Text

Insp Rept 50-443/97-07 on 971005-1206.No Violations Noted. Major Areas Inspected:Operations,Engineering,Maint & Plant Support Re Implementation of Security Program
	ML20198L604
Person / Time
Site:	Seabrook
Issue date:	01/07/1998
From:	; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	;
Shared Package
ML20198L585	List: IR 05000443/1997007; ML20198L581;
References
	50-443-97-07, 50-443-97-7, NUDOCS 9801160068
	Download: ML20198L604 (45)
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U. '. NUCLEAR REGULATORY COMMISSION ~

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REGION I

e Docket No.:'

50A43'

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. License No.:

NFF 86 '

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i i Report No.:

50-443/97 07

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Licensee:

North Atlantic 4nergy Service Corporation Facility: -

Seabrohk_ Generating Station, Unit 1

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Location:

Post Office Box 300 Seabrook, New Hampshire 03874 t

- Dates:

October 5,1997 - December 6,1997

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, inspectors:

Raymond K. Lorson, Senior Resident inspector William T. Olsen, Resident inspector Javier Brand, Resident inspector Intern J. McFadden,- Radiation Specialist -

T. Fish, Operations Engineer L. Prividy, Senior Reactor Engineer-J. McFadden, Radiation Specialist G. Smith, Senior Physical Security Inspector P. Frechette, Physical Security Inspector Approved by:

Curtis J. Cowgill, Chief, Projects Branch 5 Division of Reactor Projects

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EXECUTIVE SUMMARY Seabrook Generating Station, Unit 1 NRC Inspection Report 50-443/97-07 This integrated inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers a 9 week period of resident and specialist inspections.

Ooerations:

The plant was operated safely during the period. Shift management responded well to a residual heat remova; (RHR) system pipe leak. Several minor configuration control weaknesses were noted early in the period, however,it appears that operations management had initiated good corrective actions to improve performance during routine operations. Reportability determinations and follow-up

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operability assessments were good, y

Operators noted that an oilleak challenged the reliability of the positive

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displacement charging pump but did not ensure that it was promptly corrected (Unresolved item 97-07-01). Multiple station groups, including operations, failed to prcperly address a loose emergency feedwater steam admission valve limit switch which resulted in an necessary entry into a two hour hot shutdown technical specification action statement.

The licensed operator requalification training program was thorough and effective.

The plant oversight and assessment reviews functioned well during the period and

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contributed to plant i afety.

Maintanance:

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-The repair of a failed hot leg temperature instrument was well controlled and

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prompt. A good initit tive was identified to enhance instrument and technician procedural guidance.

The emergency feedwt ter (EFW) and emergency diesel generator (EDG) surveillance activities were r.erformod well. Troubleshooting activities to correct an adverse EDG start time trend wer-) effective. Minor weaknesses were noted in the program guidance for evaluating fluctuating gage indications, and scheduling EDG preventive maintenance activities. The licensee identified appropriate measures to address these concerns.

Enoineerina:

The licensee was taking appropriate actions to improve the content of the Seabrook 0'

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design basis. The EFW system was installed and operational consistent with the design requirements as described in the UFSAR.

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Except for several design deficiencies conceming the MS-V394 control modification, the electrical modifications we:e well supported with sound technical bases, however, the licensee has recognized a trend in design change notices (DCNs)

attributed to :nadequacies in initial engineerbg work.

The licensee took appropriate steps to evaluate and correct plant problems regarding the potential for high primary component tooling water temperatures during post-accident conditions and high vibrations in the steam piping for the steam generator feed pump turbines.

Weaknesses were noted in the control of temporary equipment in that it some non-permanent components were connected to or located near plant systems without a formal evaluation. The inspector did not identify any immediate operability concerns with the temporary equipment, and noted that the licensee initiated an ACR to review the program.

The licensee did not appear to promptly investigate and resolve identified degraded

plant conditions involving the residuai heat removal system and the positive displacement charging pump (Unresolved item 97-07 01). The licensee had a plan to address frequent control building air system air conditioning compressor falures, however, the inspector questioned whether previous f ailures had been properly categorized per the maintenance rule program (Unresolved item 97-07-03).

Plant Suocort:

The management of solid radioactive waste and of transportation of radioactive materials was generally effective. The volume of low level radioactive dry active waste (DAW) which was being generated continued to be low as the result of effective management in this area.

Weak attributes were noted in the scope of the Process Control Program (PCP)

document, in the documentation of the technical rationale for changes to the PCP, and in the procedure for updating scaling factors for radioactive waste streams.

Chapter 11, Radioactive Waste Management, of the Updated Final Safety Analysis Report (UFSAR) does not accurately reflect the current status of plant equipment and of methods used for radioactive waste processing.

Radiological housekeeping conditions in the waste processing building (WPB) were good. (Section R1.1).

The licensee maintained an effective security program. Management support was evident based on the implementation of the security program as documented in this report. Audits were thorough and in-depth, alarm station operators were-knowledgeable of their duties and responsibilities, communications requirements were being performed in accerdance with the NRC approved physical security plan (the Plan) and assessment aids had adequate picture quality. Security equipment iii

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- was being tested and maintained in accordance with the Plan and security training -

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was being performed in accordance with the NRC approved training and j

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2e Based on the inspectors' observations and discussions with security management i

-- end plant engineering, the inspectors determined that the licansee's provisions for

' land vehicle control measures satisfied regulatory requirements and licensee.

commitments. As an enhancement to the inspectioni the UFSAR initiative, Section 4.2 of the Plan, titled Thysical Barrier" was reviewed.- The inspectors detcrmined,

- by observations, that the protected area barrier was properly installed, maintained -

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and satisfied the requirements of the F1an.

