Ap-propriate . controls were exercised by the licensee to monitor the i valve's bonnet temperature. A special operations procedure OP-2000-193 Re O, Flush of the Boric Acid Mix Tank Suction Line was i developed to reduce the high boron concentration in the valve bonnet.
J The PORC reviewed the event on the day of occurrence, determined that
! appropriate corrective actions were being implemented, and requested that the maintenance department investigate insulating and heat tracing the bonnet of the valve, ,
Ouring the maintenance activities on June 29, 1988, the inspector i noted a high level of concern by the plant staff about the even , This concern resulted in good coordination and com1unication between l the varicus plant departments, proper sensitivity by plant manage-
, ment, and timely corrective action This was obviously contrasted 1 by the earlier event on June 7,1986, that received an appropriate i level of concern and attention by the site QC department, but did not i result in either a satisfactory questioning attitude or timely reso-l lution of identified issues by the maintenance department, i
l In the recent SALP Report No. 50-29/86-99, the NRC documented its
- concern for the inability of the licensee's maintenance organization
! to maintain a high level of quality maintenanc The inability of I
the maintenance department to ascertain the root cause of the failure of the CS-MOV-529 valve on June 7, 1988 and the lack of resolution of concerns identified by the site QC organization provide further evi-dence of the need for aggressive licensee management attention to the
! NRC concerns. The inspector had no further questions of the licensee on this matter at this tin:e.
l l 9. Surveillance Observations t
! The inspector observed tests or parts of tests to assess performance in j accordance with approved procedures and LCOs, t(st results (if completed,
! removal and restoratirl of equipment, and deficiency review and resolu-
! tio The following tests were reviewed:
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OP-4238, Rev. 8, Test of the Control Room Ventilation Emergency Shut-down System & CREACS Fans conducted on May 25, 1988
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OP-4716, Rev. 8, Vapor Container Personnel Hatch and CA-V-755 Leak Test conducted on June 1,1988
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OP-4702, Rev. 13, Vapor Containment Type B&C Penetration Tests con-ducted on June 2, 1988
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OP-4656, Rev. 8, Functional Test of the NRV Main Steam Line Pressure
- Channels / Switches conducted on June 2,1988
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OP-4225, Rev. 18, Turbine Throttle and Control Valve Surveillance >
Test conducted on June 17, 1988
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OP-4517, Rev. 12. Surveillance of Motor Operated Valves conducted on June 29, 1988
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OP-4210, Re , Fire System Operability Test conducted on
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July 1, 1988
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OP-8041, Re , Special Samo11ng of Primary Vent Stack Effluent
. conducted on July 31, 1983 No unacceptable r.onditions were ider+1fied as a result of reviewing the !
licensee's activites in this program aro '
10. Removal of Loose Resin and Activated Nonfuel Components From The Spent NeDool An inventory of activated nonfuel components, predominantly control rods ;
and follower sections had accumulated in the spent fuel pool (SFP) over i
the years of plant operation . The need for additional fuel storage space
! dictated that the SFP be cleaned out prior to the 1988 refueling outag Additionally, resin f rom the SFP cooling ion exchanger was found at the bottom of the SFP. During this inspection period, the licensee commenced the cleanup of the resin from the floor of the SFP, dose profiling of the
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irradiated hardware, and the rushing and shipping of the irradiated hard-war The spent resin was vacuumed from the SFP botton and into a Chem-Nuclear System, Inc. (CNSI) trarsport cask for shipment of f-sit The transport cask is the property of CNSI, and, therefore, Chem Nuclear presonnel ware j onsite to assist the license The inspector observed the licensee's ,
inspection of the transport cask to be used for resin shipping to assure '
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l that it had not sustained any damage that would af fect cask operation or integrity and to ensure that the cask complied with USNRC Certificatio Licensee and Chem-Nuclear procedures were used to handle and inspect the transport cask and pol >ethelene higr, integrity containe Procedures
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't appeared to adequately detail precautions and work steps; procedures were
