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Insp Rept 50-336/90-11 on 900530-0711.Noncited Violations Noted.Major Areas Inspected:Plant Operations,Radiological Controls/Emergency Preparedness,Maint/Surveillance, Engineering/Technical Support & Security
	ML20056A692
Person / Time
Site:	Millstone
Issue date:	07/31/1990
From:	Haverkamp D; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	;
Shared Package
ML20056A691	List: IR 05000336/1990011; ML20056A690;
References
	50-336-90-11, NUDOCS 9008090042
	Download: ML20056A692 (30)
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U.S. NUCLEAR REGULATORY COMMISSION q REGION I

 ' Report No.:' _50-336/90-11 Docket No.:- 50-336 License N DPR-65 Licensee:. Northeast Nuclear Energy Company
  ).0. Box 270
,'   Hartford, CT 06141-0270 Facility Name: Millstone Nuclear Power Station, Unit 2 Inspection at: Waterford, Connecticut c  Dates: May 30 - July 11, 1990
 - Repori,ing Inspector: Peter J. Habighorst, Resident Inspector Inspectors: William J. Raymond, Senior Resident Inspector Peter J. Habighorst, Resident Inspector Guy S.=Vissing, PDI4, Nuclear Reactor Regulation David M. Johnson, Resident Inspector, Three Mile Island Approved by: ebb Donald R.-Haverkamp, Chiff-
    , _ - 7h I b u Date Reactor Projects Section 4A
>   Division of Reactor Projects n
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Inspection Summary: Inspection on May 30 - July 11,1990 Inspection Report No. 50-336/90-11

 ' Areas Inspected: Routine NRC resident and specialist inspection of plant
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operations, radiological controls / emergency preparedness, maintenance /sur-veillance, engineering / technical support, security, and safety assessment / quality verificatio Results: See, Executive Summary-i- 9008090042 90000J -

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Executive Summary Plant Operations: A licensee identified, non-cited violation was noted for a licensee failure to have a non-intent procedure change approved by the plant operations review committee within the required fourteen day Inspector review of licer.4ee performance on the timeliness of approving non-intent procedure changes identi-fied no programmatic concern Radiological Controls / Emergency Preparedness: Routine review in these areas identified no noteworthy finding Two previously identified items were close Surveillance and Maintenance: Adequate corrective maintenance on the atmospheric dump valves was note Action to correct a dump valve positioner problem was still in progress at the end of the report period. One previously identified surveillance item was close Wide range nuclear instrumentation calibration and procedure changes were reviewed. No unacceptable conditions were identifie Security: Routine review in this area identified no noteworthy finding Engineering and' Technical Support: Reactor quadrant power tilt has increased slightly during Cycle 10, but has remained below the technical specification limit of 2 percent. The licensee demonstrated that safety limits are met and margins are maintained. Licensee actions to identify the quadrant tilt, to evaluate its cause and significance, and to plan corrective actions were thorough and extensiv The reactor is operating with 29 control element assemblins that are the same design which failed at another facility, The licensen has concluded, based on previous examinations and present evaluations, that continued operation until the next refueling-outage is acceptable. NRC review of this issue was still in progress at the end of the inspection perio NRC review of safety evaluations for various plant modifications found that some did not document a thorough justification to support the conclusion that the criteria of 10 CFR 50.59 were met; however, reasonable assurance existed such that no modifications were unreviewed safety question ES-1 ,

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Safety Assessment / Quality Verification: Quality Services Department audits and Nuclear Review Board reviews of safety equipment' lists, and the implementation of technical specifications and the equipment environmental qualification program were comprehensive and had generally good finding A licensee identified, non-cited violation was noted for failure to adhere to a technical specification limiting condition for operation for a process radia-tion monito , ES-2

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TABLE OF CONTENTS Page 1.0 Summary of Facility Activities.............................. 1 2.0 Plant Operations (IP 71707/93702)........................... 1

2.1 Control Room 0bservations.............................. 1 2.2 Plant Tours............................................ 1 2.3 Review of Plant Incident Reports ...................... 2 2.4 Previously Identified Items............................ 2 2.4.1 (Closed) Unresolved Item 50-336/87-25-04: Control Room Air Inleakage Testing and Improvements to-Ventilation System..................................... 2 2.4.2 (Closed) Unresolved Item 50-336/89-13-13: Inadequate Procedural Guidance to Operations Personnel Regarding Use of Radiation Monitor Alarm By-Pass Keys., 3 2.5 Non-Intent Procedure Change Timeliness................. 3 3.0 Radiological Controls / Emergency Preparedness (IP 71707). . . . , 4 3.1 Posting and Controls of Radiological Areas............. 4 3.2, Previously Identified Items............................ 5 3.2.1 (Closed) Inspector Follow Item 50-336/86-16-03: Complete'Insta11ation of the Health Physics Network Dedicated Commercial Telephone Lines................... 5 3.2.2 (Closed) Unresolved Item 50-336/89-24-01: Preplanned Alternate Monitoring Method'for High Range Noble Gaseous Monitor............................ 5

 '4.0 Maintenance / Surveillance-(IP 62703/61726/92702)............. 6 4.1- Observation of Maintenance Activities.................. 6 4.2 Observation of Surveillance Activities................. 6 4.3 Previously Identified Items............................ 7 4.3.1 (Closed) Violation 50-336/89-16-01:

Incomplete Technical Specification Surveillance for Engineered Safety Actuation System Bistable Actuation System....................................... 7 4.4 Steam Generator Atmospheric Dump Valve Anomaly......... 8 4.5 Calibration of Wide Range Nuclear Instrumentation and Procedural Changes (A.49.01)....................... 9

*  5.0 Engineering / Technical Support (IP 37700/92702).............. 10 5.1 Azimuthal Power Tilt Indication........................ 10 5.2 Engineering Support of Plant Operations... ............ 12 5. Control Element Assembly Failure Followup. . . . . . . 12 T-1
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5.3: : Steam Generator Tube Repair. Activities. ... . . . . . . . . . . . . . .-- 13 -

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,  sf   :5.41 Previously Identi fi ed I tems . ... . . . . . . ... . . . . . . . . ... . . . . . . . :   -15
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5;4~.1 :(Closed) Unresolved Item 50-336/89-24-06:-

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Licensee Event Report:89-011-00 Service Water-

     . Valve Reportabi-lity Timeliness.........................
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7.3.' Licensee Event Reports;........................-........ 20~

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   ',  '7.'4 Reportable Events...... ...............................:    22-  ,

4F 7. PIR.90-55: _High Energy Line Break: . : S .' An a ly s i s As sump t'i on Er ro'rs . . . .. .. . . .. . . . . . . . . . . . . . . . . . . . . . 22' , qE , 7.4.2;fPIR 90-48: Par _tial Engineered Safety. Feature-Actuation................................. ............- 22 o- 7. ' a g: PIR 90-42: :10 CFR 21 Notification-on Main 3R Steam: Safety Va1ves.................................... 23

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c7.4.4 ' PIR '90-53: Missed GrabfSampleLon RM-8132A/8,.... 24

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    -7.-5 ' Previ ou sly Identi fi ed 'I tems . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24
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     '7.5.1-J(Closed)' Unresolved Item 50-336/89-01-02:

LE . }h) Mechanism for Coordinating ^a Quality Assurance ;

     . 0verview of .Sa fety-Related' Activi ties. . . . . . . . . . . . . . . . . .
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DETAILS 1.0' Summary of Facility Activities Facility Operations Millstone _ Nuclear Power Station (Millstone 2 or the plant)-was in cold shutdown at the beginning of the inspection period. The outage began on May 8 to identify.and repair the source of primary-to-secondary leakag , The plant commenced a plant startup on June 12, achieved full rated power on June 20, and remained at rated power throughout the remainder of the " inspection perio NRC Activities The inspection activities during this report period included 173 hours of inspection-during normal working hours. In addition, the review of plant operations was routinely conducted during periods of backshif ts (evening shifts) and deep backshifts (weekend and midnight shifts). Inspections were performed during.10 backsh!ft hours and during'two hours of deep backshif A Region I specialist inspection o'. operator requalification exams was conducted.between June 18-22, 199 Results are provided in Inspection Report 50-336/90-10(0L).