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TABLE OF CONTENTS E X E C UTIV E S UMM A RY :............................................. 11

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TABLE O F CO NTENTS................. -..........

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- 1. Operations

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Conduct of Operations.................................... 1 01.1 General Comments (71707)...........................1

Operational Status of Facilities and Equipment................... 1 02.1 Routine Plant Tours (71707)............................1-04 Operator Knowledge and Performance (71707).................. 2 04.1 Response To Degraded Plant Equipment

..................2-04.2 Plant Equipment Configuration Control.................... 4 04.3 Event Reporting.................................... 4-05 Operator Training And Qualif'. ation

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Quality Assurance in Operations............................ 6 07.1 Safety Review and Oversight Meetings

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. Miscellaneous Operations issues.............................. 6

.08.1 (Closed) LER 97 007: Environmental Protection Program Non-

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Co m pli a n ce....................... ^................ 6 11. M ain t e n a n ce.................................................... ~r M1 Conduct of Maintenance (61726/62707)

.......................7 M1;1 Repair of a Failed Roactor Coolant Loop Hot Leg Temperature Instrument

.......................................7 M1.2 Surveillance Observations (EFW Turbine Driven Pump and "A" EDG)

...............................................7 M8 Miscellaneous Maintenance issues............................ 9 M8.1 (Closed) LER 97-014 Non-Conservative Residual Heat Removal Interlock Setting

...................................9 Ill. Enginee ring.................................................... 9 E1'

Conduct of Enginee:mg.................................... 9 E1.1 Emergency Feedwater System...........,.............. 9 E1.2 Mechanical Review

................................10 E1.3 Electrical Revie w.................................. 12 E1.4 Major Plant Modifications............................ 15 E1.5 Technical Resolution of Plant Problems................... 19 E2_

Engineering Support of Facilities and Equipment

........._........21 E 2.1 - Control of Temporary Equipment....................... 21 E2.2 Engineering Support For Degraded Plant Conditions.......... 22

E8 Miscellaneous Engineering issues............................ 24

. E 8.1 (Closed) LER 94-019 00: Electrical Relay Failures

...........24 E8.2 - (Closed) LER 50-443 / 97-006-00:Non Conservative Fuel Handling

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Accident Analysis Assumptions........................ 24 v

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IV. l Plant Support S... -. -.. :......... i........ >....................... 2 5

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Radiological Protection and Chemistry (RP&C) Controls.

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R1.11 Implementation of the Solid Radioactive Weste Program j........ 25

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R1.2. Compliance'with NRC and DOT Regulations for Shipping of Low Level -

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Radioactive. Waste (LLRW) for_ Disposal and Transportation of Other

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Radioactive Materials h.... -...,..............-........ 27.

R5; - Staff Training and Qualification in RP&C!....................... 28

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Quality Assurance in RP&C Activities..~.-...............-....... ' 28 '

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Miscellaneous RP&C lasues..........................-...... 29

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S1: - Conduct of Security.and Safeguards Activities -.=................ 30 S2 Status of Security Facilities and Equipment :...................... 30 ~

S2.1 Protected Area (PA) Detection Aids -...................... 30 S2.2 ' Alarm Stations and Communications.................... 31

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S2.3 Testing, Maintenance and Compensatory Measures.......... 31

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- Security and Safeguards St**f Training and Qualification (T&O)

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- Security Organization and Adm5istration -....................... 32 -

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- Quality Assurance in Security and Safeguards Activities........... 33

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S 7.1 -. A udits......................................... 3 3 Miscellaneous _ Security and Safety issues...................... 33

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S8.1 Vehicle Barrier System (VBS).........,................ 33

- S8.2 Vehicle Barrier System.. -............................ 34 -

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S8.3 Bomb Blast Analysis...... =....,................ <..... 34

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SP.4 1 Procedural Controls................................. 35

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V.-. M anage ment Meeting s.......................................... 3 5

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- Exit Meeting Summ ary................................... 3 5

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Review of Updated Final Safety Analysis Report (UFSAR)........... 35 X2

X3 Other N RC Activities.................................... 36 PARTIAL LIST OF PERSONS CONTACTED............................... 37 e

- PARTIAL LIST OF INSPECTION PROCEDURES USED

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PARTI AL LIST OF ACRONYMS USED.......... -......................... 39

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Report Details Summary of Plant Status The fecility operated at essentially 100% of rated thermal power throughout the inspection period. Operators reduced power to about 96% on October 27.to repair a failed hot leg-temperature instrument (Section M1.1), and shutdown the plant on December 5,1997 to repair pressure boundary leakage from the inlet piping to the B residual heat removal (RHR)-

pump suction _ relief valve (Section 04.1).

I. Operations

Conduct of Operations 01.1 General Comments (71707)

Using inspection F'rocedure 71707, the inspectors conducted frequent reviews of ongoing

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plant operations, in general, routine operations were perfurmed in accordance with station prncedures.and plant evolutions were completed in a deliberate manner with clear icommunications and effective oversight by shift supervision. Control room logs accurately reflected plant activities and observed shift turnovers were comprehensive and thoroughly addressed questions posed by the oncoming crew. Control room operators displayed good questioning perspectives prior to releasing work activities for field implementation. The inspectors found that operators were knowledgeable of plant and system status.

