IR 05000482/1991021
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{{#Wiki_filter:_ _ - - - _ - - - - - _ _ - - _ _ _. _ _ _ _ _ _ _ _ , ,,, , - cp58 "%, UNITED STATES . O-t NUCLEAR RESULATORY COMMISSION l - cEGCN IV b , f s11 RYAN PLAZA DRIVE, SUITE 1000 s ARUNGTON. TEXAS 79011 , SEP 23 gg QCLuJUR2 i , MEMORAGUM FOR: Martin J. Virgilio Director, Project Directorate IV-2 Office of Nuclear Reactor Regulation ' FROM: Samuel J. Collins Director, Division of Reactor Safety Region IV SUBJECT: TIA REGARDING TECHNICAL SPECIFICATION REQUIREMENTS FOR THE SURVEILLANCE OF FUEL BUILDING INTEGRITY -- The purpose of this memorandum is to request technical information assistance in reviewing a potential safety-related matter concerning the Technical Specifications for PWRs. During a recent Fuel Integrity and Reactor Suberiticality (FIRS) inspection conducted at Wolf Creek Generating Station (WCGS), it was noted that there was an apparent inconsistency between PWR and BWR Technical Specification (TS) surveillance requirements for - verifying fuel building integrity (i.e., vacuum pressure).
(see attached NRC. Inspection Report 50-482/91-21).
l i In Section 3.9.13 of the WCGS TS, the operational requirements for the fuel building emergency exhaust system are provided. Among the surveillance l requirements, Section 4.9.13.g. states that at least once per 18 months the ' licensee shall verify that the system maintains the fuel building at a negative pressure of greater than or equal to one-quarter inch water gauge relative to the outside atmosphere during system operation.
Chapter 15.7.4.4 of the WCGS Updated Safety Analysis Report states that for the fuel building accident the only pathway for release of radioactivity from l ' the fuel building is through the emergency filtration system. Other. release paths, such as degraded door or air lock seals or inadvertently created release paths by maintenance or modification activities, are not analyzed.
The licensee has a quality control program at WCGS for maintaining the integrity of fire barrier doors. There is, however, no permanently installed pressure gauge in the fuel building. Therefore, the licensee has no explicit indication of fuel building pressure except when a temporary pressure gauge is installed for conducting the surveillance. Consequently, the licensee must rely on the door control program and quality controls over work activities for the assurance that the fuel building integrity has been maintained since the last surveillance test for pressure.
For Region IV piants, we have identified that the PWR TS requirements are less restrictive than are the BWR TS requirements. Table 1 provides an overview of the TS surveillance requi.rements for Region IV plants.
In contrast to the PWR TS surveillance requirements, the level of safety afforded by surveilling fuel building integrity in BWRs is much more 9807t00100 980624 ~ PDR FOIA UNNERST90-155 PDR _ a i W t U LF ' a-
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, demanding. For instance, the River Bend Station TS Section 4.6.5.2 requires that the licensee surve9 y fuel building pressure within 24 hours prior to and at least once per 7 days during handling of irradiated fuel in the fuel building. The licensee has a permanently mounted pressure gauge in its fuel building for this purpose.
These dichotomous TS surveillance requirements appear unjustified when one considers the difference between the PWR and BWR fuel handling accident consequences. A PWR fuel assembly has a potentially greater fission gas inventory than that of a BWR fuel assembly.
In addition, the PWR fuel assembly is, in some respects, more susceptible to mechanical damage than a -- BWR fuel assembly. This is because the fuel rod cladding is thinner and there is no channel box for impact protection. Therefore, it would appear that the radiological significance of a fuel handling accident in the fuel building is more severe in a PWR than in a BWR. With only these considerations in mind, one would expect that the TS surveillance requirement for fuel building integrity would be more stringent for PWRs than for BWRs. However, such is not the case, and brings to question the adequacy of the surveillance requirement of the PWR TS. Should PWR licensees be required to perform the fuel building integrity surveillance near to the time of fuel handling? My staff has discussed this matter with some NRR personnel and has been unable to find an explanation for this difference in TS surveillance requirements.
We would appreciate your assistance in resolving this matter.
Please contact Dr. Dale Powers of my staff (FTS 728-81g5), if you have any questions regarding this matter.
' irector Division of Reactor Safety
Attachment:
(asnoted)
REGION IV== l NRC Inspection Report No. 50-482/91-21 . Operating License No. NPF-42 Licensee: Wolf Creek Nuclear Operating Corporation (WCNOC) P. O. Box 411 Burlington, Kansas 66839 Facility Name: Wolf Creek Generating Station (WCGS) , i Inspection At: WCGS, Coffey County, Burlington, Kansas .._.
Inspection Conducted: August 12-16, 1991 Inspectors: Dale A. Powers, Senior Reactor Inspector, Test Programs Section Division of Reactor Safety j Howard F. Bundy, Reactor Inspector, Test Programs Section Division of Reactor Safety . Other Accompanying . . Personnel: Denise M. Garcia, Reactor Engineer Intern, Test Programs Section j Division of Reactor Safety f4 Mf/ Approved: - . W. C. 5eidle, CKief, Test Programs 5ection Date Division of Reactor Safety - - l Inspection Summary Insoection conducted Aucust 12-16. 1991 (Recort 50'482/91-21) ! Areas inspected: Routine, announced inspection of the licensee's in-core fuel l Toacing ano fuel storage configurations, fuel handling, core component performance, outage work controls end critical path scheduling, and potential l ! for fuel-related problems identified at other facilities.
Results: The licensee's fuel handling procedures were found to be comprehensive and provided sufficient guidance to minimize fuel damage. The engineering and operations staffs, which were interviewed, were knowledgeable of refueling activitie's and associated eauipment. The licensee had an efficient means for utilizing refuelin; 'eam personnel, who were recently provided with fuel handling training.
she licensee hao ennanced safety by establishing three non-required positicns for the uptcming refueling outage.
Those positions were the on-site outage manager, the containment coordinator, - - __ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _. _
____ _ _ _ _ _ - _. . _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ j ' . . _2 . ard an extra senior reactor operator (SRO) on each refueling team. The licensee was cognizant of the industry's fuel handling experience ar.d had implemented a program to evaluate such experience fer relevance to its l operations. The licensee had experienced problems with irradiated ccre components (i.e. cefective fuel rods and centrol rods). The licensee's corrective actions taken in response to its fuel problems appeared to have been prudent. The licensee's fuel design had evolved during each reload to a design that was predicted to be of improved performar.ce capability. The 11cersee had performed ell of its relcads utilizing the provisions of 10 CFP. Part 50.59.
The inspection found no abuse of those provisions. The licensee's critical path schedule was approved well in advance of the upcoming refueling cutage.
! Kithin the scope of the inspection, no violations or deviations were identified. The licensee was urged by the inspectors to consider two minor --- procedural revisions. The first was to have a procedure to implement the requirement of Technical Specification 3.9.5 (paragraph 2.5), and the second Wds to specify procedurally the verification of fuel building integrity just prior to fuel movement (paragraph 2.12).
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- DETAILS ! 1.
PERSONS CONTACTED WCNOC
- C. Perry, Director. Quality
- R. Holloway, Manager, Maintenance and Modifications
- M. Williams, Manager, Plant Support
- H. Chernoff, Supervisor, Licensing
- W. Norton, Manager, Technical Support
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- S. Wideman, Senior Engineering Specialist
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- 0. Maynard, Deputy Director Plant Operations
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- A. Payne, Manager Supplier-Materials Quality
- P. Martin, Shift Supervisor, Operations M. Megehee, Supervisor, Compliance
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- R. Hammond, Health Physicist
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- M. Dinglor, Manager, Nuclear Plant Engitieering-Systems
- W. Lindsay, Manager, Quality Assurance
- T. Orf, Licensing Engineer, Licensing
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- P. Adam, Supervisor, Reactor Engineering
- T. Daddens, Jr., Outage Manager D. Byerley, Senior Refueling Oreratur, Operations D.,Dullum, Industry Technical Information Program Coordinator, Compliance
- T. Garrett, Manager, Nuclear Safety Analysis
- R. Flannigan, Manager, Nuclear Safety and Engineering l
S. Ferguson, Technical Enginet*, Nuclear Safety Analysis , KEPCO -
- R. Ecklund, Electrical Engineer
- W. Goshorn, Wolf Creek Coordinator During the inspection, the inspectors also contacted other licensee personnel. *
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- Denotes those attending the exit meeting on August'16, 1991.
2.
FUEL INTEGRITY AND REACTOR SUBCRITICALITY (60705 AND 86700) The objectives of a Fuel Integrity and Reactor Subcriticality (FIRS) inspection , l are to review, inspect, and detemine the adequacy of the licensee's activities related to the protection of reactor fuel.
In general, the inspectors' review of procedures and records were not detailed in nature, but rather were broao overviews to detemine that essential issues were addressed.