A Region I specialist inspection of engineering and technical support was conducted between June 11-15, 1990. Results are provided in Inspection Report 50-336/90-1 , 2.0 Plant Operations, 2.1 Control Room Observations Control room instruments were observed for correlation between chan-nels, proper functioning, and conformance with technical specifica-tions. Reactor, electrical, and safety system lineups were verified to be proper using indicators on the main control boar Alarm conditions in effect and alarms received in the control room

were discussed with operators. The' inspector periodically reviewed the night order log, tagout log, plant incident report log, key log, and bypass jumper log. Each of the respective logs was discussed

with operation department staf f. No noteworthy observations were mad .2 Plant Tours The inspector observed plant operations during regular and backshif t tours of the following areas:

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2-m o, . C'ontrol . R'oom Containment Vital Switchgear Room Diesel Generator Room Turbine Building Intake Structure ! Enclosure Building ESF Cubicles

During plant tours, logs and records were reviewed to ensure compliance with station procedures, to determine if entries were .,

correctly made, and to verify correct communication and equipment statu No noteworthy observations were mad .

2.3 Review of Plant Incident Reports The plant incident reports (PIRs) listed below were reviewed during the inspection period to do the following: (1) determine the signi-ficance of the events; (ii) review the licensee's evaluation of the 1: ' events; (iii) verify that the licensee's response and corrective actions were proper; and,-(iv) verify that the licensee reported the

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events in accordance with applicable requirements, if require The PIRs reviewed were:

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PIR 90-42, "10 CFR 21 Notification on Main Steam Safety Valves"- J' -- PIR 90-48, " Inadvertent Actuation of 'B' Enclosure Building Filtration System"

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PIR 90-49, "'A' Steam Generator Feed Pump Turning Gear Failure"

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PIR 90-50, "RM-8132A Flow Calibration"

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 -PIR 90-51, "'B' Emergency Diesel Generator Remote Manual Voltage Control"
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PIR 90-53, " Missed Grab Sample on RM 8123A/B"

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PIR 90-54, "Non-intent Procedural Change without Plant Operations Review Committee (PORC) Approval within 14 days"

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PIR 90-55, "High Energy Line Break Discrepancies" ';

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PIR 90-56, " Unexplained Closure of Intermediate Stop Valve"

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 .The.following PIRs warranted additional inspector fol. low-up: PIR 90-42 (see report' detail 7.4.3); PIR 90-48 (see report detail'7.4.2);

PIR- 90-53 (see report detail 7.4.4); PIR 90-55 (see report detail 7.4.1) and PIR 90-54 (see report detail 2.5).

2.4 Previously Identified Items 2. (Closed) Unresolved Item 87-25-04: Control Room Air Inleakage Testing and Improvements to Ventilation System This item concerned licensee actions to eliminate the source of radioactive air inleakage into the control room ventilation ducting on October 30, 1987, and verification of plant opera-tions in conformance with license conditions in regard to con-trol room habitabilit Inleakage of radioactivity into the control room was identified J by the licensee to be from a letdown sample valve inadvertently left open at the primary sample sink, which allowed reactor ,- coolant system (RCS) water to flow into the auxiliary building

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drain-syste The RCS water entered the aerated-waste drain-tank and degassed. The radioactive gas was subsequently trans-ported to the -control room air conditioning (CRAC) room through three floor drains. The gas transport occurred as a result of-a , small negative differential pressure (.08 inch water gauge) T between the aerated waste drain tank room and the CRAC roo > i: Licensee action'to prevent recurrence of this event included installation of loop seals in the floor drains in the CRAC room and repair of the control room ventilation system duct wor The corrective maintenance work on the associated duct work , eliminated the differential pressure between rooms (a'erated ! waste drain tank and.CRAC). The inspector verified loop seals ~ were installed in~the CRAC room and that differential pressure : between the two affected auxiliary and control building areas ' remained within intended limits through periodic inspection of-the traps during plant tour Inspector review of the licensee's assessment of control room

 ,  habitability during the October 30, 1987, event identified that
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, all license conditions were satisfied; specifically,10CFR20 and *

emergency procedure EPIP 4701-6 offsite release limits were met, technical specification air inleakage limits were satisfied, and

radiation-exposure to control room operators was much lower than regulatory limits. This item is close !

2. _(Closed) Unresolved Item 50-336/89-13-13: Inadequate Procedural Guidance to Ope _ rations Personnel Regarding UseLof -Radiation Monitor Alarm Bypass Keys

This item concerned inadequate procedural guidance to operations personnel regarding the use of bypass keys to bypass-local area radiation monitor alarms, i Station procedures OP-2383A and OP-2383B have been revised to require that local radiation alarms, when placed in' bypass to silence the alarm, must be returned to service when the alarm condition clears. This procedural requirement applies whenever an operator has used the bypass key to respond to-a local alar This item is close ,5 Non-Intent Procedure Change Timeliness On June 21, the licensee-identified a non-intent procedure change i that was not reviewed by the plant operations review committee (PORC)

within fourteen days of issuance. The deficiency was documented by the licensee in plant incident report (PIR) 90-5 Procedure EN-21132 Change 3, Revision 5, " Service Water Operational Test" was approved by senior reactor operators on June 5,1990, and approved by the PORC on June 2 !

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The inspector reviewed the non-intent change log to assess licensee performance on timeliness of PORC approval of procedural changes between 1989 and 1990. The inspector noted that less than 3 percent of the changes (3 of 123 changes) were untimely, all of which were internally identified'by the licensee via PIR's (PIR 90-54, 89-117, and 89-55).

Licensee corrective actions' included the following: logging of-initial change approval on station form SF-302, section 0, tracking of procedure changes in the control room shif t supervisor's 'n'on-intent change' log, and. department head /PORC weekly review of out-standing non-intent procedure changes. The-inspector considered these actions' appropriate to preclude recurrence of the late review .The inspector reviewed the licensee reportability determination of PIR 90-54 using emergency plan implementing procedure (EPIP) 4701-4 page 12, section'IV.a., 10CFR50.72, 10CFR50.73, and NUREG 1022 sup-plement 1. The licensee decision not to report this issue was ap-propriate, since the late procedural change approval is an adminis-trative violation'.that did not directly affect plant operation, and the events did not result in operation directly prohibited by the technical specification Conclusion The failure of PORC to approve a non-intent procedure change within fourteen days is prohibited by technical specification 6.8.3.c- and administrative control procedure 3.02, step 6.8. The NRC en-courages and supports licensee-initiatives for self-identification and correction of' problems. In accordance with 10CFR 2 Appendix C, subpart'G.1~, -a notice of violation is not being issued since the five criteria for exercising enforcement discretion were fulfilled by the licensee. This noncited violation is closed (50-336/90-11-01).