Operational Status of Facilit!as and Equipment 02.1 Routine Plant Tours (71707)

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Insoection Scone The inspectors routinely toured the plant and reviewed the condition of accessible safety-related systems and components. These activities included verification of: system configuration, power supplies, process parameters, support system availability, and operational requirements. Additionally, general area material condition and housekeeping status were noted. Systems inspected included:

Emergency Diesel Generators (EDGs)

Residual Heat Removal (RHR)

Safe:y injection (SI)

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Containment Building Spray (CBS)

Emergency Feedwater (EFW)

Primary Component Cooling Water (PCCW)

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Service Water, including the Cooling Tower (SW)

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Spent Fuel Pool Cooling (SF)

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L Emergency Electrical Switchgear

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l Equipment operability and material condition were acceptable in all cases. Early in the

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period the inspectors identified several examples.where scaffolding, ladoers, and other-

miscellaneous items were not secured or wore stored in close proximity to plant

equ4 ment. Some of tha~ldentified stowage discrepancies included: -

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A piece of sheet metal piece was left adjacent to safety-related PCCW system =

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piping.

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' A large amount of scaffolding material, used during refueling outege five, was-j

stored in numerous areas throughout the primary auxiliary buildian (PAB) and in

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some cases near safety-related components and piping.-

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The inspectors identified the issues to the licensee who promptly corrected the conditions.'

(Additionally, plant management and supervisory personnel perform 9d a Nmorshe,'nive

? station walkdown and identified numerous minor equipment and hous%png lasues. The -

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majority of these deficiencies we,re corrected during a two. day plant cleanup activity.

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sm * "@ Work requests were initiated to correct the materialissues where appropriete. Station

housekeeping and stowage conditions fullowing the plant cleanup were very good. The -

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- station walkdown was viewed as positive, however, the large number of it. sues identified /

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indicates that formal equipment and plant tours indicate an adverse trend in material condition of the plant. Station management initiated an Adverse Condition Report (ACR) to i

evduate the housekeeping program effectivenees.

. 04 ". Operator Knowledge and Performance (71707)

04.1 Response To Degraded Plant Equipment

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Insoection Scone The ins;;ectors reviewed the respon:e of operators and other station personnel to degraded equipment conditions.-

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- Qhservations and Findinas l

C Roaldual Heat Removal System Pipe Leak

On December 5,1997, four smallleaks were identified in the inlet piping to the RHR

suction side relief valve (Section E2.2). Shift management promptly reviewed the "

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I degraded pipe condition w'th appropriate engineering assistance, declared the B RHRi and B CBS systems, and the containment integrity inoperable, commenced a plant shutdown

. as required by TS 3.6.1.2, and made a one-hour non-amergency event notification per 10 1CFR 50.72. The operators completed the shutdown safely in accordance with operating-procedures OS1000.06," Power Decrease," and OS1000.03," Plant Shutdown From

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LMinimum Load To Hot Standby." The inspector concluded that the operability H

determination demonstrated a good safety focus, and thct the operators performed the

' plant shutdown well. -

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Emergency Feedwater Steam Admission Valve Limit Switch Deficiency

On November 12,1997, a quality assurance (OA) inspector identified that the MS V394 valve limit switch seal connectors were loose. The connectors were apparently not properly tightened following a modificatien activity that was performed during refueling outage five. The QA inspector reviewed this finding with at engineer who initiated and submitted a work request to the control room on December :3 The work request indicated that the connectors were part of the limit switch envirw. nental qualification boundary.

The control room processed this work request but did not identify that the loose connectors potentially rendered the steam driven amergency feedwater pump inoperable.

Since the A EDG had been removed from service for a planned maintenance outage about ten minutes before the MS-V394 work request was presented to the control room, the potential existed for both the steam driven EFW pump and the A EDG to be simultaneously inoperable. Technical Specification (TS) 3.8.1.1.2 requires the steam driven EFW pump to be operable anytime an EDG is inoperable. The control room personnel did not recognize the potential significance of this degraded condition sir'ce no attempt was made to secure the EDG outage or enter TS 3.8.1.1.2.

Subsequently (about 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months later), an engineer determined and informed the control room that the loose connectors rendered the steam driven EFW pump inoperable. The operators then entered TS 3.8.1.1.2 and initiated action to tighten the connectors. The connectors were promptiy tightened and the operators exited TS 3.8.1.1.2. The licensee has since performed testing which demonstrated that the connectors provided adoquste protection for the limit switches in their "as found" condition on December 12.

The inspector conc!uded that multiple personnel in several disciplines (operations, quality assurance, and engineering) were slow to recognize and fully evaluate the potential operability issues associated with the loose limit switch connectors :vhich resulted in a necessary entry into TS 3.8.1.1.2.

Positive Displacement Charging Pump Oil Leak

On November 4,1997, the inspector observed that a caution tag hung on the control switch for the positive displacement charging pump (PDP) indicated that the pump could trip off upon starting due to an oil leak from the PDP oil pressure sensing line. The inspector questioned the control room operators regarding the plans to repair this leak and determined that the operators were unaware of any plans to correct this leak.

The inspector followed up resnlution of this isne with the SE as discussed in Section E2.2. The inspector noted that although the PDP is not currently used for normal plant makeup it is credited as a boron injection source in the UFSAR, and is also used as an alternate piece of equipment in emergency operating procedure FR C.1, " Response To inadequate Core Cooling," and should have been rnaintained in a reliable condition. The-inspector was concemed that the operators tolerated a condition that may have affected

the reliability of the PDP. This issue remains unresolved pending further review. (URI 97-07-01)

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LShift management responded well to=a RHR pipe leski Othsr examples' were r id

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Linvolving multiple station groups,' including operations, that were not aggress'c in

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addressing degraded plant and equipment conditions..