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observed in use on the job site, and prerequisites had been met. Person-nel were very knowledgeable about the work, continuous supervision was observed by cognizant personnel and continuous on-site quality assurance oversight was present. Radiological control precautions appeared to have been conservative with health physics Oversight also evident. Hoses to be used to pump out the resin were recently hydrotd and equipment appeared to be in good condition. The cognizant engineer was observed to verify that a current leak test was in effect for the cask and the date a new gasket had been installe The inspector also observed portions of the dose profiling of the irradi-ated hardware in the spent fuel pit; the hardware included fuel cages, shim rods and absorber rods. Dose profiling of irradiated hardware is required for the proper characterization of the hardware for disposal, the actual movements of the components were done by licensee personnel while Chem-Nuclear performed the actual dose profiling, using their own proced-ures. Approved procedures were available and being followed, good pre-cautions were taken especially in the area of radiological control Quality assurance personnel were observed to be present throughout this work evolution also and good management oversight was available. Proced-ural precautions were observed to be adhered to and prerequisites had been met. Personnel were briefed prior to the job on ALARA and hot particle concerns, ano an ALARA review was p epared for the job, with emphasis on hot particle concern Radiological crecautions taken for this job are discussed in more detail in 3ection 5 of this repor The licensee's approach and execution of these jobs appeared to have been well planned with appropriate precautions specifically taken in the area of radiological control Personnel were knowledgeable of the job tasks and the oversight provided by cognizant personnel and quality assurance was noteworthy, Chem-Nuclear prepared procedures were approved by PORC and, together with licensee procedures, led to the efficient performance of the jo The inspector identified no concerns or inadequacie , Periodic and Special Reports Periodic and special reports submitted to the NRC pursuant to Technical Specification 6.9 were reviewe The revi'.w ascertained; inclusion of information required by the NRC including t.sst results and/or supporting information; consistency with design predictions and performance specifi-cations; adequacy of planned corrective action for resolution of problems; determination whether any information 55nuld be classified as an abnormal occurrence; and validity of reported information. The following periodic reports were reviewed:
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Special report submitted per TS 4.4.10.2 and 6.9.6 containing the inservice inspection (ISI) results for steam generator No. 2 and the re-analysis of the 1984 eddy current data for steam generators No and Monthly statistical reports submitted per TS 6.9.3 for the months of February 1988 through May 198 Special report submitted per TS 6.9 containing the licensee's Cycle XIX, Startup Phy.ics Test Report Attached to the licensee's special report submitting the results for the ISI of Steam r;enerator No. 2 was an additional special report containing the re-evaluation of previous inservice inspection data for steam gener-ators No. 3 and No. 4 performed in May and hne of 1987. This information was intended to supplement the 1984 rep et, "Steam Generator Inservice Inspection Results." Results of previous inspections done in 1985 for Steam Generator No.1 and in 1984 for steam generators No. 3 and No. 4 were also re-evaluated. This report is in response to the licensee com-mitment to submit additional information via a supplementary special report to update the 1984 results of three steam generators, per licensee
, letter to the NRC FYR 87-65 dated June 15, 1987. The inspector had no
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concerns in this area.
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The inspector reviewed the licensee's Cycle XIX, Startup Physics Test Report documented in accordance with the requirements of Technical Spec-1 ification 6.9 "Reporting Requirements". The results were summsrhed in a
! technical document entitled "Core XIX Startup Program for the Yankee Nuclear Power Station" and attached to the licensee's letter dated
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September 30, 1987. The test report contains an introduction, summary of j results, startup program mechanical, startup program nuclear, reload design reanalysis, and a reference sectio '
Ths purpose of the startup physics test program is to verify that the reea sured parameters comply with Technical Specification limits, and to validate core design calculations. The content of the test program was compared and found to be consistent with the requirements described in the FSAR and Technical Specifications. The results of the startup test pro-gram were reviewed to assure compliance with Technical Specifications and the test program acceptance criteria. All measured parameters met tha Technical Specification limit All parameters except the control rod grot.p A worth met the test acceptance criteri _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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. 19 The control rod group A worth was mea sured to be 1536 PCM. The predicted group A worth was 1370 PCM. The difference was calculated to be + 12.1%.