3.0 Radiological Controls / Emergency Preparedness 3.1 Posting and Controls of Radiological Areas During plant tours, contaminated, high airborne radiation, and high radiation areas were reviewed with respect to boundary identifica-tion, locking requirements, and appropriate control point No in-adequacies were note l

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Installation of the Health Physics Dedicated Commercial lelephone Lines This item was opened due to the lack of a dedicated health physics network (HPN) phone line at Millstone 2. Previous NRC inspection-closed this item at Millstone.1 and 3 in-inspection-reports 50-245/88-15'and 50-423/88-24, respectivel The_ inspector verified the installation of the HPN communica-tion line in the Millstone 2 designated operational support 9 cente The operability of the HPN lines was verified by document review of monthly surveillance checks (EPIP 4602-1), and observation of a successful surveillance on July 9,199 > ; This item is close .2.2 (Closed) Unresolved Item 50-336/89-24-01: Preplanned Alternate Monitoring Method for High Range Noble Gaseous

Monitor Previous NRC review of the operability evaluation for the high range noble gas effluent monitor (RM-8168) identified defic- , iencies in the alternate preplanned monitor metho The specific. deficiencies identified in the alternate monitoring channel were: inability of control room operators to appro-priately. classify emergency events; lack of equivalent opera-tional range for the monitor;:and the potential rad?ation hazard to acquire' data-during~an emergency even Licensee corrective actions included issuance of plant night <

order 2/6/90-1. The night order reinforced operator actions within the emergency plan implementing procedure 4701-2 page 1 , of 3. Specifically,. operators classify a postulated emergenc ' up to an " Alert" emergency classification with the low range stack. radiation monitor (RM-8132) in the event RM-8168 is out of service. Once an " Alert" classification'is implemented, communication between the control room operators and a health physics technician outside the plant occurs to consider up-grade of the event based on on-site radiation level The equivalent operational range of the monitor was not ad-dressed by_the licensee since no guidance in the TMI action plan or standard technical specification exists for a pre-planned alternate monitoring metho However, the additional actions taken af ter the " Alert" classification using communi-cation between a health physics technician and control room e p )

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l personnel will assure continued off-site dose projections and c_lassification capabilities prior to the avai1 ability of radio-

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The potential radiation hazard to acquire data during inoper-ability of the high range monitor (RM-8168) is considered in

,'    that the health physics technician's time is. minimized to acquire radiation data, and the data are taken using the-shielding of building e   To this end,-licensee actions to this item were adequate, however, administrative control or actions for long-term use in a plant night order was questioned. Licensee actions in this regard will be reviewed in future inspection .0 Maintenance / Surveillance 4.1 Observation of Maintenance Activities The inspector observed and reviewed selected portions of-preventive and corrective maintenance to verify compliance with regulations, use-of administrative and maintenance procedures, compliance with codes and standards, QA/QC involvement, use of bypass jumpers and safety tags,' personnel protection, and equipment alignment and retest. The following activities were included:
   -- AWO M2-90-6025, Replace #1 Atmospheric Dump Valve Positioner, June 10, in0
   --.AWO M2-90-6026, Replace #2 Atmospheric Dump Valve Positioner, June 13, 1990
   - -AWO M2-90-6176, Replace #1 Atmospheric Dump Positioner work under PDCR MP2-90-050, June'15, 1990
   -- AWO M2-90-6177,. Replace #2 Atmospheric Dump Positioner work under PDCR MP2-90-050, June 15, 1990 No notable observations were made, o

4.2 Observation of Surveillance Activities The inspector observed and reviewed portions of completed surveil-lance tests to. assess performance in accordance with approved pro-cedures and limiting Conditions for operation, removal and restora-tion of equipment, and deficiency review and resolutio The fol-lowing tests were reviewed:

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SP 2401G, Reactor Protection System Testing, June 13, 1990

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SP 2403A, Engineered Safety Actuation System Bistable Trip Test, June 11, 1990

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SP 2604F, High. Pressure Safety Injection Operability : Test - Facility II, June 20, 1990 '

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SP 2654, Operation of the Emergency Diesel' j Generators, June 25, 1990 No notable observations were mad i fi - 4.3 Previously Identified Items

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     - jl 4.3.1' (Closedy Violation 50-336/89-16-01: Incomplete Technical

< Sjtecification Surveillance for Engineered Safety Actuation System Automatic Actuation Logic i , This open item concerned a violation of technical specifica- I tion (TS) surveillance requirements'for;th*. testing of the 'j engineered safety actuation system automatic actuation logi .The procedure to satisfy the requirement of TS table 4.3-2 item 2c was deficient in that the surveillance did not test the automatic test inserter signal for the containment spray actuation module. Coupled with the specific deficiency des- t cribed above, the inspector questioned whether the licensee TS 7 surveillance program achieved the intent of verifying oper- . ability of all instrumentation identified'in the T Li On September 18, 1989, the licensee documented the root cause, f corrective ~ action, and actions to prevent recurrence of the violation. The licensee's documented root cause of not imple-menting the technical specification amendment in 1981 was ; personnel oversigh+, The licensee demonstrated immediate ! operability of the automatic test sequencer for the portion of automatic actuation logic in question by performance of an inservice test on July 26, 198 Licensee corrective actions-included a revision to monthly surveillance ~ procedure'SP 2403A f

to include.the automatic test. feature verification of the e k tainment spray actuation signal, The inspector verified th ! upgrade to procedure SP 2403 .t Actions to prevent recurrence included licensee review of 1 instrumentation listed in TS section 3/4.3 to verify plant j surveillance procedures test all functions specified.in the TS ' requirements. Department Instruction 1.07 established the ! method for completing the review and verification. The in- , spector reviewed the results of the licensee's review and 1 h identified no inadequacies. The inspector also verified for ; [ recently issued technical specification amendments, that the y actions' prescribed in procedure ACP-3.29, " Incorporation of d License Amendments", was accomplishe ] This item is close i l

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- 4.4 Steam Generator Atmospheric Dump Valve Anomaly Licensee review following the reactor trip on May 8, 1990, and sub-sequent cooldown identified a deficiency in operation of steam gen-erator (SG) atmospheric dump valves (ADVs) 2-MS-V-190A/B. The valves opened fully on the reactor trip full-demand signal, and the SG
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. safety / relief valves did not actuate. However, during subsequent cooldown using the. valves, secondary temperature and pressure spikes made it evident that the dump valves did not fully open. Further troubleshooting revealed that the maximum achievable stroke range was approximately 30 to 60 percent of full ope , Maintenance and I&C personnel evaluated the situation and provided an-analysis that the 20-50 psig air signal developed by the installed-Honeywell positioner was inadequate to support full stroke of the valve. The analysis showed that 61.5 psig was required to overcome various valve loads to stroke the valve the required travel of , inche The licensee. decided to upgrade the positioner _to one with a 20-75 psig output. The selected positioner is manufactured by Masonelian and was installe A plant. design change record (PDCR) evaluation completed by the I&C departmei: oncluded that a new positioner (new manufacturer) could be installed, with additional pressure gauges on the air supply lines-without any affect on the operation of the ADVs. The PDCR evaluation MP2-90-050.was reviewed by the inspector and no safety questions were
. generate .The. maintenance was accomplished in accordance with work orders M2-90-6025 and 6026. The inspector witnessed portions of the in-stallation work, No problems were noted. The retest of the new positioners was accomplished per a detailed test procedure and work orders M2-90-6176 and 6177. Testing was observed by the inspecto The test was adequately coordinated with operations personnel and the
.-plant heatup/startup activities that were in progress at the tim Testing.on MS-1908 was completed satisfactorily but valve MS-190A was not' completed prior.to plant startup. Both valves were verified to be able to fully stroke open upon a reactor trip full open demand signal. The licensee was experiencing difficulty obtaining'a stable position for MS-190A at approximately the 10-15 percent open posi-tion. Troubleshooting efforts were still in progress at the end of the-inspection perio _ _ . - . . . . . . . .
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. Assessment The inspector concluded that the problem evaluation and repair activities were satisfactorily completed for the MS-190A/B SG-atmospheric dump valves. Testing was complete and well documente No safety concerns were generated as a result of this repair activit Licensee corrective action to correct the ADV positioner problem was still in~ progress.at the end of the inspection perio .5 Calibration of Wide Range Nuclear Instrumentation and Procedura Changes (A.49.01)

