IR 05000336/1988002

Insp Rept 50-336/88-02 on 880101-0208.No Violations Noted. Major Areas Inspected:Previously Identified Items,Plant Operations,Surveillance,Radiation Protection,Physical Security,Ie Bulletin Followup & Outage Activities

ML20196J603
Person / Time
Site:	Millstone
Issue date:	02/25/1988
From:	Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20196J597	List: IR 05000336/1988002 ML20196J592
References
50-336-88-02, 50-336-88-2, IEB-87-002, IEB-87-2, NUDOCS 8803140413
Download: ML20196J603 (20)

Text

.

.

. .

,

.

U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report: 50-336/88-02 Docket No: 50-336 License No: DPR-65 Licensee: Northeast Nuclear Energy Company P. O. Box 270 Hartford, CT 06101-0270 Facility: Millstone Nuclear Power Station, Unit 2 Inspection at: Waterford, Connecticut Dates: January 1, 1988 through February 8, 1988 Inspectors: William J. Raymond, Senior Resident Inspector Peter J. Habighorst, Resident Inspector John T. Shediosky, Senior Resident Inspector, Haddam Neck Lynn M. Kolonauski, Resident Inspector, Millstone 1 Reporting Inspector: William J. Raymond Approved: & O bO, b :ll2s /gr E. C. McCabe, Chief, Reactor Projects Section 18 Date Summary:

Areas Inspected: Routine, unannounced day and backshift inspection (140.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />)

of: previously identified items, plant operations, surveillance, radiation protec-tion, physical security, IE bulletin followup, loss of normal power events, Engi-neered Safety Feature (ESF) sump screen divider installation, shift technical advisor staffing, and outage activitie Results: No violations or conditions adverse to safe plant operation were identi-fied. Additional follow-up is warranted on loss of normal power events (Section 8.3), shift technical advisor staffing (Section 7.0), and control room in-leakage surveillance results (Section 3.0).

8803140413 880308 PDR ADOCK 05000336 Q DCD

.

.

.

.

.

TABLE OF CONTENTS Page 1.0 Persons Contacted.................................................... I 2.0 Summary of Facility Activities.... ............. . . . . . . . . . . . . . . . . . . . . 1 3.0 Licensee Actions on Previously Identi fied Items. . . . . . . . . . . . . . . . . . . . . 1 4.0 Physical Security.................................................... 2 4.1 Devitalization of a Vital Area.................................. 2 4.2 Unescorted Contractor in Vi ta l Area . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 4.3 Unescorted Visitor in Protected Area............................ 3 5.0 IE Bulletin 87-02, Fastener Testing to Determine Conformance with Applicable Material Specifications - TI 2500/26...................... 4 6.0 Engineered Safety Feature (ESF) Containment Sump Screen Divider. . . . . . 5 7.0 Shift Technical Advisor Staffing..................................... 6 8.0 Outage Activities.................................................... 7 Refueling..... ......... ....................................... 8

-

8.2 Radiation Controls.............................................. 10 Loss of Normal Power Events..................................... 11 8.4 Failure of the Engineering Safety Feature (ESF) Integrated Tes .5 Core Load Verification.......................................... 16 8.6 ES F a c t u a t i o n o n 1/2 6 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 Surveillance.................................. ...................... 17 10.0 Committee Activities................................................. 18 11.0 Management Meetings.................................................. 18

!

i i

l l

l

1

_ _ _ ____ _ ____- -____ _ _ _ _ _ _ _ _ _ _ _ _ _

l

.- . ___

.

.

"

.

.

DETAILS 1.0 Persons Contacted .

Mr. S. Scace, Station Superintendent  !

Mr. J. Keenan, Unit 2 Superintendent Mr. F. Dacimo, Unit 2 Engineering <

Mr. D. Kross, Unit 2 Instrument and Control Mr. J. Riley, Unit 2 Maintenance Mr. J. Smith, Unit 2 Operations Mr. R. Kacich, Manager, General Facility Licensing Other members of the Operations, Radiation Protection, Chemistry, In.strument >

and Control, Maintenance, Reactor Engineering, and Security Departments were also contacte .0 Summary of Facility Activities Millstone Unit 2 was shutdown throughout the inspection period for the scheduled cycle 9 refueling outage. The major outage activities wer Cycle 9 refueling, consisting of sixty-eight (68) new Westinghouse batch t

"L" fuel assemblies, five twice burned Batch "G" assemblies, and instal- l lation of twenty-one (21) new Control Element Assemblies (CEAs) '

--

Steam generator eddy current testing (ECT) examinations

--

Engineered Safety Feature (ESF) Integrated Test

--

Containment Integrated Leak Rate Test (CILRT)

--

Human Factor Modifications in the Control Room

,

--

"D" Reactor Coolant Pump electrical testing, and motor replacement

--

Plant secondary system inspections and maintenance on the low pressure turbine, steam generator feedwater pump inspections, and extraction steam pipe replacement

--

Inspection, repair, and replacement of Service Water System piping 3.0 Licensee Actions on Previously Identified Items l l

3.1 (0 pen) Unresolved Item 87-25-05: Evaluation of Control Room Radiation Levels. The licensee tested the leak tightness of the control room en- i velope. Testing was conducted per inservice test (IST) T88-05 to verify i that control room inleakage was less than 100 cfm using a test method that made the control room pressure negative by 1/16 inch of water rela-tive to the outside areas. The inspector witnessed test activities be-tween January 28 - February 4, 198 !

l

< _ ._ _ - _ . ._ __ . _ _ _ _ . _ _ _- -

_ _ _ _ _ _ _. _ i

.- . . ..

.

.

. .