Infomation on several aspects of the licensee's activities were baseo on interview statements - taken from licensee staff members and, because of time constraints, were not always verified by review of the Technical Specifications or the licensee's procedures and records. Emphasis, however, was given to reviewing the following areas: .
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. In-cere fuel loading and fuel storage geometrical configurations that
have not been specifically approved by NRC in safety evaluation reports (SERs) and that conceivably could result in situations involving I inadequate shutdown margin or inadvertent criticality; Operational work control practices, communications, procedures, physical
systems and equipment, and training which preclude unsafe fuel movements from occurring; Licensee evaluations and corrective actions that were perfomed
subsequen't to any self-identified problems that were indicative of accident sequence precursors or that had the potential to lead to fuel . damage; and .._.
The susceptibility of the licensee's operations, procedures, and
equipment to fuel-related problems that have occurred at other nuclear power plants.
'ihis inspection effort satisfied portions of the inspection requirements specified by NRC Inspection Manual Inspection Procedure 60705, " Preparation for Refueling," and Inspection Procedure 86700, " Spent Fuel Pool Activities." This - first phase of the inspection effort was perfomed when the plant was not in a refueling outage, so as to ninimize the impact on licensee resources. The final phase of the inspection effort will be performed during the upcoming refueling outage (i.e., Cycle 6 outage or, also referred to as Reload 5 outage) whe'n fuel handling operations are underway. The final phase of the inspection effort will concentrate on the observation of the licensee's activities as they relate to the implementation of procedural requirements and safe fuel handling practices. At the time of this inspection, some of the licensee's preparations for the Cycle 6 outage were incomplete. Incomplete items, for instance, included the finalization of the special nuclear material (SUM) mcvement plan.
Such items will be further inspected during the final phase of the FIRS inspection effort.
Attachment I to this inspection report is a tabulation of related documents -
reviewed by the inspectors during the inspection.
, 2.1 FUEL-RELATED INCIDENTS AT OTHER FACILITIES The inspectors discussed with the licensee several fuel-related mishaps that have occurred at nuclear power plants. Specifically, the incidents that were discussed are described in NRC Information Notices (ins) and Bulletins that ,
were issued during the past decade. Attachment 2 contains a listing of those I ins and Bulletins. The inspectors cuestioneo the licensee as to whether such publicized problems had occurred at WCGS. Also discussed were the licensee's rationale as to why WCGS equipment and operations did not have the potential for such problems, or what corrective actions had been initiatec at WCGS to preclude such problems from occurring. Those ins that resulted from the licensee's operations or those ins that the licensee evaluated and undertook action in response to their 1ssuance are discussed herein. After discussion with the licensee, other ins were viewed by the inspectors not to be directly i l l
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . ,. . . -5-applicable to the licensee's operations. This non-applicability was because the licensee's procedures at the time of the IN issuances should have precludeo such incidents from occurring.
The non-applicable INS are not discussed in this report. During this review, the inspectors identified no situation wherein the licensee's actions appeared deficient or unresponsive to the industry's documented experience.
Relating to IN 84-93, which involved industry failures of refueling water cavity seals, the inspector learned that the licensee intended to install a permanent cavity seal ring during the upcoming refueling outage. To determine the details, the inspector reviewed Plant Modification Request (PMR) 02493.
The seal ring is planned to replace the water cans for neutron shielding and the temporary inflatable seal with a permanent seal and neutron shield. Access covers will be removed during operation to provide a ventilation flow path.
. The inspector reviewed the safety analysis and noted that WCNOC was granted an exemption'to General Design Criterion 4 for the requirement for a flexible water bag neutron shield.
It appeared that all safety issues had been satisfactorily addressed.
As discussed below in Section 2.9 ins 87-19 and 89-31 either had airect implications to or resulted from WCGS operation.
In IN 89-51, the licensee was alerted to another licensee's violation of required shutdown margin upon the placement of highly reactive reloao fuel assemblies into intennediate locations during core alterations. The licensee's representative stated that WCNGC had subsequently made procedural changes to FHP 02-011. The procedural changes provided guidelines related to the conditions under which interim parking of fuel assemblies in the core were permissible during the fuel shuffling process. These guidelines are ciscussed below in Section 2.2.
In IN 90-77, the licensee was informed of another licensee's event that involved fuel assemblies that were inadvertently lifted from the core as the upper internals package was removeo. The licensee explained that its procedures should prevent this event from occurring at WCGS. However, the . licensee was considering some hardware changes that will physically require that the upper internals package is raised high enough to preclude interference curing horizontal movement.
2.2 SHUTDOWN MARGIN The licensee's representative asserted that there had been no known inciernts e at WCGS that involved inadequate shutdown margin. As discussed in Chapter 4.3 of the Updated Safety Analysis Report (USAR), the licensee has used the NRC-approved Westinghouse design codes LEOPARD and PDQ for reactivity calculations.
The inspector examined the licensee's controls for precluding an inadvertent j criticality and maintaining the required shutdown margin when interim fuel configurations were constructed in the reactor vessel during the fuel snuffling process. Procedure FHP 02-011 allowed such temporary alternate positioning l l l
__ - __ - ___ .-_ ___ o ,- . . -6- - provided that a subject fuel assembly was placed adjacent to the core baffle wall. The subject fuel assembly also was to be separated from the nearest fuel assembly by a minimum of one fuel assembly width and from the nearest cluster ! of fuel assemblies by a minimum of two fuel assembly widths, The licensee's procedure provided operator guidance to assist with a fuel assembly that was difficult to load (presumably the difficulty might be due to either twisting or bowing geometric distortions). For such situations when a temporary guide for loading was needed, the operator was to construct a box around the target location. The licensee's procedure gave restrictions on the fuel assemblies that could serve as guides. Considerations included enrichment limitations, and the accounting for the presence of burnable absorber or control rods.
j j The licensee had the capability of temporarily storing 2 fuel assemblies in the reactor cavity; therefore, shutdown margin in such borated water should not be - violated for storage in these locations.
2.3 Service Information on Fuel Handlino Eauipment . The inspector questioned the licensee's representative about the WCNOC process for handling vendor-supplied service information on their fuel handling equipment.
In particular, of interest was the thoroughness of the licensee's , process to ensure that adequate evaluations were conducted of safety-related, post-installation information that could result in the determination of , deficient equipment.
l The' inspector found that service infomation on fuel handling equipment' was reviewed in accordance with Procedure KGP-1311. Such information would normally be received as a technical bulletin or similar document. However, if such information were received as a vendor manual change, it would be processed as a plant configuration change in accordance with Procedure KGP-1131. The inspector noted that Industry Technical Infomation Program (ITIP) 01350 concerned a Westinghouse INF0 GRAM which recommended a backfit for the refueling machine load control system. This recommendation was being implemented in accordance with PMR 3477, which provided separate load switches for rooded and * -- unrodded fuel assemblies.
2.4 Commitments Related to Fuel Handlino Activitie's . In accordance with paragraph 13.5.1.1 of the USAR, the licensee committed to Regulatory Guide 1.33, "QA Program Requirements (Operation)," Revision 2, February 1978. The Regulatory Guide endorses ANS 3.2 - 1976, " Administrative Controls and QA for the Operational Phase of Nuclear Power Plants." The I standard requires written procedures for core alteration, accountability of l fuel, and partial or complete refueling operations. Specific procedures are ' recuired for each refueling outage and for receipt and shipment of fuel.
2.5 Work Controls, Responsibilities, and Delacations The inspectors reviewed the licensee's procedural delegations of responsibilities for refueling operations. The inspectors specifically examined documentation specifications related to select managers, licenseo , l __________ - _ ____
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. e a-7- . operators, and other key outage personnel. The inspectors paid particular attention to whether clear lines of authority and provisions for internal coordination had been pre-established for outage activities. The
responsibilities, communications, and sequencing of activities during refueling-relsted activities were established by Precedure FHP 02-001.
It referenced several supporting procedures.
'he licensee perfomed fuel movement in accordance with Procedure FHP 02-011.
This procedure referenced Procedure ADM 05-600 for fuel accountability control.
Fuel transfers during refueling outages were controlled by Fom FHP-M1, which was to be initiated by the supervisor, ru ctor engineering. The fom for the upcoming refueling outage had not been completed. Equipment operation during refueling were controlled by appropriate procedures as listed in the attachment ' of the procecure. New fuel receipt was controlled by Procedure FHP 01-001.
. Procedure FHP 01-002 covered shipment of new fuel offsite. The licenste had no plans for offsite shipment of spent fuel. The licensee's representative stated that no fresh or spent fuel had been transported from the site.
l The inspectors discussed with the licensee's representatives the makeup of the refueling teams. As described in Procedure FHP 02-011, the nomal refueling team consisted of 5 individuals, and the minimum refueling team consisted of 3 , individuals. A team nomally consisted of a refueling SRO, 2 upender operators, a refueling machine operator, and a spent fuel pool bricge crane operator. The procedure stipulated that the refueling SR0 must be in containment whenever a fuel assembly was being placed in or taken out of the core. The procedure also required that 2 operators must be in the fuel building and containment whenever fuel movement was underway. Two SR0s, with no other duties were assigned to each refueling crew. The remaining 3 members are not necessarily licensed. Pursuant to Technical Specifications requirements, at least one SRO was to provide oversight in containment whenever i fuel movements are underway. There was no regulatory requirement for a l licensed reactor cperator (RO) presence ir the fuel building when fuel i movements are underway.