Several concerns regarding calibration of the wide range nuclear instruments (WRNIs) have been identified. These concerns were docu-mented in non-conformance report.(NCR) No. 290-057 and submitted to licensee managemen The inspector reviewed these items and determined that there was no-safety significance. The licensee staff was either correcting the items or had made an evaluation which satisfactorily eliminated the concer Inspector review also eliminated one .of the items through an independent evaluation. The regulatory requirements for the WRNIs are as follows:

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Two out of four channels are required in modes 3, 4 and 5 for verification of. shutdown margi Two out of four are required for operation at the remote shut-down panel in modes 1, 2 and 3 (NRC Regulatory Guide 1.97).

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 .Two out of four are required for refueling operations for I/M plot Additionally, the WRNIs have two control functions: .(1) at 10E-4 percent power the zero mode bypass _is enabled to eliminate the thermal margin low pressure pretrip on rod withdrawal, and (2) in-dication of neutron flux at the ' control room and hot shutdown panel for use in manual contro The concern with the WRNIs was that impedance matching resistors on-the J-7 test connector for all four instruments were not as specified on drawing. 23203-28500 Sheet 1114. The licensee has subsequently processed a drawing change to account for the fact that the 75 ohm resistors specified on the-plan are actually 56 ohms for channel 1, 2, and 3 and.100 ohms for channel The initial drawing for the instruments provided 75 ohm resistors as a nominal value for the test connecto During installation testing, actual measurements deter-mined that the appropriate values for the resistors were as noted abov ,

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test circuit.attenuators affect only the Campbelling circuit which is F used for indication from approximately 10E-2 to 150 percent powe t Power _ indication from 10E-8 to 10E-2 percent power is provided by th pulse circuitry.

o " The third concern was that I&C surveillance procedure SP 2401J

 " Thermal- Mrgin/ Low Pressure Calculation- Tests" had a procedure change lthat was implemented that significantly changed.the intent but was evaluAttd as a non-intent change, The change in question, chang _to revisioi 10 of SP 2401), implemented a_slightly different method
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L for adjusting the linear power drawer subchannel test potentiometers i when~ delta-T power is less ; nan 12 percent. This change assured that "

,  the delta-T pewer was' greater than nuclear power. This was essential to correctly accomplish the test. The change did not affect the out-come of the surveillance and therefore met the non-intent provisions '
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of ACP 3.02E. Both intent and non-intent changes are reviewed by PORC, with nonintent changes requiring a post-implementation review within 14 day Based on the above, the inspector determined that licensee actions in this matter were proper and no inadequacies were identified. This matter is close .0 Engineering / Technical Support 5.1- Azimuthal Power Tilt Indication Description'

Azimuthal power tilt (Tq) is defined as _ the maximum dif ference between power generated in a reactor core quadrant and the average o power of all quadrants in that half of the reactor core, divided by the average power generated in all quadrants in that . half of the ' Cor * Previous to cycle 10 operation, Tq values- averaged 0.3 to 0.5 per-g cen Tq increased during Cycle 10 to the. range of 0.7 percent to " l'.2 percent. On May-7, the licensee completed a root cause investi-gation (documented in NE-90-R-228) for the higher than normal azi- '

muthal power tilt values for cycle 10 operation. Licensee Actions w

The licensee: concluded that the probable root cause for higher than normal Tq values was core quadrant reactivity differences resulting from small asymmetries in the core flow and temperature distribu-tions. The licensee's evaluation included consideration of the fol-lowing: review of tilt angle and core location during cycle length; incore nuclear instrumentation performance; past tilt performance;

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reactor coolant system (RCS) temperature and flow gradient differ-ences within the reactor core; and', differences in the number of' tubes plugged in-each steam generator The tilt angle essentially

*  remained constant at 180 to 200 d% en F (RCS hot leg penetration).

- The licensee noted that loop f Sw e 'f w ences, loop cold. leg tempera - ture differences, and steam genera wr - # plug difference correlated directly to the tilt ragnitude cN 9ghs o "

    'rection. The tilt indication appears axially homogenout. Mr.;hout the core based on
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in-core nuclear detector data, laso w tov review identified no cor-

 -relation between the core locat%ns af suspect control element as-semblies (report detail 5.2.1) and core. radial til As 'a result of the above review of core tilt fluctuations, the licensee requested the fuel vendor (Advanced Nuclear Fuels) to review the matter for impact on the safety analysis, and to recommend-actions to minimize core tilt for the beginning of Cycle 11 opera-tio Requirements for Azimuthal power Tilt Technical Specification 3.2.5. limits Tq to less than 2 percen Actions required,for tilt in excess of 2 percent are to increase surveillance for azimuthal . power tilt and total integrated radial
 - peaking f actor (FrT). Should ti't exceed 10 percent, subsequent power operation is limited to less than 20 percent-allowed thermal power, and an investigation into the cause is required. The bases for.the Tq and FrT requirements are to ensure that assumptions used in the accident analysis for linear heat rate and local power density remain valid for control element assembly insertion limits, and to preserve margins to the applicable limiting safety system settings (i.e., thermal margin / low pressure setpoint).

Safety Assessment On June 7, 1990, the fuel vendor provided its documented evaluation which showed that azimutha power tilts less than 2 percent do not impact the safety analysis as long as the limitations on linear heat rate and total integrated radial peaking factors are met. Tq values less than 2 percent have no significant impact on core reactivity coefficients (i.e. control rod worth). The inspector verified by review of completed surveillances that linear heat rates and total integrated radial peaking factors were within required limit Conclusion Based on reviews of supporting data and documents, and discussions with cognizant reactor engineering personnel, the inspectnr identi-fied no inadequacies in licensee actions and conclusion Good identification, root cause review, and followup actions were noted for core parameter change . . . . . _ . _ . _ _ _ _

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i . r 0 5.2 -Engineering Support of Plant Operations . 5.2.1: Control Element Assembly Failure Followup On June 7, 1990, while conducting cold functional testing of f the control element assemblies (CEAs) at the Maine Yankee -{ P Nuclear iv e Station, one CEA could not be fully inserted 'l into the core. Subsequent inspection of the CEA revealed that

,   the'end cap was missing from the center CEA finger, the lower stainless steel spacer and boron carbide pellets had fallen 1   out of the center finger, and an axial crack existed at the lower end. The upper stainless steel spacer was cocked in the bottom of the CEA finger and was causing the' finger to bind.in !

the cuide tube, thus, resulting in lack of full insertion of-

,   the t,E .