. 2 Preliminary test results could not verify the 100 cfm acceptance cri-terion was met. The initial results indicated inleakage was as high as 600 cfm. Af ter repairs were made to the control room boundary, measured control room inleakage was 350 cfm at 1/16 inch negative pressure. How-ever, the licensee concluded that control room inleakage for the control room ventilation system configuration that would exist under actual accident conditions (ventilation system in recirculation and no negative pressure in the control room) would be less than that indicated by the test results, and actions were initiated to evaluate the validity of the test method. Since the chosen test method is specified in Technical Specification 4.7.6.1.e.3, and since satisfactory demonstration of the ,

100 cfm inleakage is required to satisfy the control room ventilation system Limiting Condition for Operation (LCO), the licensee ini,tiated an emergency technical specification change request to allow measurement of the control room inleakage using alternate test method Satisfactory completion of the surveillance requirement to demonstrate compliance with the LC0 is required to support plant entry into Mode Inspector verification that the licensee me.t the LC0 occurred after this inspection and will be documented in a subsequent repor .0 physical Security

,

Selected aspects of site security were verified to be proper during tours, including site access controls, personnel and vehicle searches, personnel monitoring, placement of physical barriers, compensatory measures, guard force staffing, and response to alarms and degraded conditions. No inadequacies were identified. The following item warranted inspector followu .1 Devitalization of a Vital Area At the request of the Unit 2 Operations department on January 4, security devitalized a specific vital area (VA) to allow unrestricted access by maintenance personnel. Almost 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> later, on January 5, an opera-tions Shift Supervisor conducting a plant tour discovered that the security department had allowed the devitalization of the adjacent VA as well as the intended one. The two VAs are separated by a fire door without access controls. Access controls to the adjacent VA were im- ,

mediately restored. Subsequent VA tours conducted by both the security and operations departments revealed no indications of intrusion or damage to the safety-related equipment in the VA. In any case, because the plant was in Mode 5 during this event, Technical Specifications did not require the associated safety-related equipment to be operable. The licensee made a one hour report to the NRC Emergency Operations Center i per 10 CFR 73.71c; the event is also the subject of Safeguards Event Report (SER) 88-01. The resident inspector reviewed this SER and found no inadequacies. The inspector determined that this event had no security significanc i

__ - _ _ ,- - . - - - - - - - - - - - -

. _ _ _ . _ _ _

.

.

. .

. 3 The inspector had a concern about operations department involvement in the devitalization of VAs. Discussions with security personnel revealed that, other than the initial request and weekly tracking of the area, the present procedures do not require the operations department to review devitalization of an area. To prevent recurrence, the licensee briefed all security personnel on the event and will modify the security proce-dure to require concurrent review of the devitalization by the Operations and Security Shift Supervisors prior to access control remova The in-spector will verify the adequacy of the revised procedure upon issuanc .2 Unescorted Contractor in Vital Area The licensee informed the inspector of a problem involving the , loss of escort control of a visitor. The licensee identified the problem when an individual previously escorted into a vital area (VA) tried to exit it at 11:00 a.m. on 1/15. A craft worker (contractor) was escorted into the VA to perform work in support of the refueling outage. After entry into the VA, the assigned escort turned over escort responsibilities to another worker in the area. In the turnove.r process, the assigned escort failed to properly identify the craft worker as requiring an escort in the VA. The contractor attempted to exit the VA without escort and was immediately contacted by security personnel. When questioned by security, the contractor stated that he had recently completed security training, but was unaware of the need for an escort inside VAs. The event was reported to the NRC via the ENS (Emergency Notification System) at 11:45 a.m., 1/15. This example of a failure to control visitors in-accordance with security procedure will not be cited as a violation since it was identified by the licensee, it was report, and corrective actions were prompt and appropriate to prevent recurrenc The licensee's corrective actions were to upgrade the affected indivi-dual's security training. The inspector reviewed the licensee's safety, security, and emergency training lesson plans administered to contractor employee This review noted that the training addressed VA escorts and the escorted person's responsibility if separated from an escort. The inspector reviewed the exam administered after completion of the training lesson and found it adequate. The inspector had no further questions on this matte .3 Unescorted Visitor in the protected Area The licensee reported that a visitor and his escort became separated at 4:50 p.m. on January 20. The event was reported to the NRC via the EN The separation occurred within the protected area. Upon separation, the visitor reported to the North Access Point (NAP) at 5:03 p.m. to notify security personnel of separation from his assigned escort. The escort reported to NAP at 5:10 p.m. to notify security personnel of separation from the assigned visitor. The licensee interviewed both individuals and concluded this was an accidental separation with proper corrective

_ m . . . _ _ _ - _ -

.

.

. .

,

-

actions. The inspector had no further questions in regard to this matte This example of a failure to control visitors in accordance with security procedure will not be cited as a violation since it was identified by the licensee, was reported, and corrective actions were prompt and ap- -

propriate to prevent recurrenc .0 IE Bulletin 87-02, Fastener Testing to Determine Conformance with Applicable Material Specifications - TI 2500/26 The NRC issued IE Bulletin 87-02, Fastener Testing to Determine Conformance with Applicable Material Specifications, dated November 6, 1987,.to request l licensees to review their receipt inspection requirements for fasteners, and to determine through independent testing whether fasteners in stock meet re-quired mechanical and chemical specifications. Item 2 of the~ bulletin speci-fied that the licensee draw a sample of fasteners from safety-related and non-safety-related stores, with the participation of the onsite resident in-specto NRC Inspection Reports 50-245/87-33 and 50-336/87-29 document resi-dent inspectar review of licensee actions to withdraw fastener samples from station stores and submit them to an offsite laboratory for analysi In-spector review of licensee actions to withdraw duplicate sets of safety grade and non-safety fasteners and nuts for all three Millstone units, for a total Millstone station sample size of 80 items, verified conformance with Bulletin ,

items 2 and The licensee responded to IE Bulletin 87-02 by letter dated January 12, 198 The inspector reviewed the response and found that it addressed the informa- :

tion requested by Bulletin items 1, 4, 5 and 6 concerning receipt inspection practices, controls for storage and use of fasteners in safety-related and non-safety-related applications, results of chemical and physical testing performed, and the need for additional actions based on the test result The licensee's response addressed the test results for 160 fastener specimens,

which included samples drawn from stores at the Connecticut Yankee facilit The inspector reviewed Purchase Order 864178 dated December 9, 198 It was submitted to an independent laboratory, J. Dirats and Co. , to request testing in accordance with the bulletin. The inspector verified that the purchase order instructed the laboratory to complete testing per Bulletin Item 4 in accordance with the specification grade and class applicable to each fastener, .