In the licensee's plan, one of the 5 team members was always rotatino onto break or to another position. This practice minimized the l - potential for boredom.
- I The inspectors reviewed management's general working limitations for staff who would be perfoming various sefety-related tasks during the upcoming refueling outage. The inspectors questioned the licensee's controls on contract
personnel for consistency and uniformity with those controls placed on the
licensee's personnel. The licensee's representative stated that no fuel l handling was to be perfomed by contract personnel. Procedure ADM 01-023 ' provided reasonable restrictions on working hours to preclude excessive fatigue of the refueling crews. The refueling crews ncrmally worked 12 hour shifts.
The licensee's representative stated that, according to its procedure, apprcval for work duration in excess of 60 total hours per week was to be given by the , on-site outage manager.
Procedure FHP 02-011 placed an operational limitstion on refueling water clarity. The procedure specified that water clarity should be sufficient to enable observation of the index pins on the bottom of the core. The procedure l
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allowed for the relaxation of the water clarity criterien if the index pins . I were not observable but the fuel assembly top nozzles were observable. For this situation, however, the adequacy of the water clarity was to be determined I by Peactor Engineering. The licensee did not have a permanently mounted video camera on the refueling machine mast. The licensee's representative stated I that provisions were available to have an underwater camera en the refueling machine that could be used if need arose. Another underwater camera is to be located in the spent fuel pool area. The licensee's Procedure 04-001 stated that oinoculars were available to assist in the fuel handling activities.
Also, the inspectors examined the licensee's femalized measures for altering or suspending refueling operations in the event of various potential mishaps.
. Procedure OFN-003 required the suspension of all fuel movement and crane operation in the event of a tornado warning affecting the site. In addition, - Procedure ADM 02-011 requirec suspension of fuel movement for a number of reasons. Among these reasons were inconsistencies in the loading sequence or in the recording of fuel movements, fuel assembly damage, and abnormal increase in source range count rate. Appropriate evaluation and approval was recuired for resumption of fuel movement after suspension.
In regard to a loss-of-communication between the control room and the refueling
station, the inspectors noted that Technical Specification 3.9.5 required that ' when such direct communication could not be maintained, that all core alterations were to be suspended. The inspectors noted that the licensee did not.have an implementing procedure for this requirement. The licensee's representative stated, however, that fuel movements had been stoppec when such communication had been lost during a previous outage. The inspectors urged the licensee to consider a procedural revision that would be responsive to the Technical Specification requirement. The inspectors noted that the licensee had a concomitant procedural stipulation for movement of the upper internals.
. 2.6 Critical Path Schedulino The inspectors discussed with the outage manager the process whereby critical
- path schedules were developed. Planning, scheduling, and conducting outages were performed in accordance with Procedure ADM 01-108. Outage schedules were developed by an outage planning group which was chaired by the outage marager.
Members were provided from the functional groups such as operations, ! ' maintenance and modifications, and reactor engineering. Members and alternates were assigned in writing. Representatives from other groups were reauestec to.
attend meetings, as req! ired. The inspectors discussed, in general terms, tne means by which the licensee ensured necessary safety-related equipment availability and compliance with Technical Specification limiting conditions for operation during refueling outages. The inspectors were told that the ' licensee had often based its decisionmaking for scheduling maintenance activities on risk assessment.
The inspectors reviewed the specific critical path schedule for the Reloao 5 outage, and evaluated a sample of the planned concurrent activities for adverse impact upon safety equipment availability. The inspectors observed that the schedule had been approved on July 23, 1991, well ahead of the outage, which . .
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. . -9 was planned to commence about September 18, 1991. Mode and plant status infomation was listed above the work activities for each part of the outage.
The outage manager stated that plant safety requirements, such as onsite and offsite power supplies and cooling water sources, were considered in establishing the plant status requirements. The inspectors favorably noted that the majority of work on safety-related systems was scheduled to be performed with the fuel removed fem the reactor vessel. During the Reload 5 outage, the licensee planed to utilize the services of about 700 contract , persennel.
For the Reload 5 outage, the outage manager indicated that both an on-site outage manager and a containment coordinator had been assigned for each shift.
Personnel filling these positions would have authority to stop work when deemed necessary. The inspectors concluded that the designation of these positions -- should enhance the safety of outage activities.
2.7 Fuel Movement Plan The licensee was in the process of developing the SNM movement plan for the upcoming refueling outage. The approved plan will be inspected during the final phase of the FIRS inspection effort.
i l ' The licensee's core unloading and reloading practice has not employeo spiral loading and unloading of the core. Instead, the licensee has offloaded the core by first removing peripheral fuel assemblies and then removed the central fuel assemblies in the core. This practice resulted in a bridge structure remaining across the core for most of the offloading evolution. The bridge structure effect easured that a sufficient neutron flux was available for source range detector indication. The subsequent reloading process was the inverse of the offload process.
Procedure FHP 02-011 disallowed the freestanding of a fuel assembly.
Procedure FHP 03-013 provided the fuel handler with guidance on utilizing the fuel assembly loading guide (FLG). The FLG is a device that is placed on the bottom core plate to serve as a guide for assisting in the seating of a fuel assembly P - Other varied guidance was proceduralized to minimize loading / unloading damage such as that arising from spacer grid strap interactions.
The inspectors questioned a senior refueling operator on training that was performed or planned for important fuel-handling equipment and evolutions. The licensee's representative stated that about 2 weeks ago, 12 personnel who will be conducting fuel handling operations during the next outage received 3 days of training on Westinghouse fuel handling equipment in Pennsylvania.
In addition, the licensee's representative stated that cecasionally, fuel handling training and equipmerit checkout was perfomed using the licensee's dummy fuel assembly. The dumy assembly has always been the first assembly moved during core alterations.
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__ _ _ _ _ _ _ _ _ - _ _ _. j - . . . -10-2.8 Flow Blockage The inspector questioned the licensee on what controls were employed to keep loose ! parts from falling into refueling water during times when the reactor vessel head was removed. The licensee's representative responded that during refueling operation and prior to removal of the reactor vessel head, a house-keeping exclu-sion area was established at the refueling floor and at higher areas in contain-ment. The controls are given in Procedures FHP 02-001 and ADM 01-110. The inspector found that' the refueling cavity exclusion area was well established in accordance with Procedure 02-004 The procedure did not allow work on the refueling floor and above unless specifically approved by the containment coordinator. With the reactor vessel head removed Procedure FHP 02-004 stated that all tools and equipment taken into the exclusion area adjacent to the ' refueling pool and above that level that could find their way into the reactor l .. -. Vessel should be secured. It also required that pocket items be taped in j place, glasses be secured with a strap, and hard hat chin straps be utilized.
Procedure FHP 02-001 required tying or taping all tools, equipment, ano personal items prior to entering the exclusion areas at the refueling floor and above.
An SRO for refueling operations stated that it was the licensee's practice to tape clear plastic over the grate on the refueling bridge to prevent dropping small items into the refueling cavity.
. 2.9 Fuel and Core Component Problems at WCGS The inspector questioned the licensee's representative as to whether any fuel handling problems or adverse fuel behavior had been incurred at WCGS. In regard to fuel handling problems, the discussion included matters such as undue mechanical interference between spacer grids during fuel loading or unloading, physical damage incurred to irradiated core components, etc.
In regard to adverse fuel behavior, the inquiry included mechanistic effects such as fuel rod bowing (in excess of that predicted and accounted for by the fuel rod bowing rodel); fuel assembly twist or bow; abnomal cladding oxication; broken holddown springs; and cladding perforation due to collapse, fretting, fatigue, ballooning, or pellet-cladding-mechanical interaction.
- __ As discussed in IN 87-19, WCNOC was alerted to the potential for perforation and cracking of rod cluster control assemblies (RCCAs) in Westinghouse Nuclear l Steam Supply Systems (NSSSs). Such degradation had been found in some ' Westinghouse-designed plants and was attributed to coolant flow-induced vibration of the RCCAs when in the fully withdrawn position. The vibration had, in some cases, resulted in wall thinning and cracking of the stainless steel cladding of the RCCA rodlets. The thinning had occurred where the rodlets were in contact with the stainless steel guide cards, which axially positioned the RCCAs within the guide tubes. A severely degraded RCCA rodlet cladding was of concern because it could adversely impact rod worth (due to potential loss of the hafnium absorber material through a perforation) and l scransnability (because of potential hang up of the rodlet on the guide cards). As ! discussed later in IN 89-31, eddy current and profilometry measurements at WCGS resulted in the identification of abnormal wear and swelling of 53 RCCAs.