The CEA design susceptible to this fc.ilure has a single center finger filled with boron carbide (B4C) pellets to the lower

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tip, with a spring-load sho k absor'oer at the top. The-re l maining four cor_ner fingers are filled with B4C pellets except : for the.last eight inches, which are composed of silver- ; > indium-cadmium to prevent swelling at the lower ti .j i The improved CEA design has the center finger composed of ; silver-indium-cadmium for the last eight inches of the CEA l finger, j Safety Significance .

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L The failure to insert the CEA based on center finger failure

,  results in potential multiple stuck CEAs, localized criti- =

I t; cality after a reactor trip, and invalidation of safety analy-sis initial conditions and results as described in Final

Safety Analysis Report (FSAR) sections 14.1.2.4 and 1 CEA . operability is required in part by technical specifications

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3.1.3.1, 3.1.3.3, 3.1.3.4, 3.1.3.5, 3.1.3.6'and 5.3.2. .These 1 requirements verify position indication, CEA drop times, the 'l e number of CEAs in the core, and position agreement between CEA ! groups.

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Licensee Actions I

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(cm ~ On June 15, the resident presented questions to the licensee q

regarding CEAs of the design susceptible to failure at j' Millstone 2. The topics addressed included: past CEA per- " formance history and measurements; number and age of affected . CEAs; operational tests conducted on CEAs; and, potential j detection mechanisms of incipient failures.

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V4' The licensee.has a total of 29 CEA$ of the susceptible design within the core. Of this population 13 are scheduled for ;

replacement during the September,1990, refuel outage. The j remaining 16 were installed in the core in the 1985 time frame ' p ' and are currently not scheCuled for replacement, i

.c   In March 1985 the licensee commenced a nondestructive examin-  t i   ation' program of CEA fingers to develop a basis for CEA re .  !

f placement based on the potentia) B4C swelling mechanism. The ' i examination program in part was implemented to validate the' 'i

design lifetime of CEAs, which is 10 years per FSAR 3.3. .

The nondestructive examinati.ons included eddy current testing _

.'. ofgall original CEAs (73) and profilometry of 9 CEA fingers in  :

March, 1985 to determine finger strain and. defects. The maxi- ! mum wear of any finger was 2.8 mils (finger diameter 0.948 1 inch) and the worst observed strain was 0.99 percent in the center finger of CEA'#33. CEA #33 was discharged from the ' core in April, 1985. In February 1987 the licensee conducted ! eddy current testing of 21 discharged CEAs. No cracks were ' , note On June' 22, '1990, as a result of the CEA failure at Maine ! Yankee, the licensee visually inspected (by video-recording) t 12 CEAs in the spent fuel pool that were discharged from the im

core in February 1989. No defects were observed. The CEAs inspected had exposure of 9.08 effective full power year . The licensee concluded, based on the result of_the previous examinations and present evaluations, that continued operation i until the' refueling outage is acceptabl On July 6,1990, the NRC formally requested additional infor-mation from Northeast Utilities and a commitment to an action plan to address CEAs at Millstone 2. This item is subject to l further review during futtre NRC inspections.- ' 5.3. Steam Generator Tube Repair Activities The inspector observed portions of the licensee activities to inspect, test and repair steam generator (SG) tube plug leaks identi-fied during the present outage that commenced on May 8, 1990. As of ; June 4, 1990, the licensee had completed eddy current testing and SG pressure testing which identified four leaking tube plugs of ' various types noted in NRC Region I Inspection Report 50-336/90-0 On June 5, the inspector witnessed the completion of OP 2316C Revi-sion 7, Change 1, "SG Leak Testing Using Auxiliary Feedwater and High [ Pressure Nitrogen."- The licensee was only able to pressurize the SG to approximately 150 psig due to the limited capacity of the

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nitrogen supply system. 'Two additional welded plug leaks were identified immediately adjacent to a plug that had just been weld i

. repaired. The licensee engineering staff concluded that welding-  !

activities had most.likely caused the additional-leak ;

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The. licensee subsequently completed weld repairs to these additional plug. leaks'and accomplished a pressure test on June 8 that resulted - in no: leak identification. This test per OP 2316C was also 6ccom- * plished at a reduced pressure 'of appro.ximately 150 psi * The' inspector discussed several aspects of the repair activities with l' licensee engineering personne Cleanliness control was maintained

through the entire repair evolution. The contractor personnel accom- - i plishing the weld repair activities completed cleaning and inspection ;

of the SG bottom head for loose material such as metal fragments, ; prior to removal.of equipment. Licensee personnel also completed a cleanliness inspection when scaffolding was removed. An additional-check by licensee QC personnel was also accomplishe , The inspector also discussed with licensee engineering personnel the :

possible correlation between the size and number of leaks observed during the inspection as compared to the leak estimates made from radiation monitor readings during operation prior to shutdow .

The licensee had no formal calculation, but had made a reasonable estimate of the types of leaks that were observed at the low test pressure and concluded that these leaks could reasonable have been

. expected to cause-the estimated leaks that were observed during power operation. -The pressure and temperature differentials are much
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greater when operating at power, versus the lower temperature and low- 1 pressures used during leak testing activities. The inspector con-cluded that these estimates were reasonable attempts to quantity the ' SG tube leakag The inspector witnessed various portions of the SG tube leak testing per OP 23160. The licensee personnel controlled the evolution in an

. adequate manner. The differential pressure between the two SGs was  .

maintained at less than 5 psig as required by procedure. No safety ' concerns were generated as a result of the inspectio ,

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5. /89-24-06: Licensee Event 6

, , ,   R(Closed)UnresolvedItem eport 89-011-00: Service _ Water Valve Reportability
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j , M eliness c This unresolved item was opened to determine the licensee's t e '

corrective actions to improve reportability determinations and NRC notification. timeliness, as a result of the. events in LER . 89-011-0 '

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Northeast Utilities Service Company (NUSCO) procedure NEO 2.25 - r " Identification and Implementation of NRC Reporting Require-ments" provides guidance on timeliness and completion of re- ' i portability determinations. The licensing department provides

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e the status of substantial safety hazards and reportability i evaluations bimonthly for the four NU nuclear plants. The licensee provides this status report to the NRC residents at ; Connecticut Yankee and Millston Based on the above actions, licensee corrective actions to I improve reportability timeliness is acceptable and this item is closed; however, licensee reportability actions will be

reviewed as part of routine resident inspection .! s 5.5 Review of 10-CFR 50.59 Determinations for Plant Modifications The licensee's annual report for January 1,1989, to December 31, !

+  1989, for Millstone, Unit 2, was reviewed. The report identified 24 !

plant design changes, 34 plant design change evaluations, 14 pro-cedure changes, 22 jumper, lifted-lead, bypass changes, 11 set point-changes, and 6 tests. As required by 10 CFR 50.59, the report con- !