including verification of ultimate strength, hardness and chemical properties as required by the appropriate specifications. The inspector also reviewed the sample data sheets included in the licensee's January 12 response, along !

with the test results provided by the laboratory. Review of the purchase  !

e order and test results verified that the description of the tested material matched the sample descriptions recorded by the NRC inspector as each item was withdrawn from stores and tagged on December 7, 198 i The licensee's test program identiffed 7 discrepancies in the 160 samples i

'

tested. Of the 7, two were judged to be nonconforming. Nonconformance re-ports were issued to disposition these items. The licensee concluded that both of the nonconforming items were minor specification deviations which are

,

. . . . .- .

.

'

.

not safety significant. The nonconformances involved Millstone samples drawn from QA and non-QA stores: MP-21a/B, ASTM A307 Gr B 5/8 inch bolt; and MP-3A/B, ASTM A193-78A Gr B 3/8 inch bolt. In both samples, the test results ,

showed that the material deviated from the ASTM Specificatio i The licensee determined that all samples met their specified functional re- >

quirements and could be used "as is." Specifically, the chemical composition for one bolt showed chromium outside the specified range of 18.00% to 20.00%

at 17.29%, and nickel outside the specified range of 8.00% to 10.5% at 10.8%.

The out of specification chemistry was accepted by the licensee because the variations were not enough to significantly degrade the alloy's strength, ductility or corrosion resistance. Also, the measured Rockwell B hardness for a bolt was in excess of the specified maximum of 95 at 95.5. This out of specification hardness was accepted by the licensee because the reading indicated above average strength with no significant effect on ductility and corrosion properties, and becausc the reading was within the ASTM E 10 meas-urement accuracy for property of +/-2%. '

Based on the above results, the licensee concluded that no additional actions relative to the fasteners in stock were warranted. No further actions are planned. The inspector identified no inadequacies in the licensee's conclu-sions or plan '

i The inspector had no further questions on this item. It will be reviewed further following NRC Staff review of the licensee's response to Bulletin 87-02 and evaluation of the reported test results. Further review per TI

-

2500/26 item 05.01 and 05.02 regarding receipt inspection program procedures and implementation will be conducted on a subsequent routine inspection.

'

6.0 Engineered Safety Feature (ESF) Containnent Sump Screen Divider >

On January 20, the licensee notified the inspector of a deviation from the '

Final Safety Analysis Report (FSAR), section 6.2.2. This section commits the i licensee to an assembly dividing the containment sump screens located in the i lowest elevation of containment. The screen assembly was not divided into  ;

, two independent sections separated by mesh and grating as depicted in FSAR  ;

1 figure 6.2- The inspector independently verified that this condition ex-I iste The containment sump is protected from clogging by the sump screen.

l The screen is sized for low velocity flow [i.e. less than 1 foot per second (fps)]. The screen is sized such that any particle that penetrates the screen '

is capable of passing through ESF components. Due to the low velocity across the sump screen, loose objects generated by an in-containment pipe break will ,

either settle on the sump screen or will not affect sump operation. Loose i

objects most likely to contact or pass through the sump screen would be pieces i of pipe insulation dislodged by a loss of coolant incident (LOCI). ,

The inspector asked when the non-conformance concerning the sump suction i screen divider occurred. The licensee stated this condition had existed since initial construction of the facility but had not been previously identifie '

i l

i

- - - - _ - - - -- - - - . - . , ..- = -

-. _- .. .

.

-

.

.

.

The licensee generated work order M2-88-01529, "Installation of Containment Sump Grating," on January 29. Divider installation was completed, with the mesh / grating attachment and spacing installed consistent with the grating-to-frame joints of the existing sump screen assembly. The inspector verified

,

this by review of procedure M2-88-0152 The inspector reviewed the following procedures used to install the contain-ment sump screen dividers:

--

M2-88-01529 "Installation of Containment Sump Grating"

--

Commercial Commodity Evaluation - Deck grating hold down clips

'

--

Commercial Commodity Evaluation - 1 1/4" Deck Grating

'

,

--

Inspection Status Evaluation (Quality Control)

--

Millstone Visual Weld Inspection Plan

--

WPS-605, Rev. 2 "AWS Prequalified Joint Wel' ding Procedure for Joining Structural Carben Steel" )

--

QA-1303, Rev. 4 "Visual Examination Procedure" ,

t

--

Material Issue Form #020188 "Category I Wire Mesh Cloth" i

,

!

--

Material Issue Form #012588 "Category I Angle Iron" l The inspector reviewed the Quality Control (QC) inspections of the containment sump screen divider installation, done in accordance with Procedure QA 1303, #

and had no further question .0 Shift Technical Advisor Staffing '

On January 21, the Office of Nuclear Reactor Regulation (NRR) informed the inspector of concerns involving the use of dual-role Shift Supervisor / Shift !

Technical Advisors (SS/STA) at Millstone and Haddam Neck. Millstone 1 Tech-nical Specification (TS) Table'6.2-1, "Minimum Shift Crew Composition," lists a total of seven individuals for Modes 1, 2, and 3. Because the SS serves i

in a dual capacity (SS plus STA), these seven positions are filled by six individuals. The current Millstone 1 TSs (original issue, license DPR-21,

.

l

dated October 31,1986) do not specify the dual role or otherwise contain a l reference acknowledging that the STA role may be filled by the on-shift S i

-

The inspectors discussed this issue in meetings with Millstone station man-  ;

i agement on January 21. The issue was subsequently expanded to include Mill- l stone 2 and Haddam Neck. (The Millstone 3 TSs allow a dual role SR0/STA but i

,

also require an on-shift SRO to meet the current NRC policy statement on  ;

engineering expertise on-shift.)