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-11 ! - The degradation was found to have occurred when the hafnium hydrided and subsequently swelled. It was postulated that the hydriding process became significant when the protective oxide on the hafnium was abraded away. The I abrasion mechanism was attributed to thermal cycling, and occurred at locations where the hafnium was in contact with the inner surface of the RCCA rodlet cladding.
Consequently, the licensee replaced all 53 RCCAs and has continued its surveillance program. The licensee's representative stated that in order to spread out any RCCA rodlet cladding wear axially en the rodlet that the WCGS Technical Specifications were revised to redefine the fully withdrawn position ' from the 231 step to the 222 step. This redefinition allowed the licensee the , flexibility to reposition RCCAs by one step about every 2 months. Subsequently, the licensee has replaced another 3 RCCAs curing refueling outages. The -- current surveillance program included the monitoring of hafnium in the reactor coolant system (RCS). The licensee had developed plans to inspect some RCCAs for conformance with the established integrity acceptance criteria during the Reload 5 outage.
The licensee's documentation. alto sumarized its activities resulting from fuel failures incurred in prevsous cycles of operation. Those activities included
ultrasonic and visual examination of irradiated fuel assemblies, reconstitution of fuel assemblies that were to be reused, and root cause analysis of some of the failure rechanisms. The licensee's records showed that, to date, there have been 2 fuel assemblies that were reconstituted.
In all, a total of 5 l defective fuel rods have been incurred at WCGS prior to the current Cycle 5 of j operation. Of those 5 defective fuel rods, 3 rods were investigated by failure ! analysis and replaced. The other 2 defective fuel rods were located in fuel ) assemblies that were to be discharged; therefore, no failure analysis was { conducted on them.
In regard to the 2 failed rods that were investigated, the I licensee concluded that 2 fuel rods failed because of manufacturing defects. One manufacturing defect resulted from an impurity in an end plug weld. Another manufacturing defect resulted from a lack of the necessary number of fuel pellets that were loaded into the rod. The latter problem manifested itself in* fuel rod perforation due to collapse.
In recard to the third defective fuel - rod that was investigated, the licensee was unable,to establish the failure mechanism.
The 3 defective fuel rods, which were investigated, were replaced with stainless steel rods. The licensee's representative stated that Westinghouse performed reviews of the peaking factor changes associated with those reconstitution.
The reviews revealed that the licensee's prior reload safety analysis remained bounding and that no operational limits needed revision nor was,the authorization for performing the reload under the provisiens of 10 CFR Part 50.59 adversely impacted.
l During operation in this current Cycle 5, the licensee had evidence from the RCS activity of failed fuel. Consequently, the licensee had developed plans to conduct surveillance of all Cycle 5 fuel after core unloading.
(All fuel is planned to be offloaded during the outage.) As indicated by the licensee's critical path schedule, the licensee intended to perform this surveillance, the
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- . -12- ' surveillance of RCCAs, and the reconstitution of fuel assemblies during the period of October 11 to October 19, 1991. The licensee's representative stated that Westinghouse contract personnel will perfom reconstitution activities, , and that the licensee will endorse via its quality assurance (QA) plan the ' necessary Westinghouse procedures to affect the surveillance ind reconstitution activities.
Other incidents, which have occurred at WCGS, have resulted in minor damage to irrad1ated core compcnents. Those incidents have invo1*;ed a fuel assembly that sustained a torn grid strap, a fuel assembly that was lowered upon another, and a RCCA handling tool that was moved while the RCCA was latched and positioned in a fuel assembly. The licensee's representative stated that one of the corrective actions resulting from the latter incident, was a comitment in response to Violation 90-16-01 to recuire (1) a person (spotter) to be present - to observe the operation and disengagement of the RCCA change tool and (2) a refueling SRO to be present in containment during use of the RCCA change tool.
2.10 Qualification Requirements The inspectors were informed that the licensee did not intend to utilize contractor personnel to manipulate fuel handling eouipment during the Cycle 6 outage. The licensee planned to utilized personnel who were not licensed R0s to perfom fuel manipulations during the Cycle 6 outage; however, such manipulations in the containment were to be perfomed under the oversight of a licensed SRO.
~ The inspectors questioned the licensee on what experience or training qualification requirements had been established to certify personnel to cperate fuel handling equipment. The fuel handling operators must satisfactorily complete a Watchstation Qualification Card that establishes the methoos, the responsibilities, and the requirements for the Nuclear Station Operator (NS0) Training Program. The qualification card records were being maintained by the Nuclear Training Division.
- 2.11 Fuel Storage Racks . The inspector was informed that the licensee's fuel storage racks for t, pent fuel assemblies have not been replaced or field modified since the time of operating license issuance. The licensee's documentation showed that the licensee anticipated that its spent fuel pool will provide spent fuel storage needs until 2005. The inspector inquired as to whether there were any administrative controls on the configurations of stored fuel or enrichment combinations. The licensee's representative stated that there were no regulatory restrictions and that the current safety analysis for the spent fusi pool bounded the maximum enrichment useo in WCGS fuel designs.
2.12 Fuel Building Activities , The inspector reviewed the licensee's procedural controls on materials that might be noved over the spent fuel pool. The procedures and interlocks in effect to keep unanalyzed loads (light and heavy) from crossing over the spent
fuel pool appeared adequate. This issue was previously reviewed in NRC Inspec-tion Report 50-482/85-04. The inspector confirmed that Procedure ADM 00-306 continued to require sufficient interlocks and administrative controls to t preclude an unanalyzed accident.
The inspector noted that Procedure FHP 02-001 required containment closure ! during fuel movement. Technical Specification 3.9.13 required 2 operable fuel I building emergency exhaust fans whenever irradiated fuel was in the spent fuel pool. Prior to fuel movement in the fuel building, Procedure FHP 03-007 recuired placing one train of fuel building emergency ventilation in service.
However, there was no requirement to check fuel building integrity in the procedure. Pursuant to Technical Specification 3.9.13, fuel building vacuum pressure was required to be checked at least on a 18 month frequency. The licensee did not have permanent pressure monitoring equipment in the fuel . building, but had installed such equipment when needed.
(The fuel building pressure must be at a negative pressure of greater than or equal to 0.25 inches water relative to atmosphere.) The inspector urged the licensee to consider revising the associated procedure to increase the frequency of this survei". lance or to recuire that the surveillance be performed near to the time that irradiated fuel is to be moved. An integrity check could result in the identification of a fission product release path not anticipated, nor described , in the USAR (i.e., such as a door not fully closed or a maintenance activity that resulted in an unplanned opening).
2.13 Boron Concentration . The inspector reviewed the licensee's Procedure FHP 02-001 for requirements pertaining to monitoring horon concentration during refueling operations.
Boron concentration cf the RCS was to be recorded and had to meet the Technical Specification criteria. According to Procedure ADM 04-021, while in Mode 6, chemistry personnel were required to detemine the baron concentration of the RCS and the refueling pool at least once each 72 hours.
In aodition, other precautions were specified to prevent inadvertent boron dilution. For instance, specified system valves were to be locked closed, and verified at least once each 31 days. The licensee's procedures appeared to be responsive - - to the Technical Specifications requirements.
. 2.14 Core Load and Verification - The inspector reviewed the licensee's core load procedure to determine whether clear instructions to minimize and identify loading problems were given.
Procedure FHP 02-011 required that reactor engineering personnel be stationed in the control room during all fuel loading. Section 4.20 of the procedure required that reactor engineering personnel were responsible for monitoring the inverse count rate (1/M) during loading and ensuring that the fuel shuffle be stopped if the count rate ratio fell below 0.35. The inspector noted that Section 7.3.1 of the procedure was contradictory to Section 4.20 in that Section 7.3.1 allowed reactor engineering personnel to delegate the monitoring of the source range count rate. The inspector found that two source range neutron flux monitors were required to be operable with centinuous visual indication in the control room and one with audible indication in the containment and control room.
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The inspector reviewed the licensee's fuel load verification procedure (FHP 02-011) for independence and adequacy. In summary, the procedure called for the video and audio recording of each fuel assembly identification number, and j insert type and identification number. The procedure stated that reactor engineering would perfom a secor.d core map verification.
2.15 Movement of Reactor Vessel Upper Internals Package The inspector reviewed Procedure FMP 02-013 for removing and replacing the reactor upper internals package. The procedure required operation of the polar crane at minimum speed to preclude damage to critical tolerance mating surfaces whenever the reactor vessel upper internals were being installed in, or lifted ' from, the reactor vessel. It also required constant monitoring of the load cell and the determination of the cause of any noted deviation. A safe load - path was procedurally outlined. Also, appropriate training of polar crane operating personnel and mechanics was specified. Quality control (QC) involvement was required for movements of the upper internals package.
