IR 05000271/1992019

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Insp Rept 50-271/92-19 on 920901-04.No Violations Noted. Major Areas Inspected:Simulator Control Room,Emergency Operations Facility Performance & Operatons Support Ctr
ML20126H248
Person / Time
Site: Vermont Yankee Entergy icon.png
Issue date: 12/18/1992
From: Mccabe E
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20126H232 List:
References
50-271-92-19, NUDOCS 9301050114
Download: ML20126H248 (13)
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l U. S. NUCLEAR REGULATORY COMMISSION l

REGION I

Report: 50-271/92-19 License: D12-28 Licensee: yermont Yankee NucLar Power Corooration RD. 5. Box 102 Jrattleboro. Vermont 05301

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Facility: Vermont Nuclear Power Statiot)

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Dates: Srpfember 1-4. 1992 Location: litat'lebolo ..,1d Vernon. Vermont inspectors: C. G. Amato, Emergency Prepareaness Specialist (l. cad inspector)

S. Ilansell, Operator Licensing Examiner  :

P. liarris, Vermont Yankee Resident inspector D. Schultz, NRC Consultant (Comex Corporation)

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Approved: b M M TL '

E. C. McCabe, Chief, Emergency Preparedness Section date Facilities Radiological Safety and Safeguards 11 ranch Division of R' adiation Safety and Safeguards e

Sff!De Emergency Prenaredness (EP) program and the September 2,1992 partial participation EP exerds m; f

p The licensee -demonstrated the ability to implement the Emergency- Plan and Emergency Plan Implementing Procedures so as to protect public health and safety. Strengths were identified in Simulator Contial Rooin selection of the reactor shutdown process and in Emergency Operationsi ,

Facility performan. e. A weakness was noted in Operations. Support Center over-simulation of repair activities and repair personnel non adherence to relevant instructions 'and proccdure ,

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DETAILS 1,0 Persons Contacted The following Vermont Yankee Nuclear Power Corporation (VYNPC), or Yankee Atomic Electric Company (YAEC) staff members participated in the exercise exit meeting on September 3,1992, or in a September 4,1992 supplemental exit, or in a November 30, 1992 telephone exi I J. Babbitt, Technical Training Instructor, VYNPC t G. Bristol, Community Relations Coordinator, VYNPC _;

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T. Ilurda, Consulting Emergency Planner, Stone and Webster A. Chesley, Simulator Supervisor, VYNPC i J. Itawxhurst, Senior Engineer, YAEC  !

S. Jefferson, Assistant to the Plant Manager, VYNPC i E. Lindamood, Radiation Protection Supervisor, VYNPC W. Murphy, Senior Vice President, Operations, VYNPC J. Morarity, Security Supervisor, VYNPC R. Pagodin, Technical Services Supervisor, VYNPC ,

D. Porter, Technical Programs Manager, VYNPC _

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E. Porter, Emergency Preparedness Coordinator, VYNPC D. Reid, Plant Manager, VYNPC E. Salomon, Senior Engineer, YAEC M. Schneider, Manager of Communications, VYNPC J. Sinclair, Director of External Affairs, .VYNPC R. Sojka, Operations Support Manager, VYNPC R. Wanczyk, Operations Support Supervisor, VYNPC G. Weigand, President and Chief Executive Officer, VYNPC D. Weyman, Senior Environmental Program Manager, VYNPC The inspectors also interviewed and observed other licensee personne .0 Emergency lisereise An announced, partial participation' emergency preparedness exercise was held at the Vermont Yankee Nuclear Power Station on September 2,1992, from 8:00 a.m. to -

1:00 p.m. The States of New Hampshire and Vermont and the Commonwealth of Massachusetts participated. : There was extensive simulation'of protective actions. Off-site emergency p eparedness (EP) activities were not inspected, 2.1- - Pre-exercise Activitics The exercise objectives were submitted to NRC Region 1 on June 4,1992 and, as subsequently revised, were found appropriate. On July 3,1992, the licensee submitted the complete scenario package to the NRC. _ Initial NRC review of the scenario noted that the simulated conditions onlyjustiGed declaring an Unusual Event ,

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and an Alert, and was thereby weak in providing opportunity for participants to develop, maintain and improve key skills as required by 10 CFR $0.47(b)(14). No specific violations of NRC requirements were identified in this instance, liowever, non credible plant indications may leave evaluators unable to evaluate the adequacy of response in certain areas and also reduce the responding organization's ability to demonstrate their capabilities. That could increase the potential for a post-exercise determination that all major elements of the emergency plan had not been adequately tested. After telephone discussions with the licensee's EP staff, scenario revisions