  .tained, for each change except one, a brief description of the
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' change, including a summary of the safety evaluation.. Plant Design-Change Record,MP-2-030-88, Replacement of Service Water Piping, was identified but no summary was included. This item was presented to licensee. management for action. The licensee concluded for each

v change that the modification did not constitute an unreviewed safety __ question per the criteria of 10 CFR 50.5 i Document NSAC/125, June 1989, Guidelines for 10 CFR 50.59 Safety Evaluations, prepared by the Nuclear Management and Resources Council of the Nuclear Safety Analysis Center and operated by the Electric Power Research Institute was used as inspector guidance to review the

. content of safety evaluation NSAC/125 is currently not endorsed by the NRC, and the licensee has not committed to this document as a reference in the established administrative control procedures for safety evaluations. This document indicated, for the purposes of '

performing safety evaluations, that the determination of an un-

  ' reviewed safety question can be broken down into seven separate
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questions as follows: (1) may the proposed activity increase the probability of occurrence of an accident previously evaluated in the SAR7; (2) may the proposed activity increase the consequences of an accident previously evaluated in ,the SAR?;. (3) may the proposed activity increata the probability of occurrence of a malfuretion of equipment important to safety previously evaluated in the SAR7; (4) may the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the SAR?;

(5) may the proposed activity create the possibility of an accident of a different type than any previously evaluated in the SAR7; (6)

may the proposed activity create the possibility of a different type of malfunction of equipment important to safety than any previously evaluated in the SAR?; and, (7) does the proposed activity reduce the margin of safety as defined in the basis for any technical specifi-cation? Each of the safety evaluations were reviewed for consistency with these guidelines. A sample of 3 plant design change reports and 9 plant design change evaluations were reviewed in depth to determine if acceptable determinations were performe PDCR MP-2-034-87, Reactor Coolant System Vent Upgrade This PDCR concerned the installation of manual isolation valves downstream of the reactor head and pressurizer remotely operated vent valves. Two safety evaluations (SE) were provided, one developed by plant engineering and another developed by the reactor engineering branc The corporate reactor engineering branch SE was more com-prehensive and provided justification for each of the seven questions except question 6. "May the proposed activity create the possibility of a different type of malfunctio of equipment important to safety than any previously evaluated in the SAR7" The SE by plant engineer-ing provided a good description of the change and the reasons for the change, and it drew the proper conclusion PDCP MP-2-Oll-88, Secondary Side Safety Relief Valve (SRV) Position Indication This PDCR concerned instrumentation required to monitor the safety relief valve positions (16 total valves). The SE drew the proper conclusions except for answering question 6. As in the case for PDCR MP-2-034-88, no explanation was provided to answer question 6. Al-though justification for answers to each of the other six questions were provided, they were not clearly discussed on a one-for-one basi PDCR MP-2-030-88, Replacement of Service Water Piping This PDCR concerned the replacement of portions of the service water system with upgraded or equivalent materials. Carbon steel epoxy-lined pipe was replaced with carbon steel PVC-lined pipe and the

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Stainless steel spool pieces on the discharge side of. all three ; j T, teactor building closed cooling water (RBCCW) heat exchangers were ! l replaced with carbon steel PVC lined pipe. The SE drew proper con- ' F clusions on questions 1, 2, 5 and 7, but it did not address questions ' r' 3, 4, and 6. It concluded that the modification will not produce an unreviewed safety' question. This conclusion was drawn. because the- 'a change was an. improvement in protecting the piping from corrosio ;The modification did not change the configuration of the-piping sys- - L te * l PDCE MP-2-89-012, Terniination of Containment Air Recirculation Fan A : w -This PDCE evaluated the removal of an existing electrical connection > L box on the "A" containment air recirculation (CAR) fan motor and l F > replacing it-with a new connection box in order to_ provide the-fan g ' an environmentally qualified electrical connection to the motor. The t safety evaluation provided a good description and reason for the change, and it made the proper conclusion on the basis.that the ., change.does not change the. design functional requirements of the CAR l' cooling system and.the operation of the CAR fan moto PDCE MP-2-89-020, Containment Air Recirculation Fan Outlet Duct

 ' Expansion Joint     ,

This PDCE evaluated the replacement'of the fan outlet duct asbestu ! type expansion joint to a non-asbestos-material joint. The SE pro- i vided a good description, the reason for the changes and concluded ,

that the change does not constitute an'unreviewed safety question.- PDCE MP-2-89-023, Reactor Building Closed Co9 ling Water Base Plate Modifications ,

This PDCE evaluated the modifications of R2CCW heat exchangers sup-port assemblies to correct the effects of-corrosion to the existing base plates and anchor bolts. The SE provided a good description of -

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the modification and concluded that the change does not ennstitute an , unreviewed safety question with appropriate justification for each of the. conclusion PDCE MP-2-89-024, Modification of Pressurizer Safety Valve Support . This PDCE evaluated the change involving increasing the bolt hole-diameter of the pressurizer safety valve inlet support plates and the . use of hardened steel washers with the valve bolting. The SE pro-vided a restatement of the questions in a negative sense and thus concluded the change is not an unreviewed safety questio y R L ,1 ..-

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L . r ' PDCE MP-2-89-030, Pressurizer Proportional Heater Remover from .l Operation Dise to Low Insulation Resistance i

This PDCE evaluated the removal of the electrical energy from a pressurizer proportional. heater. The SE provided a good description ; of the modification and concluded that the change does not constitute .

U an unreviewed safety question with appropriate justification for each of the conclusion ' ly ;

 .PDCE fD-2-89-035, Power Supply X-1113A/B Replacemen E This PDCE evaluated the replacement of the original equipment 125 VDC power supplies for the Auto Auxiliary Feedwater Initiation System ,

and the Anticipated Transient without Scram System with Acopian Model ,

 #A120HT350.- The SE concluded that the change does not constitute an l'

unreviewed safety question with good justification for each conclu-k, ' sion except question 7. Although it concluded that the margin of safety of the Technical Specifications would not be reduced, the SE '

:  did not support the conclusion in the discussion. It must be implied . ,

i that, because there was a one-for-one replacement with equipment of !

;  equal or better quality, there would be no reduction in margin of L  safet !
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PDCE MP-2-89-070, Fire Pump Pressure Switch Isolation ' This change installed an isolation valve in the instrument supply piping to the fire pump pressure control switch. The SE concluded ,

that.the change does not constitute an unreviewed safety question, i It addressed affects on accidents which provided justification for " questions 1 and 2. It based conclusions relating to questions 3 ': through 7 on a general discussion of the effect of the implementation , of the change on the fire suppression water system by removing it t from service in approximately 4 hours which is less than 7 days as ' allowed by the Technical Specifications. The justification,to sup- ' port the conclusions had been implied but not fully develope ; PDCE MP-2-89-083, Battery Charger Power Failure Relay This change concerned replacement of a Furnas relay with.a similar General Electric relay in the 'B battery charger. The relay is-a power failure rela The SE concluded that the change does not constitute an unreviewed safety question and provided for the proper conclusions based on-the fact that the' component is a one-for-one ' , e electrical and seismic replacemen ! PDCE MP-2-89-114, ' Reactor Coolant Pump (RCP) Conduit Improvements ! This change involved the replacement of 1-1/2 inch EMT conduit that

extended into the motor cavity on all four reactor coolant pumps with 1-1/2 inch rigid conduit with appropriate seismic supports. The .

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conduit is for the RCP speed sensor circuit. This was an improvement .