Specifically, frequent QC dimensional and mechanical integrity checks were required during the removal and installation process. The procedure recuired connunications with the control room at all times that movement was in progress. On loss of communication, movements were to be suspended. A caution ' statement required a visual inspection if contact of the upper internals with another component during lateral transfer was suspected.
- 2.16 Fuel Assembly Desion Evolution
, The inspector questioned that licensee's representative as to whether WCNCC had implemented any fuel assembly field modifications or design changes. There have been no field modifications (asice from reconstitutiens) at WCGS. All WCGS reload fuel has been provided by Westinghouse. The licensee's representa-tive provided an overview of the evolution of the WCGS fuel design. Following the initial core load, each of the subsequently loaded fuel regions have involved some changes to the fuel design.
- The licensee's representative stated that these changes were perfomed under
- the provisions of 10 CFR Part 50.59. The changes have involved the incorporation of fuel pellet chamfering and length shorting, modification to the grid straps to reduce snapging, reduction of fuel rod pre-pressurization, alteration of the top nozzle to pemit removal from the assembly, alteration of the bottom nozzle to better trap debris, elongation of the axial gap between fuel rods and the nozzle to allow increased fuel growth associated with increased burnup, and modification of fuel rod and plugs. With the fuel design changes, the licensee has increased the batch-average discharga burnup from an original target value of 33,000 MWD /MTU to a current value of 43,000 MWD /MTU.
Currently, the licensee is considering a future replacement of its hafnium control rods with silver indium-cadmium control rods.
The licensee's documentation showed that WCHOC had commenced audits of the fuel vendor's operations in 1984 Currently, WCNOC was conducting about half a dozen audits of each fabrication batch. Audit teams were composed of individuals representing QA, nuclear services, and a consulting organization.
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15- -
- ens;e submitted on May 14, 1991, a proposed revision to the Technical cations that was necessitated by the new fuel design which was to be during the Reload 5 outage. The new fuel design, which is to constitute
, R;gion B fuel, was the Westinghouse Vantage 5H fuel design. The i .or noted that the Vantage 5H fuel design has been included in the USAR.
rd for the revision to the Technical Specifications was that the new fuel was to employ a slightly smaller diameter control rod guide thimble The smaller diameter was planned to result in increased scram times of the increased flow resistance in the thimble tubes. This proposed .n to the Technical Specifications was under review by the Office of R: actor Regulation. The licensee's representative anticipated the 1 sh:rtly.. , - pector noted that the Vantage EH fuel design also employed Zircaloy 4 The inspector questioned the licensee en the impact of using.a weaker terial on the core seismic analysis. The licensee responded that the margin of the Vantage 5H Zircaloy grid was superior to the original , grid.
In support of this assertion, the licensee stated that en 2, which required some plant-specific seismic analysis in the SER on tage 5 fuel assembly cesign, had been removea in the SER on the Vantage assembly design.
- p;ctor questioned the licensee on the degree of fuel assembly thimble ar (wall thinning) anticipated becasue of flow-induced RCCA vibration (see ion above in Section 2.9). The licensee's representative responded that ' eric safety analysis report on thimble tube wear in Westinghouse NSSSs cd upon 225 weeks of continuous rodded operation. For the current lanned operation of extended 18-month cycles for WCGS, no fuel assembly n;d for rodded operation in excess of 225 weeks.
est-Irradiation Examination
ens;e's documentation stated that the licensee followed Westinghouse's ' conditioning guidelines. The licensee's representative stated that the e trendr,d RCS activity (xenon, iodines, cesium, and some heavy metals) * - icatico of defective fuel.
, pector questioned the licensee on whether any corporate policy existed . dressed post-irradiation examination (PIE) of spent fuel. The licensee's - ntative responded that PIE is addressed, although in an implicit manner, Failed Fuel Action Plan. The inspector noted that the plan generally responsibilities related to fuel integrity to various licensee staff ns. Also, the plan presented four action levels associated with reactor specific fission products that might be sustained for a period of 7 ring steady state operation. The action levels were keyed to measured uivalent iodine. The most severe level (i.e., action level 4) was ! ted with the approach to the Technical Specification RCS activity limit.
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- vel was to result in, amongst other action, the consideration for
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- ing an unscheduled outage to replace the defective fuel.
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.. . . -16-As discussed in Section 2.9, the licensee had developed plans to conduct PIE of fuel assemblies and RCCAs during the Cycle 6 outage.
I No violations or deviations were identified in the review of this program area.
3.
Exit Meeting On August 16,1991, the inspectors met with members of the licensee's organization denoted in paragraph 1, and sumarized the scope and findings of this inspection. The licensee did not identify, as proprietary, any infomation used in the perfomance of this inspection.
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. ,. l ATTACHMENT 1 ! l Documents Reviewed
1.
WCNOC letter ET 91-0074, dated May 14, 1991 2.
WCNOC Procedure KGP-1311. " Industry Technical Information Program."
Revision 2. July 16, 1990 3.
WCNOC Procedure FHP 01-001, "New Fuel Receipt," Revision 13, change MA 91-039, June 12, 1991 4.
WCNOC Procedure FHP 01-002, " Shipment Of New Fuci Assemblies," Revision 1, June 1, 1988 , '
5.
WCNOC Procedure FHP 02-003, " Control Rod Drive Mechanism Ventilation - Removal And Installation," Revision 3. March 13, 1991 6.
WCNOC Procedure ;0M 02-210. " Operations Watchstation Qualifications," Revision 9. June 27, 1990 7.
WCNOC Procedure FHP 02-001, " Refueling Procedure " Revision 11, October 10, 1990 !- 8.
WCNOC Procedure FHP 02-004, " Refueling Cavity Exclusion Area," Revision 1. July 10, 1991 . ' 9.
WCNOC Procedure FHP 02-005, " Reactor Vessel Head Insulation Removal And Installation," Revision 2, August 15, 1989 10.
WCNOC Procedure FHP 02-006 " Instrument Port Conoseal Disassembly / Assembly," Revision 7. July 31, 1991 ' 11.
WCNOC Procedure FHP 02-007, " Reactor Vessel Closure Head Removal / Instal-lation," Revision 10, May 1, 1991 12.
WCNOC Procedure FHP 02-008, " Reactor Cavity Seal Ring Installation And
- Removal," Revision 3. Change MI 91-495, July 12, 1991 13.
WCNOC Procedure FHP 02-009, " Reactor Yessel Stud Removal, Cleaning, And Installation," Revision 7, March 20, 1991 14.
WCNOC Procedure FHP 02-010. " Reactor Vessel Closure Head 0-Rings Removal And Installation," Revision 6, September 12, 1990 15.
WCNOC Procedure FHP 02-012. " Control Rod Shaft Unlatching / Latching Toei Operating Instructions," Revision 5. January 8, 1989 16.
WCNOC Procedure FHP 02-013, " Upper Internals Removal And Installation," l Revision 6. June 5, 1991 17.
WCNOC Procedure FHP 02-011. " Fuel Shuffle And Position Verification," Revision 11. March 14, 1990 l
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' 18.
WCNOC Procedure FHP 02-015. " Calibration Of RCCA Latch / Unlatch Tool," Revision 1. February 26, 1990 19.
WCNOC Procedure FHP 02-014. " Surveillance Capsule Retrieval," Revision 1, November 3, 1926 20.
WCNGC Procedure FHP 02-016. " Lower Internals Removal and Installation," Revision 2. May 1, 1991 21.
WCNOC Procedure FHP 02-017, " Permanent Reactor Cavity Seal Access Covers," Pevision 0, March 27, 1991 22.
WCNOC Procedure FHP 02-018. " Removal And Installation Of CRDM Cables And Conduits," Revision 0, February 7, 1991 - 23.
WCNOC Procedure FHP 03-001, " Refueling Machine Operating Instructions," Revision 6 July 12, 1991 24.
WCNOC Procedure FHP 03-002, " Thimble Plug Handling Tool Operating Instructions," Revision 3, March 5, 1990 25.
WCNOC Procedure FHP 03-003 " Spent Fuel Assembly Handling Tool Operation," Revision 4. March 27, 1990 26.- WCNOC Procedure FHP 03-004, "RCCA Change fixture Operating Instru?tions,"
Revision 2, October 15, 1990 27.
WCNOC Procedure FHP 03-005, "EPRA Handling Tool Operating Instructions " Revision 4, March 5, 1990 28.
WCNOC Procedure FHP 03-006, " Fuel Transfer System Operating Instructions," Revision 4 September 21, 1990 29.
WCNOC Procedure FHP 03-007, " Spent Fuel Pool Bridge Crane Operating
Instructions And Daily Checks," Revision 10, March 29, 1990 - 30.
WCNOC Procedure FHP 03-008, " Cask Leading Crane Operating Instructions And Checkouts," Revision 8. February 9, 1990 31.
WCNOC Procedure FHP 03-009, "New Fuel Elevator Operating Instructions And Daily Checks," Revision 4, December 20, 1989 32.
WCNOC Procedure FHP 03-010 "New Fuel Handling Teol Operating Instructions," Revision 3. September 22, 1987 33.