The acceptance criteria for individual groups i s 7.5%. The inspactor i found this deviation acceptable for the following reason:
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The total rod worth met the acceptance criteri The measured percent difference was within recommended test criteria in industry standard The small predicted worth of gtoup A (1370 pcm as opposed 2300 pcm for group B) caused deviations in the measured value to have a higher percent differenc In summary, the startup physics test program and results were reviewed and found to be acceptable. The repcrt submittal was in accordance with the i requirements described in Section 6.9 of the Tcchnical Specification No violations were identified in the review of this area, j j 12. Onsite Review Committee Activities The inspectors attended regularly scheduled meetings of the Yankee NPS on-site review committee (PORC) on June 21 and July 12 to ascertain that the provisions of T.S. 6.5.1 were me PORC meeting minutes were observed to be written and sent to PORC for l their review in acco otnce with TS requirements in a timely manner, demon-
, strating the appropriateness and aggressiveness of the licensee's correc-tive actions in response to prior NRC concerns identified in this area
! (Inspection Report 50-29/83-09).
) Also, previous Systematic Assessment of Licensee Performance (SALP)
reports have enumerated inspector concerns about PORC meeting minutes
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! lacking adequate level of detail to reflect the level of detail of the
, discussions. However, recent PORC meeting minutes reflect a greater i amount of detail and are therefore more representative of the meetings and i the subjects discusse The inspector considers the more recent PORC j meeting minutes to be an improvement over previou, minutes and a positive j trend. Centinued improvement in this area will enhance the oversight of PORC and provide a greater documented insight of the safety rev%s con-ducted and discussed by the PORC, l
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On July 5,1988 at PORC meeting No. 88-67, the committee reviewed a series
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of procedure char.ges that resulted from a change in the manner in which the emergency turning c ar oil pump (ETGOP) is operated. A YNSD memo, YRP
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746/88, providea the bas!s to remove the control switch for the ETG0P from the "trip pullout" position during plant shutdown and startup evolution This condition of operation resulted from a limitation imposed by the
- original station battery No. 3 size and short-term load Due to the replacement of the original battery No. 3 with a larger battery, and in recognition that unavailability of the automatic operation of the ETGOP increased the probability of turbine bearing damage during certain loss of power and equipment conditions and places an additional burden on oper-ators, the YNSD recommended the operational change to the plant. The YNSD recommendation became the basis for the prc:edure changes. Because the existing 75 ampere battery charger for battery No. 3 cannot support the
, normal load on the de bus No. 3 and the ETG0P running current (both total
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91 amperes), and with the projected use of the battery for the two hours
, in this mode, the terminal voltage will drop to .116.8 volt The YNSD
! memo and the PORC review did not address the issue of the battery voltage
! falling below the 125 V de surveillance requirement of TS 4.8.2.3.2. The inspector determined that the written safety analysis by YNSD, and the subsequent review by the PORC, were not of the high quality usualli deter-mined to be performed. Although the inspector ut terstood the basis for the licensee's actions, anc igreed that these acti ons were "ot inappro-palat It was necessary to point out to the lit.ansee that the written basis for safety evaluations must be complete and fully document all TS considerations. Additionally, TS considerations, especially when planned operations can result in limiting conditions for operations becoming applicable, must be clearly documented to insure tha. the operational pro-J cedures make the operators aware of the conditio This issue was dis-cussed with the YNSD Project Manager, who acknowledged +.he inspector's comments and concernt, and indicated that a revised memorandum will be issued. On July 29, 1988 YNSD issued memorandum YRP 902/88 that was fully responsive to the inspector's concern The inspector had no further questions on this matte . Plant Information Reports Plant information reports (PIR's) prepared by the licensee per AP-0004 were reviewe The inspector determined whether the conditions were reportable as defined in the TS and whether the licensee's system of problem identification and corrective action is bcing effectively uti-11 ze Ihe following PIR's were reviewed:
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PIR N Occurance Date Report Date Subject 88-02 3/23/88 4/22/88 Loss of primary vent stack sample line heat trace