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04.2' Mont EM.r.t Connguration Control

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The inspectors noted three minor configuration control events which occurred at the

_ beginning of the period involv!ng: mispositioning the rod control switch for about one shift,

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improperly securing a local leak rate testiand starting the wrong fan during testing. The

inspector was concerned that although each of these events was of minor significance, collectively they represented a potential adverse performance trend.

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The inspector discussed this concern with the operations manager who indicated that the

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- f station had performed a common caese investigation in May 1997_ to determine the root

"**couse(s) for en observed increase in mispositioned composiants during routine operations.' -

The inspector reviewed the investigation report.and noted that it provided several

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recommendations including improved component labeling, and in-progress activity tracking.

The inspector noted that operations management ned developed actions to improve component labeling and use of the procedure in-progress log. Additionally, the inspector

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- reviewed the " Component Mispositioning" performance indicator and noted that operations

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performance did not appear to be declining. The inspector concluded that the licensee had implemented good actions designed to improve human performence during routine -

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- operations. The inspectors will continue to closely monitoi this area.

04.3_ Event Reporting L

The licensee made several 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months non-emergency event reports during the period for seals found within the plant' intake structure. The reporting was consistent with Section 4.1 of the Environmental Protection Plani Appendix B of the_ Technical Specifications. The

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On October 8,1997, the licensee made a one-hour non-emergency event report per 10 CFR 50.72 to report that a developmental fue! performance model calculated reduced fuel

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rod internal pressure margins when compared to the previously licensed code. The fuel vendor determined that for certain limiting examples the internal pressure could cause the-

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no gap reopening criterion and the 17% clad oxidation limits to be exceeded. The licensee y

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performed an operability determination that concluded the Seabrook fuel would meet its,,

desien requirements due to_several mitigating design differences between Seabrook's fuel y

and tho'goneric industry bounding analysis. A detailed analysis is currently being (.

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i nperability determination and concluded that the assumptions used to support the b<

4 conclusion that the Seabrook fuel would be expected to develop'less internal pressure than e

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- the industry bounding case were valid.

OriNovember 5,1997 the licensee reported per 10 CFR 50.72 that the automatic

blocking setpoint for the anticipated transient without scram mitigation system actuation

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circuit (AMSAC) was non conservative. The design discrepancy appears to have occurred during the initial AMSAC circuit design and had the effect of removing the AMSAC blocking setpoint at 40% turbine vice reactor power. The significance of this issue at the time of the discovery was minimal since reactor power was at 100% and the AMSAC circuit was armed. The licensee was developing a modification to revise the blocking setpoint.

The inspector oncluded that the operators properly made the event notifications as required by 10 CFR 50.72, and the licensee follow up actions to review operability and.

develop long term corrective actions appeared appropriate.

Operator Training And Qualification a.

insoection Scoce (71001)

From November 17 21, inspectors evaluated the Seabrook licensed operator

' requalification training program,

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Observations and Findinos Recualification examinations: The inspectors reviewed the training sample plan and

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l determined the Seabrook staff adequately covered required knowledge and performance items. Overlap of examination material between each of the weeks was acceptable. The scenarios were well constructed with valid critical tasks, detailed expected operator actions, and clear competency standards. There was good variation in scenario events as well as minimal event duplication from scenario to scenario. Job performance measures (JPM1) were also well written and covered an appropriate range of topics. Also, the training staff used an excellent data system to track the overlap of JPMs administered-during the requalification training cycles.

- Examination administration: The facility examined one staff crew and one shift crew on

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the simulator. The evaluators, which included the assistant operations manager, critically assessed crew anu individual operator performance. Tho training staff conducted equally good evaluations during administration of the JPMs. The operators performed well and all passed their exams, which was consistent with the inspectors' conclusions. Throughout the examination week, the inspectors did not detect any indications of exam compromise.

Feedback system: The inspector reviewed a representative sample of licensed operator comments and noted the training staff effectively implemented feedback where appropriate. For example, based on operator feedback, the staff revised the emergency operating procedures to be consistent with the procedure operators use to open the main generator output breaker if it fails to open following a main turbine trip.

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Remedial trainino oroaram: The inspectors reviewed training records of operators who had failed portions of requalification exams and concluded the training staff had prescribed i

appropriate remediation.. lmproved scores on subsequent tests for these individuals indicated that the remediation had been effective.

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Ooeratcr license conditions:- The inspectors reviewed watch standing records and

determined that operators had met proficiency and license reactivation requirements. ' Also,

based on a review of eleven operator medical records (roughly 20% of all operator '

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-licenses), the inspector concluded that physical exams were being performed and documented properly.

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Conclusions

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The inspectors determined that the exams for licensed operator requalification training were challenging and included an acceptcble sampling of various kr owledge and -

performance areas. Where appropriate, the training staff revised the requalification program to incorporate lesson plans and test items that addressed changes to the plant.

Evaluators critically assessed crew and individual operator performance. Seabrook staff ensured licensed operators satisfied the medical and watch standing conditions of their licenses. The inspectors concluded that the f acility had a thorough and effective licensed

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operator requalification program that contributed to safe plant operation.