 .inthat it provided seismic qualified' supports for the RCP speed i L
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,  sensor circuit. The SE concluded that the change does not constitute i
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an' unreviewed safety question and justified the response to the 7 ,

,  questions based on the improvement in the seismic qualification of- i 9 .
 'the speed sensor circuit. It fully developed the conclusion relating to effects of. analyzed accidents and the margin of safety. The justi- fication to support the responses relating to the other questions was '
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implie . Conclusions on Review of 10 CFR 50.59 Determinations j Based on" review of PDCR MP2-034-87, PDCR MP-2-011-88, PDCR MP-2-030-88, i F PDCE.MP-2-89-035, and PDCE MP-2-89-070, the inspector concluded that q safety evaluations documented did not record a thorough justification i for each criterion of NSAC/125 to fully ',upport the conclusion that ; the change did not constitute an unreviewed safety question. Th '

licensee, however, is not committed to NSAC/125 in established 80 min-istrative control procedures for safety evaluations. Notwithstend- ! ing, the licensee has a good 10 CFR 50.59 determination process. The ;

 -inspector concluded that there is reasonable assurance that the changes i meet.the criteria of 10 CFR 50.5 :
 ' 6.0 Security l

Selected aspects of site security were verified to be proper during in- ! spection tours, including site access controls, personnel-searches, per- r sonnel monitoring, placement of physical barriers, compensatory measures, , guard force, staffing, and response to alarms and degraded conditions. No r notable observations were mad ~ r 7.0 Safety Assessment /0uality Verification 7.1 Licensee Self-Assessment Activities The inspector reviewed the results of selected audits conducted'by the I licensee quality services department and the nuclear review boar : The review considered the. purpose of the audit, findings of the l audit, licensee response and corrective actions, verification of t corrective actions, and assessment of overall audit attributes.

, Specific audits reviewed were: i

  - A22019, ' Verification of Millstone 2 Technical Specification bnplementation'   i
  - A30141, ' Material Equipment and Parts List'
  - A30144, 'NUSCo Assessment of Equipment

. Environmental Qualification (EEQ) Implementation' r:

 - The inspector concluded the audits were comprehensive and had generally good findings.

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7.2 Periodic Reports y .

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 . Upon receipt, periodic reports submitted pursuant to. technical specifications.were. reviewed. This review verified that the reported i is

' information was valid and included the required NRC data. The in-

 'spector also ascertained whether any reported information should be  -

classified as'an abnormal ociurrence, The following reports were reviewed: r

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Monthly Operating Report - May 1990 ,

b , No notable observations were mad , 7.3 ' Licensee Event Reports >

L Licensee event reports (LERs) submitted during the period were re- : viewed to access LER accuracy, the adequacy f corrective actions and compliance with 10 CFR 50.73 -eporting regt aments, and to determine if there'were any generic implications or ' any further information i was require The following LERs were reviewed: i

LER 90-06, " Manual Reactor Trip": L< [ . Previous NRC review of the adequacy of r,orrective actions and generic implications concerning LER 90-06 were documented in inspection 3 l report 50-336/90-09 sections 2.3 and 4.1.1. No inadequacies were , noted regarding 10 CFR 50.73 reporting requirement ,

 .LER 90-05, " Potential High Energy Line Break via Auxiliary $ team

, , Lines Located in $afety ReTated Areas": Licensee immediate corrective actions were to isolate-auxiliary steam ! toLthe unit. Inspector verification of auxiliary steam isolation was ' accomplished by review of tag order 2-648-90 and verification with control drawings 26026 sheets 1, 2, and r

Licensee corrective actions to develop plant raodifications to restore auxiliary steam system is ongoing, as well as a review of the impact ! '

 !on equipment required for safe shutdown following a high energy line  .

break. This matter will be 11 owed in future inspections by review l of the supplemental report t, :ER 90-05, expected to be completed by ;

*  the licensee on December 31, : 390.

NRC review of the reporting requirements of 10 CFR 50.73 identified I no inadequacie LER 85-002-01. " Unplanned Actuation of Containment Purge Valve":, On May 30, 1990, the licensee documented additional corrective actions for a previous LER (85-002) dated April 1,1985. The :

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corrective actions were necessary to prevent inadvertent actuation of-(+ an engineered safety actuation system signal, specifically, the con-tainment purge valve isolation (CPVI). The 1985 actuation occurred . L as a result of holding the radiation monitor reset. button too lon .The result was tnat the monitor achieved the " failed" indication > ! resulting in an actuation of CPV j The corrective actions included: installation of covers and a warning sign over the reset button for the associated radiation monitors to

. prevent inadvertent use of the reset button; reinstruction of opera - :

P tors that a CPVI signal'may be processed if the reset button is held too long; and~ issuance.of a generic troubleshooting procedur~ The inspector verified.the licensee corrective actions, No inadequacies were noted. The, inspector further reviewed the reporting require- .i ments and identified no inadequacies,, 'l LER 90-002-00, " Failure to log into Techncal Specification Action Statement while performing Calibration of RM 8132 A&' ( A.44.01): On' April 20, 1990, the licensee reported to the NRC under 10CFR 50.73 (a)(2)(1)(B) a condition prohibited by the technical specification ; that occurred when the control room operators failed to enter into 1-technical specification 3.3.3.10, Table 3,3-13 item la. and le. for RM 8132A/B " Millstone 2 stack radiation monitor." ' RM 8132A/B was released for calibration (procedure SP 2404AF) on March 20, 1990. The calibration actually began on March 21 and the operator recognized the error on March 22 at 12:40 p.m.- The actions not. performed as required included chemistry grab sampl.s every 12 . hours and estimation of sample flowrate every 4 hour ! RM 8132A/B is required to be operable on a continuous basis except for outages up to 12 hours for. surveillance (i.e. calibrations). The operability requirements of the monitrr is based on 10CFR 50 Appendix A, design criteria 60, 63 and 6 Licensee evaluation concluded that during the time the monitor was 1 out of service (March 21 - March 22), no-increase in gaseous or-particulate radiation levels within the auxiliary building occurre The inspector verified the radiation levels were acceptable in the-- auxiliary building by review of RM-8434A/B and 8997 "Radwaste Vent

Monitor," RM-9095 " Waste Gas Monitor," RM 8145A/B " Fuel Handling Exhaust" and RM-5099 " Steam Jet Air Ejector Monitor" outputs. The

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inspector also noted no decrease in the waste gas decay tank leve ' The inspector verified licensee corrective actions as documented in LER 90-002, and noted that reportability and timeliness requirements per 10CFR 50.73 (a)(2)(1)(B) and NUREG 1022 were me No inadequa- ' cies were identifie .