(

i I

!,

t f

E I

__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . - - - - _ . -

. . . - . . - . . - ..

.

'

.

7 r

.

In response to an October 4,1983 licensee request to amend the TS for both Millstone 1 and 2, NRR processed License Amendment 95/92 by two letters dated

! February 16, 1984. One letter would have issued a version of TS Table 6.2-1 I

which recognized the dual role SS/STA. However, the associated transmittal

, letter stated that, since the Commission policy statement on engineering ex-pertise on shift was still in draft form, the proposed TS for the dual-role -

l SS/STA was deemed unacceptable at the time. Consistent with this considera-

'

tion, a revised version of TS Table 6.2-1 was issued without reference to the dual-role SS/STA, with the associated transmittal letter stating that the issue of an on-shift SRO filling the STA position would be handled separately, i On January 1,1934, Millstone 1 began using a dual-role SS/STA without change '

,

to the TS, after having informed the NRC of their plans to do s In an unrelated licensing action, the NRC staff issued Amendment 104'/102 for MP-1/2 on August 6, 1985 to approve an organizational change as found on page 6-3 of the TS. By mistake, a copy of page 6-4 which contained the version of TS Table 6.2-1 which allowed the dual role SS/STA was included in the ,

transmittal. The licensee recognized the mistake and did not distribute this !

version of Table 6.2-1 in their copies of the TS >

On January 22, 1988, the licensee's Manager of General Facility Licensing met -

j

"

with the inspectors to discuss STA staffing. Also, a meeting between the NRC -

staff and the licensee was held in Bethesda, Maryland on January 27, 1988 to :

discuss the issue. The NRC asked the licensee to develop a plan to achieve :

compliance with the applicable TS. The staff stated that there was no com- {

pelling safety reason to take immediate action prior to Commission review and ,

approval of a license amendment. After subsequent communication with the ;

j staff NU prepared a proposed TS amendment which would allow the SS/ STAS to !

continue to serve in the dual role without additions to current shift staffing,

l 4 but with future dual role SRO/STA personnel to meet the current Commission l policy statement on engineering expertise on shift. (The proposed amendment, !

dated February 17, 1988 was sent to NRC subsequent to the inspection period.)

J

, The inspector will continue to follow the progress of the newly proposed l

! amendment and other developments involving the dual role SS/STA issue.

) 8.0 Outage Activities i The inspector reviewed the conduct of the Millstone Unit 2 Refueling /Mainten- i i ance outage through direct inspection of activities in the containment, ;

auxiliary building, turbine building intake structure and control room. Por- '

tions of the following activities were reviewe Installation of Steam Generator Nozzle Dam Repositioning of Control Element Assemblies in the reacto Fuel movement in the reactor, reactor cavity, and fuel storage pool.

1

_ _ _ _ _ . _ _ _ _ . _ _ - - _ _ _ _ _

i

'

.

--

Stea 'enerator Tube Eddy Current Testing (ECT).

--

..emoval of a broken drill bit and the associated tube plug from the N Steam Generator tube shee Inspection, repair and replacement of "A" Service Wate'. System pipin "D" Reactor Coolant Pump Motor Electrical Testin Plant Secondary System Inspections and Maintenaree includin *

Low Pressure Turbine inspections;

Steam Generator Feedwater Pump Turbine inspections; and

• Extraction Steam Pipe replacemen The inspector observed that there has been a significant improvement in the performance of the steam generator nozzle dam system during this outage. The installation of these devices was much more trouble free than in the pas The inspector determined that this performance i.mprovement was the direct result of meaningful preoperational testing, additional personnel training on the Steam Generator Mock-up, and direct managemer.t attentio The inspector also reviewed the program for assuring quality in the analysis of Steam Generator Tube Eddy Current Testing (ECT). A new program not only makes use of experienced analytic personnel but also provides, first, a training program and, second, a qualification examination. The intended re-sult is that the personnel involved in ECT examinations be familiar with unique and specific ECT signals of the Millstone Unit 2 Steam Generator Tubes, and that they have a clear understanding of acceptance criteri The inspections and repair of the "A" Service Water System Piping were re-viewed. Implementation of this program required that sections of the piping be removed to allow for inspection of its inside surface. The inspector found that a high level of attention was being placed on this program. Because of the piping location, system function, and its past performance, this activity is very important to overall plant safet There were no unacceptable conditions identifie The inspector reviewed control roem activities on 4:00 p.m. on January 23, 6:30 p.m. on January 30, and 11:30 p.m. on February 10, to observe activities in progress. Shift staffing and control room protocol were proper. No inade-quacies were identifie .1 Refueling The inspector reviewed the following procedures in regard to the refuel-ing activities:

1). SP-21025 Boron Requirements for Spent Fuel

_ . _ . __ _ _ - _

.

t

.

'

9  ;

11). SP-21064 Refuelirg Machine Load Test iii). PDCR 2-47-87 Cycle 9 Reload iv). EN-21008 Refueling Worklist Administrative Control v). OP-2209A Refueling Operations vi). OP-2614A Daily Technical Specification (TS) Surveillances during Refueling During refueling operations the inspector periodically witnessed opera-  !

tions in the control room, spent fuel pool areas, and refueling, cavity area. Reactor engineers, plant operators, and non-licensed operators were interviewed during refueling operations to determine responsibili-ties, duties, and knowledge level. No inadequacies were note .