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- Quality Assurance in Operations 07.1. Safety Review and Oversight Meetings During this inspection period, the inspectors atteaded multiple self assessment and

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- oversight meetings including:

The Nuclear Safety Audit Review Committee (NSARC), held on October 15,1997;

Various Station Operational Review Committee (SORC) meetings;

Various Management Review Team (MRT) meetings.

The inspector found that the NSARC meeting was well organized and conducted. The meeting reviewed activities important to plant safety, and the technical presentations were thorough. The MRT meetings critically reviewed the adverse condition reports (ACRs).

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The SORC members generally asked probing questions.

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The inspectors concluded that the plant oversight and assessment reviews functioned well-duried the period and contributed to plant safety.

Miscellaneous Operations issues D

08.1 (Closed) LER 97-007: Environmental Protection Program Non-Complianct, On 'Aptil 25,1997, the licensee identified Environmental Protection Plan (EPP) program and l

' reporting deficiencies. These deficiencies ware attributed to weak program guidance and oversight. : The licensee committed to enhance the program controls, assign responsibility for the program elements, ano to increase the frequency of program. audits. The inspector x--

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< concluded that these actions were adequate, and this LER is closed.

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II. Maintenanca l

~M1 Comiuct of Maintenance (61726/62707)

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M1.1 Repelr of a Failed Reactor Coolant Loop Hot Log Temperature Instrument a.

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1The inspector reviewed the response to a failed reactor coolant loop hot leg temperature

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instrument on October 27. The inspector reviewed completed work packages, adverse condition reports (ACR), pre job phnning and briefing activities, and verified that TS action statements were entered as required.

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Obse:vations and Findinas

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qThe operators promptly placed the hot leg instrument in the " Trip" position as lequired by l

'TSs. > Prior to troubleshooting per work request (WR) 97 WOO 3204,the instrument was W'

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placed in the " Bypass" position which removed the trip signal, and the required TS action statement was entered.- Station instrument and control technicians attributed the failure-to a circuit card in the instrument circuitry. The technicians replaced the card and satisfactorily retested the instrument as directed by station work reque it 97W3204 and repetitive trisk sheet 97Rl05472603.

During review of the AC81 generated to review this activity the management review team L

(MRT) recommended that special l&C procedures be developed to expedite the

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troubleshooting and repair process. This recommendation was intended to reduce the amount of time that the instrument would be left in the bypass condition for troubleshooting.

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Conclualona l

The inspJctor determined that the licensee properly entered the required TS action

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l statement when the instrument troubleshooting commenced. Maintenance personnel promptly and effectively diagnosed and repaired th; instrument and returned it to service

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approximately twelve hours later.1he initiative to develop procedures to expedite and formalize instrument and controls repair activities was positive.

M1.2 Survelilance Observations (EFW Turbine Driven Pump and "A" EDG)

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Insoection Scoce a.

On October 6 and on November 13, respectively, the inspectors observed the turbine driven emergency feedwater pump (EFW) quarterly surveillance test and the "A"

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EEmergency Diesel Gencrator (EDG),. monthly surveillance test. The inspectors reviewed the

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. surveillance procedures, the Updated Final Safety Analysis Report (UFSAR), Technical Specifications (TS), previous NRC inspection findings, instrument calibration data, and reviewed the test results. The inspectors also observed troubleshooting and maintenance activities to correct a slowly increasing start time trend on the " A" EDG.

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Observations and EiDdings EFW turbine driven pump test

The surveillance procedure was per'ormed satisfactorily and the TS required acceptanco criteria were met. Control room operators and field personnel communicated and

= coordinated the test activities well. The required system test parameters such as flow, turbine speed, pump suction and discharge pressures were satisfied. Measuring and test'

equipment (M&TE) were verified to be in current calibration.

The inspector noted that the pump discharge pressure (PI-4248) was fluctuating between approximately 1550 and 1620 psig. A representative value was recorded during the test, however, the inspector questioned operations and test personnel to determine Seabrook's policy for gage fluctuations. The inspector noted that guidance in this area was not clear, and discussed this issue with Operation Management who indicated that the Operations Management Manual (OPMM) would be revised to clarify the guidance for evaluating gage

= fluctuations. The inspector concluded that the concern was properly addressed.

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"A" EDG monthly surveillance:

The inspectors verified that the "A" EDG properly started from a simulated SI signal, and reached the specified voltage and frequency within the required 10 seconds. The inspector also verified that the EDG was properly loaded per the test procedure. Field personnel properly monitored the EDG parameters.

Prior to the test the SE had identified an increasing EDG start time trend. The slower start

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times met TS limits, however, the SE developed a plan to investigate this trend. The plan included enhanced monitoring of key EDG parameters, and inspections of the air start system components. During these inspections, the SE determined that dirt in the air distributor intet filters had caused an increased starting air pressure drop. The air filter condition was correcteo,'and the EDG start time improved. The SE planned to develop a periodic maintenance activity to clean or replace the filters.

The inspector noted, based on discussions with the SE, and by review of a maintenance work order that the licensee did not have poskive controls to ensure that periodic cycling of the fuelinjector rods would not impact the EDG test results. The SE indicated that enhanced controls for scheduling this activity would be developed to address this concern.

The inspector was satisfied with this response, c.

Conclusion The EFW and EDG surveillance activities were performed well. Troubleshooting activities to corrout an adverse A EDG start time trend were effective. Weaknesses were noted in

. the, program guidance for evaluating fluctuating gage indications, and scheduling EDG

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preventive maintenance activities. The licensee identified appropriate measures to address these concerns.