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were made to simulate a Site Area Emergency. That was done by increasing containment radiation monitor readings by a factor of 1000 without the core being uncovered or clad damage sufficient to produce a gap fraction release giving rise to the containment monitor readings. Due to time considerations, further revisions to add realism were not made. The as-revised scenario was deemed adequate to meet minimum requirements for testing major portions of the Vermont Yankee Nuclear i Power Station Emergency Plan and Implementing Procedures. It also provided the opportunity for the licensee to demonstrate areas previously identified by the NRC as ;

in need of corrective hetion.110 wever, the licensee was advised that a more realistic scenario is needed to meet NRC and FEMA standards for the 1993 full-participation, ingestion pathway emergency exercis NRC observers attended a licensee briefing on September 1,1992. Changes to the scenario were discossed during the briefing. The licensee stated that certain emergency response activities would be simulated and that controllers would intercede in the exercise as necessary to prevent disrupting normal plant activitie .2 Exercise Scennrh The scenario included the following simulated events and condition * Initial (8:00 a.m.) conditions: reactor operating at steady state power for six months; a control rod drive pump was tagged out for repairs and scheduled for re'..irn to service within three hours; off-gas radiation has been increasing; and the Vernon tie line has been removed from service (return to service scheduled by early afternoon).

  • The "B" control rod drive pump tripped; two hydraulic control units (IICUs) were inoperable in a 9x9 fuel arra * An Unusual Event,

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valve * Reactor scra _ - . . _ ~ _ _ _ . . . ._ _ - - - . . _ . _ _ __

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  • liigh level alarms for drywell equipment and floor drain sump * An Alert.-
  • Torus pressure above 4 psig and drywell radiation at 450 R/h * Drywell radiation increased to 1000 R/h * A Site Area Emergenc * less of normal power on Bus The inspector compared the key events and players for the annual exercise scenario and for the preparatory licensee drill. While several aspects were common to both scenarios, the scenarios were found to be sufficiently

. different to provide an acceptable test for the annual exercis A review of the key personnel lists showed that three of four key nunagers played the same role during the exercise and preparatory drill. No associated violation of NRC requirements was identified, but rotation of drill / exercise players is an asset to developing the personnel defense-in-depth essential for successful accident management, The inspector asked the licensee to review this consideratio .3 Activities Observed During the exercise, the NRC observed activation, augmentation and functioning of the Emergency ~ Response Facilities and the Emergency Response Organization. The following were observe * Use of control room procedures, i c

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  • Detection, classincation, and assessment of scenario event * Direction and coordination of the emergency respons * Notincation of licensee and Massachusetts, New llampshire, and Vermont personnel; and communication of pertinent plant status information to Commonwealth and State personne * Communications and information flow, and record keepin * Assessment and projection of off site radiological dose and consideration of protective action * Accident analysi * Accident mitigatio .0 Gaulilallion of Eicrthc EhldhlES Emergency preparedness exercise findings were classi0ed as follow .1 Eurcise Strengthi performance that provided a strong, positive indicator of ability to effectively cope with an emergenc .2 Esercise Wenkness performance that could have precluded effective response to an aspect of an emergency. (An exercise weakness does not, of itself, constitute _ overall response inadequacy.) Area for Im11r11Ettic1111 performance which did not have a signincant negative ,

impact on emergency plan / procedure implementation, but which should be evaluated by the licensee to determine if corrective action could improve performanc .

4.0 Esercise Observntions The NRC team noted that the licensee's activation of the Emergency Response Organization, Emergency Resp (mse Facilities (ERFs), and use of these facilities were generally consistent with their Emergency Plan and Emergency Plan Implementing Procedures. NRC inspector observations and evaluations were made in the ERFs as described belo .1 Shimlator Cot 11rol Rqonl-The following exercise strengths were identifie . ,

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  • Selection of the plant shutdown process, liigh radiation in the drywell and two secondary containment areas indicated clad failure and a primary system leak into the drywell. Secondary containment radiation was near 1000 R/11 The Emergency Operations Facility (EOF) staff thought reactor vessel emergency deprenurization was needed. The Simulator Control Room staff advised the EOF that plant conditions and secondary containment control procedures called for plant shutdown and not emergency depressurizatio Operators correctly followed OE 3105, " Secondary Containment Control Procedure."
  • Emergency Operating Procedure (EOP) use that resulted in keeping the core covered and maintaining containment isolation, effectively safeguarding the public against a radiation releas The following expected actions were done wel * Plant operators recognized the entry conditions for applicable Emergency Operating Procedures (EOPs) and Abnormal Operating Procedures (AOPs).