In witnessing control room operations, the inspector verified periodically that a minimum of two source range neutron. flux detectors were operating !

with continuous visual indication in the control room as required per i TS 3. The refueling machine was independently checked by the in- ;

spector for load cut-off limit Scrveillance was completed within the i allowable time frame (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) prior to moving fuel assemblie ;

,

The inspector verified proper communications between the control room and fuel handling areas had been established. While communications were lost on January 21, fuel movement was halted. The control room was notified during transfers of fuel assemblies, fuel assembly positioning over the core, unlatching of fuel assemblies in the core, and placement /

removal of fuel assemblies in the upender. The inspector witnessed fuel assembly L-65 movement from the upender into the transfer canal on Janu-ary 21. The licensee properly completed Materials Transfer Form (ENG Form 21001-1) step-by-step. The movement of fuel assembly L-65 was timely and in accordance with approved procedures.

, According to PDCR 2-47-87, Rev. O, cycle 9 reload consisted of 68 new

Westinghouse Batch L fuel assemblies, five twice-burned Westinghouse l Batch G assemblies, four once-burned Westinghouse Batch H assemblies, and 21 new Control Element Assemblies (CEAs) to replace 21 old CEAs.

i The only fuel design change between the Batch L (new fuel) and the pre-vious Batch K fuel is a change in the fuel pellet design. The old design was a square end, dished pellet; the new design has a shorter length (0.457 inch vs. 0.600 inch), a chamfered edge, and reduced dish volum The chamfered edge and the reduced pellet length are expected to result

.

in improved fuel performance through reducing pellet-clad interactio The inspector reviewed the safety evaluation for cycle 9 reload. The

! safety evaluation was divided into (i) impact on design basis accident (ii) potential for creation of a new unanalyzed event, and (iii) impact i

I l

.. _-- _ _ _ . . . - - - . . - _ . - . - . . . . . -

r

.

.

l

.

'

. 10 on the margin of safety. The review was comprehensive. The inspector had no further questions. The following items were periodically verified ,

during the refueling to assure licensee compliance with the TS: *

Shutdown cooling loop operability TS 4.9. '

Refueling Water Storage Tank Temperature TS 4.1.2. Spent Fuel Pool area ventilation TS 4.9.14 ;

Engineered Safeguard Actuation System TS 4.3.2. i Offsite breaker alignment AC Sources TS 4.8. ;

Beron Concentration (RCS) TS 4.9.1. !

The inspector identified no outstar. ding issues in reg ed to refuelin In summary, the inspector concluded that the cycle 9 refueling was well manage .2 Radiation Controls i The inspector reviewed the licensee's program on posting of radiological control points, radiation work permits (RWPs), and the health physics i technicians' duties and responsibilities. Also, the inspector physically exercised access control into and out of posted "High Radiation Areas", ,

radiologically controlled areas, and contaminated areas during this in- '

spection period. The inspector observed adherenta of the licensee to approved procedures involving access into radiologically centrolled crea L Control of high radietion areas in accordance evith technical specifica-tion (TS) 6.12 and Procedure SHP 4906 (Posting of Radiological Controlled l

-

Areas) was implemented an1 well coordinate The inspector interviewed personnel at the health physics (HP) control >

point on use of the containment elevator and the possibility of uncon- l trolled access into high radiation areas. That elevator has an elec- .

trical interlock to prevent operation at the -24 foo+, elevation (a posted [

high radiation area). This interlock safeguards against uncontrolled access to a high radiation area. The licensee informed the inspector that no past events concerning uncontrolled access to the -24 foot level

,

+

of containment have occurre l The inspector interviewed high radiation area guards to assess their re- )

sponsibilities. Responsibilities include controlling access to high i'

radiation areas, insuring personnel have the proper dosimetcy and appro-priate protective clothing, and communicating with the health physics control poin Thn inspector had no further questions in this are '

The inspector reviewed SHP 4912 (Radiation Work Permit Completion and Flow Control, Rev. 10) and verified implementation at the health physics control point. This procedure provides guidance for development of RWPs designed to specify radiation safety and control requirements in areas of possible significant exposure, and their utilization as an exposure recording system. Radiation Work Permits are required for high radiation l

i

!

,

[

_____..---,,,,--...,.-----r -

. _ - -- - .. . _ . ___ . - -

.

- .

.

. 11 ,

areas, contaminated areas, neutron radiation areas, airborne radiation areas, and maintenance or inspection of contaminated or radioactive equipment in excess of the following limits;

--

Removable contaiination in excess of 1000 disintegrations per minute

. per 100 square centimeters (OPM/100 CM2) of beta gamma and/or 20 DPM/100 CM2 alph Radiation dose rates in excess of 100 mrem /hr or where a whole body exposure of 100 mrem could be received within one work wee ;

, The inspector reviewed radiation und contamination levels within con-tainment and verified proper postings and the implementation of,RWPs.

, The inspector had no further questions in regards to this matte l The inspector toured the containment on January 23, 1988 to review-acti-

'

vities in progress and to-verify compliance with radiological controls

-

as established by RWPs. The inspector verified that workers wore re-quired dosimetry and that protective clothi.ng required by the RWPs was i

, in use. Workers were cognizant of dose rates in assigned work areas and

of Icw dose area

, for Leeping exposures as low as possible. The inspec- .

.

tor verified the posting of barricades and augmented controls established t for work in high dose areas was proper.

The inspector verified for the steam generator inspections in progress

at 4:30 p.m. on January 23 that: health physics (HP) personnel at the

control points for the steam generator work were knowledgeable of duties and administrative controls; worker monitoring in high dose areas was proper; and HP personnel were cognizant of activities in progress. HP .

personnel received periodic briefings on work activities for the shift i and on anticipated changes in plant operational status. The licensee had established good control of worker exposure for work activities in- ,

j side the containment. No inadequacies were identifie ;

8.3 Loss of Nortral Power Events i 8. Event 1: Inverter Failure i At 9:20 a.m., on January 19, 1988, with the unit shutdown, the l j licensee reported vital bus 24D was de-energized by an under- e

!

voltage function of the Engineered Safety Actuation System j J (ESAS). The ESAS undervoltage signal was caused by a failure  ;

t of DC/AC inverter no. 1. Each of the four ESAS cabinets (A, '

B, C, and D) are powered from four vital DC-AC inverters which  ;

are supplied from two vital 125 VOC battery buses. Each bat-

tery bus supplies two respective inverters. Each inverter  !