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M8 Miscellaneous Maintenance issues M8.1 (Clored) LER 97 014 Non Conservative Residual Heat RemovalInterlock Setting The licensee determined that the increasing pressure reset for the RHR low pressure open permissive interlock was sbout 30 psig non-conservative. This could have potentialiy allowed the RHR suction isolation valves to be opened above the 365 psig TS limit. The significance of this event was minimal since the licensee had administrative controls in place to prevent opening these valves above the TS limit, and if inadvertently operated the valves would have shut at 394 psig well below the 495 psig RHR suction side relief valve setpoint. The licensee determined the root cause for this condition to be poor incorporation of the log lc deadband into the surveillance requirements, plans to submit a TS change to resolve this discrepancy. The inspector concluded that the licensee actions appeared reasonable. This LER is closed.

Ill. Enoineerina E1 Conduct of Engineering Insoection Obiectives and Methods The first inspection objective was to assess engineering pe-formance by evaluating the design of the emergency fee Avater (EFW) system. The inspectors reviewed calculations and engineering documents used to support system performance during normal and accident conditions. As a second objective, the inspectors evaluated the enpineering activities associated with the implementation of plant modifications and the technical resolution of plant problems.

The inspection consisted of mechanical and electrical reviews of the EFW system as described in various sections of the Seabroo)< Updated Final Safety Analysis Report (UFSAR) and in Design Basis Document DBD EFW 01, Revision 1, dated July 29,1997.

The inspection was performed in accordance with the applicable portions of Inspection Procedure (IP) 93809," Safety System Engineering Inspection" and IP 37550,

" Engineering". Inspection findings pertinent to the EFW system reviety were evaluated within the context of the licensee's response (Letter NYN-97012, dated February 7,1997)

to the NRC s 10 CFR 50.54(f) letter of October 9,1996.

E1.1 Emergency Feedwater System System Desian The Emergency Feedwater (EFW) system, as described in the Seabrook UFSAR, is required to provide the capability to remove heat from the reactor coolant system during emergency

- conditions-when the main feedwater system is not available. The system consists of two 100% capacity pumps (EFW-P-37A is turbine driven and *FW-P-37B is motor driven) with the primary water source from the condensate storage !

(CST). Each pump discharges to a common discharge header which serves 4-inch sup;ly lines to each steam generator.

Two normally open, motor operated globe valves are located in oach supply line (one each

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for Train A and B). Stop check valves (1-FW V76, V82, V88, and V94) are included in the individual supply lines to prevent main feedwater backflow to the EFW system portion located in the pumphouse which is normally depressurized.

The turbine driven pump receives its steam supply from steam generator A or B. The motor driven pump is poweied from the safety related 4160 volt bus E6 which is powered from the B train emergency diesel generator upon a loss of offsite power. The EFW system is started automatically on a loss of offsite power, low low level in any of the steam generators, a safety injection signal or an anticipated transient without scram mitigation system actuation signal.

Additional pumping capability is provided by the startup feed pump (FW P 113) that is part of the main feedwater system. However, this capability is not obtainect automatically since Jhe startup feed pump controls and the valves necessary to align flow to the EFW system are administratively controlled.

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E1.2 Mechanical Review a.

Insoaction Scoos (93809)

The inspectors reviewed calculations, englar ing and work control documents, and test procedures associated with the condansate storage tank (CST), the air operated, steam admission valves to the turbine driven pump, and the supply line stop check valves. Plant walkdowns of the system were also performed.

b.

Observations and Findinos CST Inventorv-The CST is a stainless steel tank located outdoors adjacent to the turbins building, it contains about 400,000 gallons and is equipped with a stainless steel floating cover.o minimize oxygen in the CST water. About half of the tank capacity is dedicated for use by

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the EFW system. Design information concerning the CST is included in UFSAR Section 9.2.6, " Condensate Storage Facility".

The inspector reviewed Calculation 737 51, Revision 2, " Inventory in C:adensate Storage Tank" and had discussions with the cognizant design engineer to understand the basis for the Technical Specifiestions requirement to maintain a minimum CST inventory of 212,000 gallons. The inspector deterrnined that this calculation had been reviewed during the licensee's review of the EFW system design basis infnrmation in response to the NRC's 10 CFR 50.54(f) request. As a result, the licensee found numerous inconsistencies between this calculation and information included in UFSAR Sectint; 9.2.6. This warranted a-detailed review of UFSAR Section 9.2.6 to substantiate the CST volumas with those found

.

- in design calculations. Many of the corrective actions to correct these inconsistencies, including an' update to the UFSAR, were identified in Adverse Condition Report 96-1442

' (December 20,1996).

a o

la response to ACR 961442 the licensee provided an evaluation of the CST calculations and supporting documentation to determine whether the present Technical Specifications limit of 212,000 gallons contained sufficient margin to account for all CST unusable volumes and still contain the design basis inventory required to cool the plant to residual heat removalinitiation. A major effort in this evaluation was being performed by Yankee Atomic in calculation SBC 792," Update of Seabrook Sttdion NSSS Cooldown Study".

While this calculation had not been fully documerited, the rebults and conclusions had been independently verified and the licensee was confident that the Technical Specifications minimum hventory of 212,000 gallons was sufficient to account for the unusable volumes due to the floating cover and the CSTlevelinstrument inaccuracies.

The inspector also noted that the licensee had opened the CST during the past refueling outage primarily to inspect and upgrade the floating cover including the neoprene material used to form the seal betweer. the over and the tank wall. Some plants had experienced problems with such tank covers as notad in NRC Information Notice 91-82. The inspector

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noted that the component inspection r+;.s o for this activity indicated that the overall

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condition of the CST was very good.