The crew correctly performed all the required actions in the procedure When plant conditions warranted, the crew updated the Technical Support Center (TSC) to inform them of additional Alert classification dat * The Shift Supervisor (SS) briefed the crew on all significant plant condition changes that affected the event mitigation strategy, lie prioritized and remained informed of all troubleshooting and repair of inoperable safety equipment. The crew's communications throughout the exercise were noteworth The following areas were improved over previous performanc * The communicator (chemistry technician) performed all off-site notifications within the required time. lie was familiar with all communication equipment and used it with confidenc * The Shift Engineer (SE) kept the Shift Supervisor updated on plant radiation trends and primary containment parameters. The SE also recognized the failure of the "B" emergency dicsci to automatically start within seconds of the malfunction, a

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The following area for improvement was identifie * The crew did not detect a failure of residual heat removal pump "B" to automatically start in an appropriate amount of time: about 16 minutes elapsed before the crew recognized the failure of this pump to automatically star No exercise weaknesses were identifie .2 ItthDkal Stwngt1_ Center (TSC)

No exercise strengths, weaknesses, or areas for improvement were identified. The following expected actions were done wel * The TSC was promptly staffed. The transfer of Plant Emergency Director '

responsibilities from the Shift Supervisor to the TSC Coordinator (TSCC) was effective. The TSCC frequently briefed the staff on conditions, priorities, and accident mitigation. Congestion and noise were kept to a minimum, except during the period prior to TSC activation. It was observed, however, that a department manager was the primary communicator with the Operations .

Support Center and that such assignment might impede fulfillment of departmental emergency responsibilitie * The TSC staff demonstrated a conservative safety perspective. TSC directions to operators were sound and were carefully reviewed by the TSC staff prior to being communicated to the Control Room. Probing and detailed questions on component failures and equipment restoration were observe Event classincations and emergency action levels were frequently reviewed to assess worst case scenarios, potential release pathways, and subsequent component failures and plant challenges. Frequent and detailed communications on equipment status and corrective actions were established with the Site Recovery Manager and the control room. The TSC Coordinator effectively delegated responsibilitics, prioritized repair efforts, and made decisions based on engineering evaluations, plant procedures, and operational experienc * The Engineering Support Group (ESG) provided accurate and technically sound evaluations to the TSC, Independent efforts, such as core damage assessment and primary system leak location, contributed to analyses of the event However, an ESG recommendation regarding reactor vessel depressurization tvas not fully evaluated as to the overall effect on plant operatio .

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The following improved areas were identifie * Areas improved since the last exercise included the use of status boards and the accountability of personnel. Accurate and timely plant parameter information was frequently ulxlated on the status boards, but meteorological data w's not observed. The use of the site access computer and prompt management involvement contributed to the timely resolution of a personnel accountability deficiency (see Detail 6.0).

4.3 Operntluns SumwrLCenttr_f03.C)

Operations Support Center performance was marred by excessive simulation of activities and by non-adherence to procedures and good practice (especially in regard to selection of replacement components). The following basic requirements were involve * There were speciGc licensee written and oral instructions to controllers and players that, to demonstrate capabilities as realistically as possible, simulation of response activities was to be avoided wherever possibl * Paragraph C of Procedure Ap 0021, Work Orders, Revision 20, required initiation of an Emergency Work Order (EWO) to accomplish emergency work and to document that such work on safety-related equipment was accomplished. No exception to AP 0021 was identified as being granted, and 10 CFR 50.54(x) was not invoked to authorize deviating from or exceeding requirement Specific examples of related problems follo * Operating Emergency Procedure OE-3107, Revision 1, Appendix H, Vent the Control Rod Drive Over Piston Volume, required two auxiliary operators (AOs) to obtain the "HCU venting tools" located in the " friction testing tool bee in the reactor buildmg instrument and controls shop, along with sound powered phones for communications. A single AO proceeded directly to the area of the hydraulic control accumulators, and explained to the controller what he would do. When questioced about how he would hook up the vent hose to the withdraw riser bkxk, the AO did not identify or refer to OE-3107 but replied that he would use the vent adapter in the tool box that was up in the tool eag * The Emergency Diesel Generator "B" air stirt air 61ter differential pressure was simulated high, and the Glter was diagnosed as clogged. Repair work was initiated to obtain a new Olter insert, remove the installed filter, and install the new filter. No work instructions or post-maintenance test steps l