I supplies 120 VAC vital instrument buses (VA), which in turn l

supply the four ESAS cabinets.

.

I I

.

l

_ -_- . _ . . _ , _ , _ . . _ . . _ _ _ . - . - .. - _ _ _ _ _ _ . . . . . _ , , . . _ . . _ . _ . . , , . _ _ . . - - . - . . _ _ _ . . - , _ . _ . - - . . . ,

-

.

. .

.

j The failure of inverter no.1, and the unavailability of its l non-vital backup supply (inverter no. 5) due to scheduled +

maintenance, resulted in a loss of VA-1 Instrument bus VA-30 i was out of service at the time due to scheduled maintenanc I This particular loss of 2 out of 4 vital instrument buses re-  !

suited in an ESAS undervoltage and subsequent start and load  :

sequencing of the "B" emergency diesel generator (EDG). The r

"A" EDG was not available because of scheduled shutdown main-tenanc The inspector reviewed licensee drawings 25283-28150(Engi-neered Safety Logic Sequencer and Channel output), 25203-30024-(single line diagram 125 VOC emergency and 120 VAC vital in-strumentation),and 25203-39047 sheet 11 to verify the licen-

.

!

see's root cause determination. The inspector verified minimum required AC electrical buses and DC vital buses were operable i at the time of the event in accordance with Technical Specifi- l cations 3.8.2.2 and 3.8.2.4. No inadequacies were note The failure of inverter no. I was caused by a short across DC capacitor C1. The licensee determined the failure of C1 in inverter 1 by work order M2-88-0102 The vendor (Mepco Cen-tral Lab) informed the licensee that possible age degradation r causing dielectric breakdown may have caused the failure of capacitor C1. In response, the licensee developed a DC capaci- p tor in-service. test (T-88-20) for vital and non-vital inverters  !

-

to detect capacitor dielectric breakdown. The objective of this test is to apply working voltage across DC capacitors used 1 in vital and non-vital inverters to determine the quality of  :

dielectric material for all DC spare capacitors prior to in- l stallation. The test determines if the capacitor will hold "

a charge and will operate under a working voltage, and verifies manufacturer's capacitance values. The inspector observed  :

Plant Operations Review Committee (PORC) meeting 2-86-33 con- ,

cerning approval of in-service test procedure T-88-20. Licen- '

see review during the PORC meeting was thorough and addressed the safety evaluation and environmental impact of in-service *

test T88-2 ;

On January 22 and 27, the in-service test was performed on all spare DC capacitors used in vital and non vital inverters associated with ESAS. All capacitors passed the acceptance criterion of holding a constant voltage potential for fifteen -

minutes after the initial charging time of one minute, and met the capacitance static test value as prescribed on the capaci-to Tested capacitors have been installed in all vital and i non-vital inverters associated with ESAS. Further, the licen- *

see is considering a program to periodically replace DC capacitors associated with vital and non-vital inverters. The >

inspector had no further questions on this matter, ,

!

,

I

. ,- , _ _ _ _ _ _ . _ _ _ . _ _ _ _ . _ . _ _ _ _ . . _ _ _d

. .- - . .. .. . - ...

-

.

. i

'

'

, 13

.

8. Event 2: 4.16 KV Tie-Breaker Trip On January 30, at 3:47 p.m., with the unit shutdown for re-fueling, the licensee reported vital bus 24 0 was de-energized by the opening of 4.16 KV tie-breaker 52-A410. The tie-breaker supplied power to the 24D bus from the facility I normal sta-I tion service transformer. The "B" EDG did not start due to

.

inoperability because of scheduled maintenance. Vital bus 240 and the affected loads were restored within about one minute

,

in accordance with the licensee's operating procedure The licensee's investigation into the opening of breaker 52-A410 revealed abnormal operation of ESAS Facility 21 and Z2 cabinets. The ESAS cabinets process undervoltage actuation signals from the vital buses (240, 24C) to sequence and load the-EDG to the affected vital bu The investigation deter- '

mined the followin ESAS Facility Z2 was inopera.ble at the time of the even The Facility Z2 cabinet normally has two voltage supplies, 24 VOC and 15 VDC, for each of the respective actuation modules. Improper lead reconnection while performing work package M2-88-0176 introduced 125 VOC into the Facility Z2 cabinet. Its 24 VDC supply fuses blew, and 24 VOC and 15 VDC were lost to Facility Z2. This did not, however, cause the opening of tie-breaker 52-A410. The licensee responded to the Facility Z2 event by performance of in-service test T-88-21 to identify the damage. Details *

concerning in-service test T-88-21 and restoration of the j Facility Z2 cabinet are discussed in Detail At approximately 3

30 p.m. on 1/30, the licensee reported ,

a failed +15 VDC power supply (PS-503) to the undervoltage actuation module in ESAS Facility Z1. The power supply c

was replaced, and successfully retested by the licensee 3 on 1/30 under work order M2-88-01613. This event also

did not cause the opening of tie-breaker 52-A41 ] The inspector concurred that the Facility Z2 and Z1 problems

>

did not cause the tie-breaker opening after reviewing work  !

orders M2-88-01613, M2-88-01613, and M2-88-01682, and inter-

<

views with the licensee. The licensee has not identified the  !

root cause for the inadvertent opening of 4.16 KV tie-breaker

52-A410. However, successful 4.16 KV breaker interface with l

,

Facility 21 and 22 ESAS cabinets was verified during the inte- l grated ESF test on February 5. The inspector interviewed lic-

! enses personnel, independently verified results of the inte-I

! grated ESF test, and concluded that breaker operability has i been adequately demonstrated despite the inability to determine  !

;

'

t

.,  !

'

,

I F

,w - - . . e-g - , - , r-- - _ -.