The inspector concluded that the lic; >see was taking positive mps to maintain the CST -

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and improve the Seabrook design basis information as a result c1 lessMs learned from the l

EFW system 10 CFR 50.54(f) review.

l l

Beckuo Air Sunolv - Turbine Driven Pumo Steam AdmissicQdsta l

Steam is supplied to the turbine driven pump from steam generators "A" or 'B" via air-operated valves,1 MS V393 or 1-MS V394, respectively. Piping downstream of these valves leads to a common supply header to the turbine. Pressure from the service air heaoer or a safety related backup air supply (consisting of high pressure gas bottles, L

pressure regulators and interconnecting tubing) maintains 1-MS 393 and 1-MS-394in the closed position when operation of the turbine driven pump is not required, in the event.'

a loss of pressure in the service air header, two in-series check valves are installed to isolate the non safety related portions of the service air system from the backup air supply.

UFSAR Section 9.3.1," Compressed Air Systems", specifies a backup air supply capacity requirement to operate each valve four complete cycles in a 10-hour period. The licensee i

had implemented this requirement in accordance with plant modification DCR 90 032. The inspector reviewed the post modification test procedure ES 91-1-44," Backup Nitrogen Gas Supply Test" and noted that it adequately demonstrated fulfillment of this UFSAR requirement.

The inspector questioned how the licensee periodically verified the Leat tightness of the in.

series check valves. The licensee indicated that a quarterly test is performed that consists

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of a qualitative backseat check of each valve under a large differential pressure. A periodic visualinspection of the valve internals is performed per the Seabrook check valve inspect!on program. While these valve tests have been satisf actorily and seem to demonstrate acceptable check valve performance, the inspector observed t. hat no periodic quantitative leak tightness check was being performed to assure that the backup air

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e capacity requirement of UFSAR Section 9.3.1'was being mat. The licensee l issued an :

Adverse Condition Report to evaluate this inspector observation.

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, Stan Check Valve Backleakaga

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' Stop check. valves 1-FW V82 and 1-FW V88 function to separate the pressurized main ifeedwater piping from the normally.depressurized EFW discharge, header'as stated in'

UFSAR Section 6.8.3. They also prevent potential vapor binding of the pumps and assure that the piping in the EFW pumphouse is maintained in accordance with its mode: ate ' a

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energy design criteria.

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During a walkdown in the pumphouse, the inspectors observed audible flow noises and j

elevated piping surface _ temperatures which evidenced system backloakage from these fl valves. The licensee had evaluated a backleakage concern regarding these N

'

refuel outage ORO5 sinco they had been repaired during refuel outage ORO4. The licensee.

gy%,." chose to monitor the piping temperatures adjacent to the valves since previous be'ckloakage problems had been experienced. -

The inspector noted that a temperature monitoring program using thermocouples attached to the piping surface was in place to continuously trend the EFW stop check valve backloakage based on piping temperatures. Continuous temperature monitoring was boing plotted and assessed on an ongoing basis by the cognizant system engineer. The highest temperature (approximately 130*F) was associated with the supply line associated with'

stop check valve 1-FW V82_('8" steam generator). In addition to monitoring piping

- temperatures, the licensee verified that the piping in the EFW pumphouse remained a depressurized despite the check valve backleakage.

'

Since the licensee was adequately monitoring piping temperatures, the inspector had no-immediate concerns regarding the piping operability. Deficiencies for these valves had

" * + been issued to effect permanent repairs.

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Conclusions

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The inspector concluded that the mechanical design of the EFW system was sound.- M a

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result of lessons learned from the EFW system 10 CFR 50.5A(f) review, the licensee was

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taking appropriate corrective actions to improve the content of the Seabrook design basis -

' information.

.E1.3 Electrical Review a.

Insoection Scone (93809)

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. mThe inspectors reviewed the EFW system electrical instrumentation and controls (l&C)

' included in the design basis documentation, design drawings, calculations, analyses, and bother engineering documents that are used to support system performance during normal

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and accident conditions.L

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-bc Observations anf Findmps Desian Basis Documentation Review The design' basis documentation described the electrical and I&C design of the EFW system and other supporting and interfacing equipment (steam generator, emergency diesel

. generator (EDG), safety injection' actuation system inputs, and CST). The bistables,.

teensorsi relays,viring, and controls needed to initiate an automatic start of the EFW ~

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system were includod consistent with the UFSAR requireinents. The inspectors noted that the motor driven EFW pump w'as appropriately powerod from the safety-related 4160 volt-bus (E6)'and backed up by the "B" train EDG The power supply requirements for the control devices of the EFW system and its supporting equipment, such as the steam admission valves to the turbina driven pump and the hiitiation logic sensors, were -

adequately designed and powered from the applicable power supplies required to support c

- EFW system operation during normal and accident operation, s

&^*4 During plant walkdowns the inspector verified that the manual controls for the EFW system pumps, control valves, the CST, and associated instrumentation were appropriately provided in the main control room, at the remote safe shutdown (RSS) panel or locally to support the proper operation of the EFW system on demand. During this review the -

i

. inspectors identified a deficiency in DSD-EFW 01. Specifically, the DBD erroneously indicated that two motor-oporr.ted valves (FW V156 and FW-V163fwere controllable from s the RSS panel.- The licensee issued an adverse condition report to correct this

< documentation error during the next update of the DBD. The inspectors also observed that the indicator status lights for two components on the "A" RSS panel were not lit. The licensee replaced the defective indicator bulbs.