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88-03 3/23-24/88 4/21/88 Unauthorized removal of the swinging gate and radiation area / radio-active material postings for the prima ry chem-istry sampling cubicle PIR 88-02: This event was reviewed in Inspection Report 50-29/88-05, Section 7h. Licensee corrective actions that have been implemented due to this event included modifying the primary auxiliary operator log to require a once per-eight-hour check of the heat tracing in the primary vent stack (PVS) monitoring building and a once per-shift check of PVS operabil'ty conducted by an RP technician to ensure the normal operation of the h0at trace system. The licensee also developed a descriptive label containing the information necessary for personnel to evaluate the oper-ation of the syste PIR 88-03: This event was reviewed in Inspection Report 50-29/88-05, Section 7 The PIR provides good documentation of the licensee's inves-tigation and corrective actions to prevent recurrence of this even The inspector had no further questions regarding this matte The inspector identified no violations regarding the licensee's actions associated with the events and noted that the licensee determined the cAuses of the occurrences and specified appropriate short-term and long-term corrective action . Part 21 Report The licensee issued a Notification of Potential Existence of a Design Defect in accordance with 10 CFR 21.21 (Part 21) on April 26, 1988. This Part 21 Report notified the NRC that the licensee considers that a design defect may exist with respect to Automatic Switch Company (ASCO) Tripoint SB12BMR/TL 10A22 pressure switche The licensee has experienced recurring problems with the main steam line pressure switches (Section 10 of Inspection Report 50-29/88-09). Recent testing and inspection of four failed pressure switches indicated the failures and increased deadband widening are due to the generic failure /
extrusion of the polyurethane dis The licensee experienced similar
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. 22 problems with these pressure switches in 1985 and inspections and test of those defective pressure switches showed the same problem; i.e., the polyurethane disc extruded around and on top of the piston in the '
cylinder. ASCO appears to not have given notification to other utilities of the possible pressure switch failure mechanism and therefore, the licensee submitted a Part 21 repor The inspector had the following concern. Although the licensee has exper-iened past similar problems with these same pressure switches and visual inspection revealed the same defect, the Part 21 report does not mention the previous pressure switch failure The fact that this is a recurring defect lends credence and inportance to the Part 21 report. The Part 21 report should have indicated that similar problems were experiencej with these pressure switches in 1985. The inspector had no further questions at this tim . Site Visit by NRC Commissioner and Staff On July 8, 1988, Cormissioner Kenneth Carr and an Executive Assistant conducted a visit to the Yankee NPS. A meeting was held with the resident inspectors and a site tour was conducted by the Plant Superintendent, j Following the tour, discussions were held with corporate and site manage- l ment on NRC and Yankee NPS issue . Management Meetings During the inspection period, the following managnent meetings were con-
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ducted or attended by the inspector as noted below:
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The inspector attended an exit meeting held on June 24, 1983 by an NRC:RI specialist inspe: tor at the conclusion of Inspection 50-29/
88-12, which included a review of the licensee's solid radvaste pre-paration, packaging and shipping progra The inspector attended an exit meeting held on June 24, 1933 by an
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NRC:RI op3rator licensing examiner at the conclusion of operator
- licensing examination 50-29/83-11 to discuss results and identify strengths /weakr e s ses in the licensee's operator licensing training
progra The inspector attended a treeting on June 22, 1988 with corporate <
management representatives to discuss items of mutual interes ,
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The inspector attended an exit meeting held on July 15, 1988 by an NRC:RI security specialist a t, the conclusion of Inspection 50-29/
88-13, which included a review of the security organization, security system power supplies, personnel access control and changes in vital area bevndaries.
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The inspectors attended a meetirg at the Region I Office in King of Prussia, PA on July 14, 1988 between Region I and Yankee Atomic Electric Company to discuss the results of the NRC Systematic Assess- ;
ment of Licensee Performance review conducted for the period of October 7,1986 through March 31, 1988. This assessment is docu-mented in SALP Report 50-29/86-9 The inspector attended an exit meeting held on July 22, 1988 by an NRC:RI specialist inspector at the conclusion of Inspection 50-29/
88-14, which included a review of the licensee's IST progra At periodic intervals during the course of the inspection period, meetings were held with senior f acility management to discuss the inspection scope and preliminary findings of the resident inspector l t
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