-_------m----. . - . , . -._g g,-,,,,-av_s.g m a -,a ,.y ,,,,,m,- <,,,-wm,-nm, -

. - - . . . - . _ =

_ - ..

.

.

-

. 14 ,

!

the breaker opening cause. The inspector will continue to  !

review surveillance and preventive maintenance on the breaker during future routine inspections, d

8. Event 3: 4.16 KV Bus 24C Loss At 8:54 p.m., on 2/4, the unit was shutdown (Mode 5) when a loss of normal power to 4.16KV vital bus 24C occurred. The 1 cause of the event was improper placement of 24A/C breaker 52-

A304 in test in preparation for the integrated ESF (SP-2613C)

tes That caused reserve station service transformer (RSST)

supply breaker 52-A302 to vital bus 24C to open because the control circuit reflected both breakers as being closed. On loss of vital bus-24C, the "A" EDG started on undervoltage, however, output breaker 52-A312 did not initially close because of the placement of breaker 52-A304 in test. In this configur- ,

ation, the control circuit for the "A" EDG output breaker had '

"closed" contacts for both the normal station service trans-1 former (NSST), and RSST supply br.eakers to vital bus 24C, re-

sulting in "A" EDG inability to assume loads on the 24C vitai  !

bus.

,

The operators, on verification of this condition, manually shut the "A" EDG output breaker and restored vital loads within three minutes of the start of the event. However, load se-quencing was not as specified. The licensee's corrective ac- >

' tions included replacement of the "A" EDG automatic load se-  !

quencer in ESAS Facility Z1 cabinet per work order M2-88-01887, t

Also, change No. 3 to ESF integrated test SP-2613C was issuad, cautioning operators on the effect upon mechanical coupler de-vices when placing breaker 52-A-304 in test. The inspector had no further questions in this are .4 Failure of the Engineered Safety Feature (ESF) Integrated Test  ;

On February 5, the licensee performed integrated ESF test SP-2613C, Re , Change 4. The objectives of the procedure are to demonstrate proper  :

de-energization and load shed of emergency buses 24C, 240, and 24E, pro-per diesel start and energizing of emergency buses and permanently con-nected loads, and determination that the emergency diesel generators (EDGs) will sustain tripping of large load ,

,

The inspector discussed with the licensee two specific changes to SP-i 2613C, The first change prescribed ambient conditions for the EDG. Am-bient conditions as defined by the licensee are: EDG secured for at 1 cast

'

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, or at EDG ambient temperature. The reason for this change was to meet the intent of having the EDG at ambient conditions without tying satisfaction of the initial condition to a specific number of hour ,

-

The licensee based ambient conditions (i.e. crankcase temperature, lube

!

'

i

'

i E

l r

. _ _ _ _ _ , . . _ _ . , . . . - _ _

.

.

.

oil temperature, EDG room temperature) on an EDG which had been secured for greater than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The revision was discussed and approved at Plant Operations Review Committee (PORC) meeting 2-88-55 on February The second change to SP-2613C developed a prerequisite concerning a Mode 5 operation option. The prior procedure SP-2613C, prerequisite was:

--

Mode 6 with the following conditions: (i) The reactor vessel head removed and the refueling cavity ready for filling, or (ii) the reactor vessel head is detensione The change allows the license to perform SP-2613C with the plant in Mode 5 with the following conditions: (i) High Pressure Safety Injection (HPSI)

system header stop valves 2-SI-654 and 2-SI-656 closed and safety-tagged to shift supervisor control; (ii) one pressurizer safety valve or power-operated relief valve (PORV) removed and its flange safety-tagged to shift supervisor control; (iii) The steam generator (SG) nozzle dams removed and the primary manways installe The inspector independently verified that these conditions were in con-formance with Technical Specification 3.4.2.1, 3.4.3, and 3.5.3. con-cerning plant mode changes. The inspector interviewed licensee personnel on this change and had no further question The licensee reviewed and formally approved this change to SP-2613C in PORC meeting 2-88-48, on February .

The inspector reviewed the results of SP-2613C and discussed the results with the licensee. The unacceptable results of SP-2613C were as follows:

--

Initial start of Facility I EDG Sequence 1 actuation times for ser- ,

vice water pumps A and B, HPSI pumps A and B, and Containment Air Recirculation Fans A and C varied between 0.179 and 0.249 seconds outside the required 2.2 second actuation time (based on initial closure of the Facility 1 EDG output breaker). The inspector noted that, on restart of the Facility 1 EDG, all sequencer 1 actuation times were within limit On initial start and subsequent restart of facility II EDG, Sequence l 2 actuation time for the "C" charging pump was in excess of the j allowable 8.4 seconds. In these starts of the EDG, the "C" charging ,

pump started 0.350 and 0.450 seconds in excess of the limit, based on the computer alarm and event recorde The licensee stated that failures like this had not previously occurre l The inspector expressed concern to the licensee on the potential cause ;

for the failure, and if the failures to SP-2613C were interrelated to i the previous Engineering Safety Actuation System problems. The inspector I I

.

.

.

will pursue this issue further in future inspections. At the closure of the inspection period, the licensee had not considered SP-2613C suc-cessfully completed, and investigation was continuing on the results of the integrated ESF test. (SP-2613C was successfully completed after the inspection period.)