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Conclusion- ! L : Failure to adhere to the. limiting condition for operation (3.3.3.10 ~ table 3.3-10) within the technical specification is a violation.- In accordance with 10CFR 2 Appendix'C, subpart G.1, a notice of viola- : tion is not being issued.since the five criteria for exercising en- 'l 6 forcement discretion were fulfilled by the licensee. This non-cited ;

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  -violationisclosed(50-336/90-11-02).  *
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7.4 Reportable Events

7.4.1' High Energy Line Break Analysis Assumption Errors ! L', On June 21 at 4:15 p.m., the licensee reported under 10 CFR i , 50.72(b)(1)(ii)(a and b) an unanalyzed condition that was ! L outside the design basis and which significantly compromised -i L plant safety. The unanalyzed condition was documented in PIR '

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90-55.- During a plant walkdown, the licensee identified two ' , " access door conditions that would invalidate the assumptions ! in the HELB analysis. Specifically, the north access door- ,

into the emergency 4160 kilovolt room located at .the 54 foot elevation of the turbine building, and the access door on the - north side of the. D.C switchgear room between the auxiliary . i bui.lding hallway and the switchgear room could not withstand a , design pressure of 0.75 psig during a mainsteam line brea ,w -Thus, equipment in the switchgear rooms could be affected during a postulated main steam line break even '

Licensee cor.rective actions included: tagging open the' 1 y turbine roll up door upon identification of the condition to preclude pressurization, and the installation ~of bypass . Jumpers 2-90-35 and 2-90-3 Bypass jumpers 2-90-35 and' ! 2-90-36 installed a strong back and reinforcement plat respectfully on the affected access doors. The-inspector' reviewed:the bypass jumper support documentation, verified strong back installation and tagging, and reviewed the ! engineering calculations supporting the temporary modification "

to satisfy design parameters. .No inadequacies were note . Partial Engineered Safety Feature Actuation

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   'On June 17, the licensee reported an actuation of an engineered safety feature per 10 CFR 50-72(b)(2)(ii). =The ,

actuation was documented in PIR 90-48. At approximately 1:18 a.m., a partial actuation of the "B" train of the enclosure building filtration actuation system (EBFAS) inadvertertly initiated the enclosure building filtration fan (F-25B), the control room filter fan (F-328), and opened the associated fan discharge dampers (2-EB-41 and HV-210). The control room operators immediate actions were to: open the suction

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verify whether any other ESF. equipment was in operation. No i other ESF equipment inadvertently started. The operators I ,j l restored the ventilation system to its normalistandby align-men ,

At the. time of the actuation, an operator was replacing an EDG sequence 2 indicator light.on ESF actuation cabinet "B." The . EBFAS-fans and discharge damper are aligned to sequence 3 and ; ~The inspector verified, based on discussions with th ; licensee and review of diagram S9N21-6 sheets 1 and 2, that no

relationship exists between the indicator light and the EBFAS logi At the end of the inspection period, licensee investigation into the cause of the partial actuation was ongoing. The inspector will review licensee actions in future inspection . CFR 21 Notif' cation on Main Steam Safety Valves ; i On May 17, Northeast Utilities was informed by Dresser Industries, Inc., of a variation in the "K" factor used to

establish main steam safety. valve setpoint using the 1566 hydraulic setting device (hydroset). This concern was docu- ', mented in PIR 90-4 The hydroset reduces spring pressure

without raising main steam line pressure, allowing the safety valve to lif In-the hydroset hydraulic pressure measure-ments, a valve K factor constant is used to determine the set pressure of the valv ' The licensee verifies main steam safety valve setpoints with l procedure MP2730B in accordance with technical specification ., requirement 4.7.1.1. -The acceptance criteria (+/- 1 percent)

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is consistent with the specifications of ASME;section XI, , subsection IWV-3510, 1974 edition, t On May 30,-the station director initiated an internal action item in response to the Dresser-Industries, Inc. 10 CFR 21 report. On June 11, Millstone 2 initiated procedure change 1 - to MP 2730B-1 and nonconformance' report-(NCR) 290-058 to re-flect the valve K fact r changes recommended by the vendo The licensee evaluated the most recently completed valve sur-- veillance data and recalculated the set pressures with the revised K factors for the main steam safety valve Five out of sixteen safety valves exceeded the acceptance criteria, with a worst case setpoint of +1.7 percent of set pressure. On June 11, with the plant in hot standby, the licensee conducted an acceptable safety valve test on the five valves exceeding the desired setpoin , j ,

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The' inspector verified the procedure change, reviewed the ' W revised setpoint calculation implementation, and reviewed the f' , completed surveillance and NCR disposition. Based on previous l , NRC review,.in inspectien report 50-336/89-03 section 5.2 and LER 89-02-00, the current out-of-specification safety valves

setpoints did not invalidate the design functions to mitigate main steam pressure for overpressure events, or to limit off-site doses for postulated steam generator tube rupture event , Assessment

Licensee actions in' response to the vendor 10 'CFR 21 report ! were acceptable, timely and sufficient.-

7.4.4 Missed Grab Sample on RM-8132A/B l 7 ' , On June 20,'the licensee documented a failure to take a chemistry grab sample every 12 hours as required per technical specification 3.3.3.10 Table 3.3-13 item 1.a. , This matter !

'f   was documented in pIR 90-53. A preliminary licensee evalua- i tion concluded this was a condition prohibited by the techni- '

cal specification and a licensee event report will be issued, g

Inspector review noted a previous similar event as documented ;

,   in LER.90-02, and licensee actions will be tracked as an un- "

r resolved item to review the adequacy of controls for process radiation monitors (50-336/90-11-03).

7.5 previously Identified Items -;

7.5;1 (Closed) Unresolved Item 50-336/89-01-02: No Mechanism ; for Coordinating a Quality Assurance Overview of Safety- - JelatedActivities !

This item questioned the adequacy _of quality assurance (QA)

,   overview of safety-related activitie In addition to the coordination concern, the timely responses to the QA sur-veillance were also questione [
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 ,  The statiun quality assessment services supervisor presented
  'to the inspector his 1990 integrated assessment plan as_evi- .

dence that the licensee is integrating and coordinating the quality services department (QSD) verification activitie The coordination effort was supported by the plant quality services manager and supervisor. The plan describes the methods for performing the assessment function, i.e., audits, surveillances, and inspections; the ma.jor activities to be assessed; and includes a description of how the assessment activities will be accomplished during 1990, i t 4 .

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The inspector reviewed.the revised ACP-QA 9.070, " Quality

    ' Services' Surveillance Program,". that-tightened the response requirements regarding surveillance findings. The automated  ;

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surveillance = tracking: log was also reviewed. Only one overdue ; o item was listed for the station as_ compared to an August 16, '

1989,. report that indicated _ that Millstone 2 had 15 overdue ' ' i tems '.

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{}r'f'     Ba' sed upon these findings, the item is closed. . -

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8.0 Management Meeting "1 ' i Periodic meetings were held with station management to discuss inspection ., ; 4 , findings'during the: inspection period. -A. summary of findings was also _ ! discussed at the conclusion of the: inspection. 'No proprietary informatio Li

  - was covered within the scope of the inspection. No written material was Y";-  ' j;  given to the licensee during the inspection perio ;
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v t e Inspection Report - Millstone - 1990011
80 \| 81 \| 82 \| 83 \| 84 \| 85 \| 86 \| 87 \| 88 \| 89 \| 90 \| 91 \| 92 \| 93 \| 94 \| 95 \| 96 \| 97 \| 98 \| 99 00 \| 01 \| 02 \| 03 \| 04 \| 05 \| 06 \| 07 \| 08 \| 09 \| 10 \| 11 \| 12 \| 13 \| 14 \| 15 \| 16 \| 17 \| 18 \| 19 \| 20 \| 21 \| 22 \| 23 \| 24 \|
1990 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 82(0) 83(0) 7(0) 8(0) 9(0) 10(0) 11(0) 13(0) 14(0) 16(0) 19(0) 23(0) 80(0) 8(0) 9(0) 10(0) 11(0) 18(0) 19(0) 27(0) 27(0) 8(0) 10(0) 12(0) 19(0) 80(0) Security Operator Exams Emergency Planning Other

IR 05000336/1990011

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