WCNOC Procedure FHP 03-011, "New Rod Control Cluster Assembly Handling
l ' Tool Operating Instructions," Revision 2, April 2, 1991 l 34.
WCNOC Procedure FHP 03-012, " Rod Cluster Control Change Tool Operating Instructions," Revision 3, Change MI 90-485, May 21, 1990 , l t
... ,. . 1-l-3- ' 35.
WCNOC Procedure FHP 03-013, " Fuel Assembly Installation Guide Operating l Instructions," Revision 0, April 4, 1990 l< ' 36.
WCNOC Procedure OFN 00-018, " Fuel Hanoling Accident," Revision 4 May 23, 1990 l
37.
WCNOC Procedure FHP 04-001, " Spent Fuel Inspection," Revision 2, September 19, 1990 38.
WCNOC Procedure ADM 01-023. " Guidelines For WCGS Staff Working Hours," Revision 6. December 5, 1989 39.
WCNOC Procedure ADM 01-108, " Outage Planning And Implementations " - Revision 7 March 27, 1991 40.
WCNOC Procedure ADM 01-221. " Failed Fuel Action Plan," Revision.1, November 7,1990 41.
WCNOC Procedure ADM 05-401, * Reactor Engineering Personnel Qualification And Training," Revision 4, October 3, 1990 . 42.
WCNOC Procedure ADM 05-600, "Special huelear Material Safeguards And Accountability," Revision 7. November 14, 1990 43.. WCNOC Procedure ADM 04-020. " Chemistry Surveillance Program," Revision 21, March 21, 1991 44.
WCNOC Procedure ADM 04-021 " Chemistry Surveillance During Refueling Shutdown," Revision 4, October 27, 1987 45.
WCNOC Procedure ADM 08-306, " Control Of Heavy Loads And Special Lifting Devices," Revision 2, December 20, 1988 46.
WCNOC Industry Technical Infomation Program (ITIP), ITIP No. 01350, " Westinghouse INF0 GRAM: Refueling Machine load Control System Backfit," * - August 20, 1990 . 47.
WCNOC ITIP No. 00050, " Reactor Cavity Seal Failure Recomencation: Training And Pre-Fueling Briefings for All Personnel in Refueling Activities " January 30, 1985 48.
WCNOC Procedure SYS GG-200, " Fuel Building Emergency Exhaust Operations " Revision 4, June 9, 1991 49.
WCNOC " Critical Refueling Schedule," Revision 0, June 23, 1991 50.
WCNOC Procedure OFN-003, " Natural Events " Revision 10 51.
WCNOC Procedure ADM 08-306. " Control Of Heavy Loads And Special Lifting Devices," December 20, 1988 _ __________________ _ _
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lant F. modification 02493, "Rx Cavity Seal Ring (Remove WTR Cans)."
cvision 5. June 14. 1991 CNOC Procedure KGP-1131. " Plant Modification Process," Revision 7. PCN 3 .
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, I.
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- ATTACHMENT 2 j Fuel-Related Information Notices (ins) and Bulletins I A.
IN 81-23, " Fuel Assembly Damageo Due to Improper Positioning of Handling , Equipment" B.
IN 84-93, "Potertial for Loss of Water From the Refueling Cavity" (see also Bulletin 84-03) I C.
IN 85-12, "Recent Fuel Handling Fvents" D.
IN 86-06, " Failure of L1fting Rig Attachment While Lifting the Upper ) Guide Structure at St. Lucie, Unit 1" ' E.
IN 87-19. " Perforation and Cracking of Rod Cluster Control Assemblies" F.
IN 86-58, " Dropped Fuel Assembly" G.
IN 88-21. " Inadvertent Criticality Events at Oskarshamn and at U.S.
Nuclear Power Plants" H.
IN 89-31. " Swelling and Cracking of Hafnium Control Rods" - I.
IN E9-51, " Potential Loss of Requireo Shutdown Margin During Refueling Operations" (see also Bulletin 89-03) , , J.
IN 89-69, " Loss of Themal Margin Caused by Channel Box Bow" (see al'so Bulletin 90-02) K.
IN 90-77 and Supplement 1 " Inadvertent Removai of Fuel Assemblies from the Reactor Core" L.
IN 91-26. " Potential Nonconservative Errors in the Working Fomat Hansen-Roach Cross-Section Set Provided with the Keno and Scale Codes" -._.
- l l
l ! ..
_ _ _ _ _ _ _ ,,... ATTACHMENT 2 ' l Fuel-Related Infomation Notices (ins) and Bulletins A.
IN 81-23, " Fuel Assembly Damageo Due to Improper Positioning of Handling Equipment" i B.
IN 84-93. " Potential for Loss of Water From the Refueling Cavity" (see I also Bulletin 84-03) C.
IN 85-12. "Recent Fuel Handling Events" D.
IN 86-06, " Failure of Lifting Rig Attachment While Lifting the Upper Guide Structure at St. Lucie, Unit 1" E.
IN 87-19. " Perforation and Cracking of Rod Cluster Control Assemblies" F.
IN 86-58, " Dropped Fuel Assembly" G.
IN 88-21 " Inadvertent Criticality Events at Oskarshamn and at U.S.
Nuclear Power Plants" ' ' H.
IN 89-31 " Swelling and Cracking of Hafnium Control Rods" I.
IN E9-51, " Potential Loss of Requirec Shutdown Margin During Refueling Operations" (see also Bulletin 89-03) , J.
IN 89-69, " Loss of Themal Margin Caused by Channel Box Bow" (see also Bulletin 90-02) K.
IN 90-77 and Supplement 1. " Inadvertent Removal of Fuel Assemblies from the Reactor Core" L.
IN 91-26. " Potential Nonconservative Errors in the Working Fomat Hansen-Roach Cross-Section Set Provided with the Keno and Scale Codes" i S - l )
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i I NUCLEAR REGULATORY COMMISSION
wassmorow, o.c. anus
- ...*
April 10, 1992 MEMORANDUM FOR: Robert A. Capra, Director Project Directorate I-I Division of Reactor Projects I/II, NRR Walter R. Butler, Director Project Directorate I-3 Division of Reactor Projects I/II, NRR John F. Stolz, Director Project Directorate I-4 Division of Reactor Projects, I/II, NRR Elinor G. Adensam', Director Project Directorate II-I Division of Reactor Projects, I/II, NRR !
Ledyard B. Marsh, Director I Project Directorate III-I Division of Reactor Projects II, IV, & V. NRR Theodore R. Quay, Director Project Directorate V Division of Reactor Projects III, IV, & V NRR . THRU: Charles E. Rossi, Director - #er /////f p
Division of Operational Event ssessment , Office of Nuclear Reactor Regulation FROM: Christopher I. Grimes, Chief Technical Specifications Branch Division of Operational Events Assessment Office of Nuclear Reactor Regulat1on , SUBJECT: OPERABILITY REQUIREMENTS DURING TESTING AND REQUIREMENTS FOR ALTERNATE TRAIN TESTING Several concerns have been raised regarding the staff's position on entry into technical specification (TS) Limiting Conditions for Operation (LCO) when equipment is removed from service for testing. The staff's position on this matter was reiterated in a memorandum to S. J. Collins, Director Division of Reactor Projects, Region IV from C. I. Grimes, Acting Assistant Director for
Regions IV and V Reactors, Division of Reactor Projects III, IV, and V, Office ! ' of Nuclear Reactor Regulation, dated April 17, 1991 (Enclosure I). The staff's position is that equipment is inoperable when it is removed from , ervice and incapable of performing its safety function during the performance s .. .. m y.3gcogpzXM 'fp,
... . ,,
April 10,1992 of a TS surveillance requirement. This position was recently restated in Generic Letter 91-18, "Information to Licensees Regarding Two NRC Inspection , Manual Sections on Resolution of Degraded and Non-Conforming Conditions and on I Operability."
Some licensees have not been adhering to the staff's position, which can result in operation with a loss of safety function. For those licensees, implementation of the staff's position can add operating limitations. The purpose of this memorandum is to emphasize the need for adherence to the staff's position and to provide guidance on avoiding unnecessary operating limitations.
If the staff's position is not followed, loss of safety function could be caused by performance of a scheduled test when the opposite train is already . inoperable. Generally, this can be avoided while adhering to the staff's position by rescheduling the test within the TS schedule tolerance, which is 25 percent of the Surveillance Test Interval. However, some TS still include a requirement for a special test of the alternate train when one train is inoperable, and thus force a loss of safety function during the test such that shutdown may have to be initiated. That requirement reflects an earlier staff position that has been revised and not included in Standard Technical Specifications (STS) or newly issued TS for several years. Therefore, . licensees that have not already done so should be encouraged to request an amendment to delete such requirements.
OTSB and PD IV-2 have spot checked the TS for all plants licensed before 1981.