8.5 Core Load Verification After completion of fuel loading, the licensee videotaped the top of the reactor fuel to verify the core loading agreed with the desired pattern as specified in the cycle analysis report. The inspector reviewed OP-2209A, Refueling Operations, discussed licensee activities with the Reactor Engineer on January 22, and checked the final core configuration against the desired pattern depicted on Form 2-RE-A4, Cycle 9 Core Ma The inspector's review included consideration of: confirmation of proper fuel assembly serial numbers, fuel bundle orientation, loading of neutron startup sources, fuel bundle and control element seating, and inspection of the top of the core for debris. Also, the inspector verified that core load verification by the utility included independent review of the loading pattern and involvement by QC personnel. No inadequacies were identifie The inspector reviewed the quality of the core loading videotape and noted it was adequate. The clarity of the fuel assembly serial numbers ,

on most of the tape was very good. However, the serial numbers for fuel '

assemblies at core locations N-15, J-20 and J-3 were not clear and, for

-

location N-15, the verification that the assembly loaded there was L-15 relied upon confirmation that assembly L-13 was loaded in location R- The clarity of the videotape for locations N-15, J-20 and J-3 could have been improved by optimizing the position of the underwater lighting relative to the videocamera as the taping was performed. This item was discussed with the Reactor Engineer, who noted the inspector's comment The inspector had no further comments regarding core load verificatio While verifying the core load pattern and videotaping the core on January 22, 1988, the licensee identified a small spring lying on top of fuel assembly J28 in core location S18. The inspector viewed the spring on the core videotape. It was approximately 1 inch long and 1/8 inch in diameter. The licensee retrieved the spring using long-handled tools, and measured contact dose rates of 300 R/hr. The licensee reviewed the object and determined that it was not a fuel pin plenum spring, and that it did not resemble any known component in the fuel or internals. The licensee concluded the spring was debris dropped into the system at some previous time. The licensee plans no further followup actions. The inspector reviewed the licensee's actions and identified no inadequacies.

>

,

b m , _ - _ -

--

.

.

.

8.6 ESF Actuation At 10:37 a.m., 1/26, the licensee reported under 10 CFR 50.73(A)(2)(iv)

that an ESF actuation occurred on Facility Z2 ESAS. The actuation was the Auxiliary Exhaust Actuation Signal. This AEAS trip resulted in stoppage of the spent fuel outside air supply and diversion of the spent fuel area exhaust to the Enclosure Building Filtration System (EBFS),

and in a control room ventilation shift to the recirculation mode. The cause of the AEAS was improper verification of test equipment used to test ESAS cabinet voltages during preventive maintenance (procedure IC-24308, Rev. 3). The preventive maintenance procedure objective is to verify ESAS system power supply voltages (+24 VDC, and +15 VDC) are pro-per. The technician performing this procedure did not properly, verify the isolation transformer associated with the oscilloscope test equipment floated above ground (ungrounded). The subsequent ground on the isola-tion transformer during testing of the +15 VDC on PS 603, Facility II ESAS cabinet, usulted in a noise spike, causing a trip of Facility II ESA The AEAS actuation functioned as required. Normal line-up was restored af ter identification of the cause. The inspector had no further questions on this matte The licensee discussed with the inspector the reportability of this even The inspector utilized NUREG 1022 Supplement 1 (Licensee Event Report- i System) as guidance for reportability of an ESF actuation under 10 CFR ,

50.73(a)(2)(iv). The AEAS signal was spurious in nature, however, the  !

system actuated and performed its intended function, thus, the actuation is reportable even if the system is not required to be operational. The licensee concurred with this explanation, and reported the ESF actuation per 10 CFR 50.73(a)(2)(iv). The inspector has no further question .0 Surveillanco ,

On 2/3, the licensee performed in-service test T-88-21, Rev.0, Change 2 for

'

.

'

the Engineering Safety Actuation System (ESAS) Facility Z2 actuation cabinet.

'

The ESAS actuation cabinets detect accident conditions and initiate safety features which are designed to localize, control, mitigate, and terminate such '

incidents. In-service test T-88-21 was developed to document the operation

of the ESAS Facility 22 actuation cabinet following corrective maintenance

during the refueling outage and subsequent anomalies during cabinet actuatio .

! The test accomplished the following.

i i) Visual check of each actuation module within the ESAS Facility Z2 actu-j ation cabine ) Electrical checks on circuit card diodes which may have been damaged

during a previously identified overvoltage conditio i

.

i iii) Installation of each actuation module in the ESAS module test assembly and bench test of module operability.

I i

$

) i

g .y or c w w-- -- - - -#_ - .-w p.-m - ---s-- -._..,_..-,s 9 -,,. -3, y. .g_.-m m..m,-w .%,y ,.g-,asmus.-%p4.t-2---M T9'=*W--e.e ( .

'

.

'

.

iv) Ability of Actuation Module AM-628 (undervoltage module) to supply a sequencer input to load shed the "B" ED The subsequent anomalies on ESAS Facility 22 actuation cabinet were investi-gated and determinad by the licensee to be a "crowbar" actuation. According to the licensee, a "crowbar" actuation occurs when a voltage perturbation on the regulated (+15 VDC) results in a power interruption to the unregulated (+24 VOC) power supplies for each actuation module. This type of actuation response prevents the modules in ESAS Facility Z2 from receiving the required voltage potential, thus inhibiting actuation of ESA The inspector-witnessed a portion of the visual inspections performed on the actuation modules, electrical checks of the circuit diodes, bench test of actuation module operability, and restoration of the ESAS Facility Z2 cabine The inspector reviewed test data and the licensee's conclusions of the results of the test, and developed no further questions on this matte Inspector follow-up concluded that the licensee did not fully address the cause of the "crowbar" actuation. The inspector, expressed concern to the licensee on the adequacy of the test in determining the root cause for the

"crowbar" actuation, and the apparent urgency to complete in-service test T- l 88-21. The licensee informed the inspector this issue will be presented to i the work group responsible for in-service test T-88-21, and to the first line supervisors at the daily outage meetin The inspector will continue his follow-up in future inspection .0 Committee A:tivities

'

t The inspector attended Unit 2 Plant Operations Review Committee (PORC) neet-ings on January 21, January 26, February 1, and February 5,1988. Technical Specification 6.5 requirements for committee composition and quorum were met, t The inspector observed good regard for safety by the PORC in matters under i review. No inadequacies were identified, s

11.0 Management Meetings <

At periodic intervals during this inspection, meetings were held with senior ,

plant management to discuss the finding No proprietary information was identified as being in the inspection coverage. No written material was *

provided to the licensee by the inspector, t

!

l

,

E C -