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9 were prepared. Rather, a short verbal description was delivered to the controller by the on-scene mechanic concerning what he would do to accomplish the tas * No tools or other materials (e.g., ladder to reach isolation valves) were used to repair the diesel air start filter. Similarly, although a spare breaker was >

available to perform required testing during the simulated work on the residual heat removal pump breaker, the spare was not use * An Auxiliary Operator, in order to repair the diesel air start filter, simulated shutting two air receiver isolation stop valves (supply to the air filter), and opening the vent / drain valve on the filter body to depressurire the filter, lie then stated that he had been authorized by the Shift Supervisor to act as a

" human tagout" for the repositioned valves (the NRC evaluators in the control room or TSC did not observe the granting of such permission). It was nfat possible to observe the supply valves and the vent / drain valve at the same time from the AO's location. Further, Procedure AP 0140, Vermont Yankee local Control Switching Rules, Revision 16, did not provide for a " human Ggout."

  • Paragraph B of Procedure AP 0806, issuance and Return of Material, Parts, and Components, Revision 7, required "all ... fields of the MPAC Stock issue Request (SIR) entry screen" to be completed (Maintenance Planning and Control Computer). The electronically prepared Warehouse issue Ticket (WIT) was then required before Stores personnel could issue any parts or material. These procedural requirements were not complied with to procure the replacement diesel air filter cartridge,
  • Repair of the diesel air start filter required identification and procurement of the filter cartridge for the Dollinger air filter assembly. The repair teant memler who was responsible for this activity was unable to access and use the appropriate computer data base for approximately 35 minutes. As a consequence, a non-controlled, superseded data base (IPI database of the now defunct MAXIM computer program) was used to identify the safety-related component and initiate procurement from stores. The inability of the MPAC computer operator to find and identify the correct stock number of the diesel air start filter for an extended period indicated inadequate training of

.mechani ca l ma itn enance personnelin use of the MPAC compute * Residual heat removal (RHR) pump motor supply breaker trouble-shooting identified blown closing circuit fuses. Electrical technicians did not identify the correct replacement fuses (the information available was that the fuses were 15 amp, with no fuse type information), and decided to install "like for like" fuses. The electrical foreman, when questioned about that practice,

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concluded he would obtain permission from the Shift Supervisor to obtain a set of non safety,4kV breaker closing circuitry fuses, and install them in the RiiR breaker. The inspectors observed no obtaining of such permission, and no identincation to emergency response management of the intention of using non safety fuses in safety equipment. Further, review of electrica maintenance procedures found no guidance on correct selection of fuses. The

" Fuse 1.isting," a controlled document (P.B. Corbett memo dtd 1/29/91, titled Controlled Fuse Document) did not specify the type of fuses for the R11R pump breaker. Since suf0cient information was not available at the scene, this matter should have been referred to the TSC. (Note: NRC Region 1 Division of Reactor Safety inspection No. 50 271/92 81 later identified fuse replacement provisions as a violation; correction of that violation is relevant to this aspect of OSC performance, but demonstration of proper fuse and other spare parts selection capability during the next NRC-observed emergency exercise remains appropriate.)

  • The foreman responsible for trouble-shooting the RiiR pump motor supply ,

breaker ordered a test of the racked-out breaker after replacement of the closing circuitry fuses. The technicians simulated wheeling the 4KV breaker to the Breaker Test Source (BTS) area, and then described to the controller in vague terms how they would perform an operational test before breaker re-installation in its cubicle. No procedure was used to perform the operational test. When questioned about the absence of a test procedure, the electrical foreman stated that such testing was within the skill of the craft, and that the least experienced technician in the crew could perform the testing Hawlessl Procedure OP-5222,4kV AC Circuit Breaker Inspection, Calibration,' and !