We found six plants which (1) have alternate train testing of systems in the l TS and (2) have not submitted an amendment proposal to remove such testing from the TS: Indian Point 2 & 3 H. B. Robinson Maine Yankee Palisades San Onofre 1 i Three Mile Island 1 l We request that the PMs and PDs contact these plants' and encourage them to prepose TS amendments to remove alternate train testing. Our spot check may l-not have identified all the plants which still have some alternate train test requirements. Therefore, we are discussing with PD IV-2 the possible need for generic communications on this issue.
The alternate train testing. requirement was included in early TS to provide a , positive demonstration that a loss of safety function had not c,ccurred. Later ' it was deleted with the realization that the added assurance by testing is not sufficient to justify the loss of safety function during the test, provided required periodic surveillance testing is current and that there are no known reasons to suggest that the alternate train is inoperable.
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April 10,1992 Requests for amendments should not include deletion of requirements for testing alternate emergency diesel generators. However, licensees that have not already done so should be encouraged to include a request for
modifications consistent with current STS requirements for testing alternate emergency diesel generators. That change would eliminate the need for such testing if an emergency diesel generator is taken out of service for preplanned preventative maintenance or testing.
If taken out of service for other reasons, testing of the alternate emergency diesel generator is required. However, most plant designs are such that an emergency diesel generator in the required test remains operable by being capable of responding to an emergency start signal.
Until an amendment is issued the staff's position and plant situation may - ' result in a need for temporary enforcement or licensing actions to avoid shutdowns that would not be required after the amendment is issued.
The Office of Enforcement concurs with this position. Should you have any questions regarding this matter, please contact Calvin W. Moon (301) 504-3136.
Original Signed by: C.1. Gritnes , Christopher I. Grimes, Chief Technical Specifications Branch Division of Operational Events Assessment
Office of Nuclear Reactor Regulation ' DISTRIBUTION: . WTRussell OTSB Members OTSB R/F DOEA R/F JGPartlow Background Books Central Files EHTrottier ' SAVarga ACThadani FMReinhart NConicella GCLainas BABoger CHBerlinger DHDorman JAZwolinski Central Files AEchaffee FWilliams MJVirgilio CERossi JLieberman GKalman CIGrimes RLEmch GRKingler RWHernan JACalvo RJGiardina CWMoon AMasciantonio WHBateman AGody MWHodges, RI CWHehl, RI DMCrutchfield LAReyes, RII AFGibson, RII EGGreenman, RIII . HJMiller, RIII SJCollins, RIV ABBeach, RIV RPZimmerman, RV l Document Name: A:\\PDGRIMES. CAL PMC's Diskette Marked " Moon" l
- See previous concurrence
' OTSB:DOEA:NRR OTSB:DOEA:NRR OTSB:DOEA:NRR C:0TSB:DOEA:NRR D: DST:NRR l RJGiardina* CWMoon* RLEmch* CIGrimes* ACThadani* 01/ /92 01/31/92 01/31/92 02/13/92 g 02/17/92 OE: D - J an
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0FFICIAL RECORD COPY Q V-Y
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_-______ __________________________ _ _ _ _ _ _ _ _ _ _ _ _ _ ,.- A
April 10,1992 Requests for amendments should not include deletion of requirements for testing alternate emergency diesel generators. However, licensees that have not already done so should be encouraged to include a request for modifications consistent with current STS requirements for testing alternate emergency diesel generators. That change would eliminate the need for such testing if an emergency diesel generator is taken out of service for preplanned preventative maintenance or testing.
If taken out of service for other reasons, testing of the alternate emergency diesel generator is required. However, most plant designs are such that an emergency diesel generator in the required test remains operable by being capable of responding to an emergency start signal.
Until an amendment is issued the staff's position and plant situation may result in a need for temporary enforcement or licensing actions to avoid . shutdowns that would not be required after the amendment is issued.
)
The Office of Enforcement concurs with this position. Should you have any ! questions regarding this matter, please contact Calvin W. Moon (301) 504-3136.
' - i i,.. A, Chris I. Grimes, Chief ) Technical Specifications Branch ) Division of Operational Events Assessment Office of Nuclear Reactor Regulation l .
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,. \\ UNITED 8TATES $ NUCLEAR REGULATORY COMMISSION $ - ! CASMlWGTON,0. o. 30665 \\, * =l*** April 17,1991 MEMORANDUM FOR: Samuel J. Collins, Director I Division of Reactor Projects Region IV FROM: Christopher I. Grimes, Acting Assistant Director for Regions IV and V Reacturs Division of Reettor Projects III, IV, and V Office of Nuclear Reactor Regulation SUBJECT: ENTRY INTO A TECHNICAL SPECIFICATION (TS) LIMITING CONDITION
FOR OPERATION (LC0) DURING THE PERFORMANCE OF SURVEILLANCE TESTING (TACNO.79763) . In response to your request dated February 11, 1991, we have reviewed the available guidance regarding the entry into Technical Specification Limiting Conditions for Operation during the performance of surveillance testing.
It is, and has been, the position of NRR that any surveillance or maintenance activity which renders equipment unable to perform its design function should - result in the equipment being declared inoperable and the licensee entering , the applicable Technical Specification Action Statements. This position is I considered consistent with the definition of operability which currently exists in the Technical Specifications at most plants (including Cooper).
l The fact that some licensees have not operated in accordance with this position
has led NRR to prepare generic guidance addressing this issue and other i operable / operability issues. Issuance, in the form of technical guidance for i the Inspection Manual, is expected in the near future. The proposed technical guidance, which reflects existing policy, states: "If technical specification surveillance require that safety equipment l be removed from service and rendered incapable of performing its safety function, the equipment is INOPERABLE. The LC0 action statement should ' be entered unless the technical specification explicitly direct otherwise (e.g. some instrumentation requirements). Upon completion of the , ! surveillance, the licensee should verify restoration to operation of at
least those portions of the equipment or system features that were altered l to accomplish the surveillance,' It should be noted that the above policy is also applicable to preventive maintenance (pM) or other planned activities that render equipment ingspable of performing its safety function.
The proposed Inspection Manual Technical Guidance points out that the policy is not intended to impose hardships on licensees' surveillance programs.
License amendments or procedure changes should be pursued if a surveillance procedure cannot be performed within the limitations of the existing technical specification LCO allowable outage tiee. Discussions with licensees should . '-
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.. arding expedite the implementation of the policy but enforcement actions reg involved existing or past practices need not be pursued unless such practices In oxtended equipment outages and an actual degradation in plant safety ft Features regard to a related issue at Cooper, several of the Engineered Sa e y be Technical Specifications require redundant trains to be " demonstrate This wording is generally when a train is declered inoperable.
l ith the referred to as alternate train testing and will introduce a prob em wNRR reco operable implementation of the above stated policies.
hi l with the licensee mention that other licensees have i The staff has found this approach to llow ment other than the diesel generators.
be acceptable and encourages Cooper to submit an ther unnecessary testing of safety equipment.
492-1133.
Address any questions regarding this response to Bill Reckley at please note that various programs such as Resident Inspector seminars ar f the policy.
planned to assist in licensee and regional office imple Inspection Manual Technical Guidance.
^^Qh i Christopher I. Grimes, Acting Assistant Director for Regions IV and V Reactors Division of Reactor Projects !!!, IV, and V Office of Nuclear Reactor Regulation C. W. Hehl, Region I ec: L. A. Reyes, Region II W. L. forney, Region III R. Zimmerman, Region V i I . ' ' ~ ~ - - - - - -. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
April 17,1991 ,- - expedite the implementation of the policy but enforcement actions regarding existing or past practices need not be pursued unless such practices involved s extended equipment outages and an actual degradation in plant safety. In regard to a related issue at Cooper, several of the Engineered Safety Features Technical Specifications require redundant trains to be " demonstrated" to be operable when a train is declared inoperable. This wording is generally referred to as alternate train testing and will introduce a problem with the implementation of the above stated policies. NRR recommends that discussions with the licensee mention that other licensees have pursued Technical Specification revisions which do not require alternate train testing for equip.
ment other than the diesel generators. The staff has found this approach to be acceptable and encourages Cooper to submit an amendment in order to allow an orderly implementation of the generic operability policy and.pravent other unnecessary testing of safety equipment.
Address any questions regarding this response to Bill Reckley at 492 1133.
Please note that various programs such as Resident Inspector seminars are planned to assist in licensee and regional office implementation of the policy.
OTSB is now the primary NRR contact for questions regarding the proposed Inspection Manual Technical Guidance.
m Lk Q Chr1stopher I. Grimes, Acting Assistant Director for Regions IV and V Reactors Division of Reactor Projects III, IV, and V Office of Nuclear Reactor Regulation DISTRIBUTION Central file CWHehl,* Region I PD4 1 Reading LAReyes, Region 11 B. Boger (M513E4))WLForney, Region III C. Grimes (M513E4 RZimmerman, Region V P. O'Connor D. Wigginton W. Reckley P. Noonan DFC ;PD4-1/LA :FD4 1/FE
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- /8/91 DFflCIAL RECDRD CDPY Document Name: MEMD GRIME 5/CDLLIN5/79763 l
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WASHINGTON. D.C. 20665 I . %*****/ ' June 22, 1992 MEMORANDUM FOR: Herbert H. Berkow, Director Project Directorate II-2 Division of Reactor Projects I/II, NRR FROM: Christopher I. Grimes, Chief Technical Specifications Branch Division of Operational Events Assessment, NRR SUBJECT: ST. LUCIE CONTAINMENT VESSEL INTEGRITY (TIA 91-23) (TAC NO. M82930) During a recent inspection of St. Lucie, Region II identified a concern on a Technical Specification interpretation with regards to TS 3/4.6.1.1 " Containment Vessel Integrity". The concern involves the status of vent, l drain, and test valves in containment penetrations with regards to containment integrity and how to treat these boundary conditions. The region requested guidance from NRR on this subject, by asking seven specific questions. OTSB . has reviewed the information provided by the region and provides the following ! responses to the questions: 1.
Confirm that vent, drain, and test valves in the containment penetrations are considered containment penetration valves affecting CONTAINMENT VESSEL INTEGRITY per TS 3.6.1.1 and that the 31 day verification of TS 4.6.1.1.a.1 applies to them.
Response: 10CFR Part 50, Appendix J defines Containment Isolation Valves (CIV) as any valve which is relied upon to perform a containment isolation function. The staff considers any valve associated with a containment penetration no matter how small to be a CIV. Therefore, vent, drain, and test valves are considered CIV's and the requirements of TS 3.6.1.1 apply to them.
2.
Determine whether or not the phrase " secured'in their positions" in TS 4.6.1.1,a.1 and in the definition of CONTAINMENT VESSEL INTEGRITY applies te valves, blind flanges, and deactivated automatic valves or just to deactivated automatic valves. Stated differently, must the subject vents, drains and test valves be locked or sealed.
Response: Standard Review Plant (SRP) 6.2.4 " Containment Isolation System" states the following with regards to closed manual valves and blind flanges: . Contact: R. J. Giardina 504-3152
A W 2 92/h
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l-2-June 22, 1992 . ~ " Sealed closed barriers include blind flanges and , l sealed closed isolation valves which may be closed manual valves, closed remote-manual valves and closed automatic valves which remain closed after a loss of coolant accident. Sealed closed isolation valves should be under administrative control to assure that they cannot be inadvertently opened. Administrative control includes mechanical devices to seal or lock the valve closed, or prevent power from being supplied to the valve operator."
Based on the above the phrase " secured in their position" applies to manual valves, blind flanges and deactivated automatic valves.
3.
Confirm that, for the containment spray lines, the area between the first valve outside the containment and the check valve inside containment are " containment penetrations" and that TS 3.6.1.1 and TS 4.6.1.1.a.1 apply to them.
Response: By the definition of CIV provided in the response to 1 l above, the first valve outside containment and the check valve inside containment are considered CIV's. Therefore, the area between the valves is considered a containment penetration and the requirements of TS 3.6.1.1 and 4.6.1.1.a.1 apply.
, 4.
Determine what leak testing must be applied to vent, drain, and test valves in the containment penetration areas.
Is the soap bubble test of TS 4.6.1.2 [e - Unit 1] [f - Unit 2] the appropriate leak test for these valves.
Response: 10CFR50 Appendix J and the Technical Specifications specify the leakage test and how they are to be performed. The soap bubble test is to be used on penetrations not individually testable by the applicable Type B and C tests, and is only good for detecting leakage at welded joints, flanges, and at valve stems. Therefore, it is not the appropriate test to determine leakage from the vent, drain, and test valves.
SRP 6.2.6 " Containment Leakage Testing" states the following with regards to leak testing of vent, drain and test valves: i " Leak testing.... test, vent and drain (TVD) connections that are used to facilitate local leak testing and the performance of the containment integrated leak rate test, should be under administrative control, and should be subject to periodic surveillance, to assure their integrity and verify the effectiveness of administrative. controls."
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_ __ ,. - -3-June 22, 1992 , ' 5.
Identify acceptable compensatory measures that might be used without entering TS 3.6.1.1 when desiring to open a vent or drain valve in a penetration area when containment integrity is required.
For example, is a dedicated operator posted by the valve with a radio, or outside penetration room door with a radio, acceptable.
Is a check valve inside containment an acceptable boundary such that no special compensation was needed in this instance.
Response: 10CFR50, Appendix A General Design Criterion 56 required CIV's both inside and outside containment. Enclosure 1 describes the compensatory measures that may be used in lieu of automatic actuation when manually operated test, vent or drain valves in a containment penetration are located outside containment and are required to be opened.
It also states the criteria used to determine whether the compensatory measures are applicable. Since the GDC requires two valves, a check valve inside containment would not constitute an acceptable boundary such that no special compensatory measures are needed.
6.
Identify any NRC requirements that licensees must clearly identify as having a containment integrity function (i.e. all valves and fittings involved in containment integrity).
Response: SRP 6.2.4 Section III " Review Procedures" states "Th'e CSB [ Containment Systems Branch) also ascertains that no single failure can prevent isolation of the containment. This is accomplished by reviewing the containment isolation provisions for each line penetrating containment to determine that two isolation barriers in series are provided....." SRP 6.2.6 Section III states for Type B tests that "All containment penetrations should be listed in the test program" and for Type C tests that "All containment isolation valves requiring a Type C test should be listed in the test program." By reference.to the piping and instrumentation diagrams and CIV listings provided in the Final Safety Analysis Report (FSAR) the reviewer confirms that all the above items have been listed. The Technical Specifications also contain a listing of CIV's which may not-include all the valves that are classified as CIV's by the plant licensing basis. Generally the FSAR identifies those , l valves that are classified as CIV's. Generic Letter 91-08 " Removal of Component Lists from Technical Specifications" removes the listing from the TS, but also modified the LCO, remedial actions and surveillance requirements so that they apply to all valves that are classified as CIV's by the plant licensing basis. These are the only requirements specified by the NRC which require the licensee to identify components or systems as having a containment integrity function.
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-4- ' June 22, 1992 ] 7.
Consider issuing the requested guidance as part of NRC Inspection Manual Chapter 9900 to ensure national consistency in this area.
Response: OTSB has been tasked with updating and revising the NRC Inspection Manual Chapter 9900. The task scheduled to begin in December,1992 will involve the review of all pertinent TS interpretation documentation to determine its current applicability and generic application to TS problems. This
document will be considered at that time for possible inclusion in Chapter 9900.
Christopher. Grimes, Chief ' Technical Specifications Branch Division of Operational Events Assessment, NRR Enclosure: As stated cc: G. C. Lainas J. Norris . I
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ h-Q f?&;A Novemoer 1,1990 ' l, p Enclosure 1 MEMORANDUM FOR: Richard Barrett, Director / ' Project Directorate III-2 Je/a ( d / c'U 'W - Division of Reactor Projects III, IV, V ( ph. 7. f.,,,,,, 4Q, and Special Projects . - J X-w y 7~ e FROM: Conrad E. McCracken, Chief gg 4 t [.f.e A. Plant Systems Branch - Division of Systems Technology ' * " " SUBJECT: ACCEPTABILITY OF DEDICATED INDIVIDUAL TO PERFORM PRIMARY CONTAINMENT ISOLATION FUNCTION (TAC N05. 77531 THROUGH77534) Reference: Memorandum from Edward G. Greenman to John A. Zwolinski, dated August 14, 1990, Subject: Request for Assistance Regarding Acceptability of Dedicated Individual to Perform Primary Containment Isolation Function In response to the referenced memorandum, the Plant Systems Branch (SPLB) is giving its view on the subject practice described below. As described in the reference memorandum, the drywell manifold sample system at Dresden and Quad Cities contains a number of sample lines. Each sample line has two manual containment isolation valves, which can not be isolated automatically when opened for sampling. The current practice at these plants is to provide a dedicated individual, who is constantly in contact with the control room, to close the manual isolation valves when containment isolation is required.
The SPLB Tinds that a similar practice has been allowed in the past for other plants where the existing design of g'ation valve, and a backfit of plant )),containmentpenetrationlines did not have any automatic containment iso codification was not justified. Additionally, in the proposed Standard Technical Specifications (STS) of the ongoing Technical Specification Improvement Program, the owners groups proposed to add a note in the STS allowing certain locked closed containment isolation valves to be opened intermittently under administrative control. These controls consist of stationing a dedicated operator, who is in continuous communication with the control room, to close the valve. The penetration can be rapidly isolated when containment isolation is indicated. Based on the past experience and proposed STS, the SPLB believes the current practice at Dresden and Quad Cities to be an acceptable alternate safety measures in lieu of the containment isolation provisions provided in Standard Review Plan (SRP) 6.2.4 and TMI Action Plan, Item II.E.4.2, which requires all nonessential systems be automatically isolated by the containment isolation signal.
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