Testing, Revision 11, Section 6.1.4., Breaker Preliminary Operational Checks, covered the tests ordered by the foreman. It contained three pages of cautions, notes, and steps requiring specine pushbutton manipulation not easily committed to memory. Those were not routinely performed steps, and the inspector concluded that the breaker test was not a routine craft activity .

that was likely to be properly be performed without procedure us Collectively, the above items constitute a weakness in OSC conformance to instructions and procedures and a potential for inadequate plant con 0guration control in an emergency (IFI 50-271/92-19-01)

4.4 litungency Operations Fncility and Site Recovery Managu The following exercise strengths were identine * An ad hoc meeting of senior executives in the Senior Vice President's of0cc area led to declaration of an Unusual Event and a decision to activate

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Emergency Response Facilities as soon as possible and place them in standby readiness to respond to any degrading plant condition * There were frequent and detailed briefings of Commonwealth and State representatives in the EO * Thorough review of conditions which might lead to failure of the three Ossion product barriers and a need for protective actio * Very good analysis of the electrical problem and development of an excellent circuit Jiagram for the malfunctioning emergency diesel-generator and electric busse * The Resource Manager made extensive efforts to line up a diesel generator,

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fuel supplies, a day tank, and hose No exercises wealnesses were identiDe .

The following areas for improvement were identine * Noise control in the Site Recovery Manager's of0ce. (Too many individuals were talking too loudly and at the same time.)

  • Procedurally identifying the executive ad hoc committee and it's function * The extent of training of the Site Recovery Manager and staff, and the EOF  :

staff, in Emergency Operating Procedure analysis (licensee identined).

5.0 Kunttdat_ninLPerfunilance of DuticL0hhihig) ,

in order to determine if reactor operators were trained in the symptoms of and responses to Site Area Emergencies (SAEs) and General Emergencies (GEs), licensed operator-requalincation (LOR) training records were reviewed. Results of that record review for 1992 were as follows:

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LOR Cycle 92.h Each ~ w participated in two or three SAE scenarios, LOR Cycle 92.2; Each crew participated in one SAE scenari LOR Cyrle 923: Each crew participated in one GE scenari No training inadequacy was identified by the above revie ,

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. I 12 Dutigency Prutattdness nnd Securitv Intrrfact Section II.D.79 of Appendix 11 to 10 CFR 73 requires a security response to emergency situations other than security incidents. A new main guardhouse had been built and placed in operation. In addition, accountability has been " computerized." The Site Security j Manager was interviewed with respect to accountabilit .

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The inspector was informed that computerized accountability was used for the first time during this exercise. That accountability was accomplished in 30 minutes. However, this phase of the exercise was not observed by the inspector .0 Lkttut_Crillnur The NRC inspection team attended the licensee's exercise critique on September 3,199 At that critique, the licensee's lead controllers and observers discussed observati ons of the exercise. The licensec's controllers noted unnecessary entry into the Emergency Operating Procedures, scenario weakness in controls over excessive simulation, a need for improved training in the Materials, Parts and Components data base, and a need for procedure -

revision to assure proper safety-related fuse selection. Also the licensee noted the same OSC concerns identiGed in Detail 4.3 and initiated action to address these after the exit meetin .0 DDttgency Respor1Se Fncilltv IInhllahluly_

Ability of the Control Room and Technica: Support Center (TSC) to protect emergency responders from radiation was reviewed. That review identiGed reported dose values indicatirig shielding would limit Control Room staff radiation dose to the 10 CFR 50 Appendix A General Design Criterion 19 (GDC-19) value (five Rem whole body). .

However, no documentation was located indicating the ability of the TSC ventilation system ,

to limit thyroid dose to the equivalent of the GDC-19 live Rem whole body dose or to the licensee-committed 30 Rem value. This matter was identitled for further NRC staff consideratio . Exit Meeting Following the licensee's self critique, the NRC team met with the licensec's representatives listed in section i on September 3,1992 and September 4,1992. During these meetings, the NRC team leader summarized the inspection findings. Further discussions of the inspection findings were held by telephone with the licensee on December I and 15, 199 _ _ __._ _ ___ . - . . . _ .-_ ,_ -_ __. - _ .

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During the September 3 and 4 and December i discussions, the licensee was advised that the  !

preliminary NRC conclusion was that exercise performance demonstrated the ability to ,

implement Control Room procedures, the Emergency Plan, and the Emergency Plan implementing Procedures in a manner that would adequately provide protective measures for -

the_ health and safety of the publi ;

The licensee acknowledged the inspection findings and stated the intention of reviewing them *

further and taking appropriate actio On December 15,1992, the findings of the preceding report sections were highlighted to the licensee by telephon ,

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