Calculation Review The inspector reviewed the following design calculations:

9763 3 ED OO-8.? F, Revision 5, " Emergency Diesel Generator (EOG) Loading

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Calculation" i.

9763 3 ED 00-66-F, Revision 2, " Control Circuit Voltage Drop Calculation"

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9763 3-ED 00-14F, Revision 9, "125 V de System Battery, Charger, Motor Feeder

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j-Calculation" The EDG loading calculation in.;luded the maximum breke horse power rating for the motor

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._ driven EFW pump, in addition to this conservatism, the licensee had included the starting -

L loads of all applicable process related motor-operated valves (MOVs) in the initial starting

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. sequence of the EDG loading, even though these MOVs change their state later in the load

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, sequencenThe review of the ' worst-case voltage dip during the starting load sequences -

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indicated an' approximate 14% dip of rated bus voltage which was_ within the 20% voltage

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= < dip requirement limit shown in the UFSAR (Sections 8.3.1.1.e.5 and 8.3.1.2.b.2). 'The l

' inspector concluded this calculation to be acceptable.

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Regarding the review of the de system calculations associated with the EFW system

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components, the inspector cbserved that the turbine driven EFW pump steam supply valve solenoids would be adequately powered to open the respective valves automatically and i

admit steam to the turbine. The inspector also noted that ;he licensee had appropriately assumed the minimum battery system voltage of 105 V h in the calculations and applied-appropriate held cable voltage losses in evaluating voltage drops. The inspectors i

concluded that the de system component and control circuits associated with the j

automatic operation of the EFW system had sufficient voltage and were appropriately.

powered from the respective do buses to perform the intended design function.

.)

Instrument Calibration Data and Setooint Calculations The inspector reviewed the calibration data and associated setpoint calculations for eight water level, flow, and temperature instruments. The inspector identified one documentation deficiency and a procedural methodology issue during the review of these items.

Condensate storage tank (CST) water level instruments (CO-LT-4079 and 4096) are provided for local and main control room indication to assure a minimum CST inventory of 212,000 gallons during normal power operation (Technical Specification 3/4.7.13). The inspector determined that the existing instrument setpoint calculaion (737-37, Revision 4)

for these instruments did not include sufficient detail to account for instrument loop inaccuracies. However, as listed in the setpoint calculation to calibrate these instruments,

. the inspector noted that the licenses had besn using conservative loop inaccuracy values of i1.4 and il foot in calibrating the indication and alarm loop respectively. The licensee included the missing supporting data in en updated calculation on September 24, 1997, which the inspector considered to be appropriate.

[

During the review of the CST water level instrument (1 CO-LT-4096) calibration check procedure (IS1618.212) performed on July 17,1997, the inspector also noted thr.t one

"As Found" data point was out of the desired range. The l&C technician had corrected tho l

condition by adjusting the inst.ument back within the acceptable range and informed his l

supervisor as required by the proceduro. The inspector was concerned that the negative l

results from this instrument were not required to be reported to the operations department l

which was required to monitor and record this instrument per Technical Specifications.

l The inspector later determined that this instrument was classified as a non-Regulatory Guide 1.97 instrument where Seabrook "lS" category procedures are applicable and require r

notification to only the l&C supervisor for out of-calibration data. Other "more critical"

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Regulatory Guide 1.97 instruments, such as refueling water storage tank level instruments, requins the use of Seabrook "lX" category procedures that require notification to the control room shi,ft manager when an instrument is found out of calibration. The inspector had no immediate concern since the instrument was corrected properly prior to restoring it to operation. At the exit meeting of October 3,1997, the licensee agreed to evaluate this

, instrument calibration practice and later issued ACR 97-2207 to include a review of

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industry practices as part of this evaluation.

!

The inspector also reviewed the measuring and test equipment (M&TE) used to calibrate the EFW and steam generator instruments (1FW-FT-4214-2,1FW-PT-4253,1 RC-PI 405-1,

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[and 1FW U4257).EThe ' measuring devices used and the process measurement tolerances s M

l Included in tha'setpoint calculation for the loop calculation met the' accuracy _ requirement in ?

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z UFSAR Section_.17.2.12.3 of at least four times the accuracy of the equipment being

! calibrated.lHowever, per discussion with the licensee; the inspectors noted that the :

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H licensee was in the process'of reviewing and updating the many M&TE specification sheets-to reflect the accuracy changes or calibration ratio restrictions of the equipment.- The -

=

zinspectar will review licensee progress in ongoing efforts with M&TE sheets in a future.

inspection.i(IFl bO-443/97 07-02)s s The inspectors concluded that the licensee in general was appropriately calibrating and Jmaintaining the EFW system instruments.

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Surveillance Tests u.

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34The inspectors reviewed _the results of the May 1997 surveillance test (EX1804.015, Revision 6),* associated with the_"B" train Emergency Diesel Generator 18 month.

W + c:: uperability.and engineered safeguards pump and valve response time. The surveillanca test

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data indicated that (1) all EFW system components functioned properly including 3he auto 94 start of the motor driven pump and the functioning of steam adm:ssion valves as required;

_.

(2) the ecceptance criteria were well defined; and (3).the licensee had adequately

_

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= demonstrated the functionality of the EFW system and other supporting equipment such as

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the EDG,

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The inspector concluded.that the EFW system was installed and being operated in a

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r manner consistent with the design bases.

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Conclusions.

Based on the electrical review, th'e inspector concluded that the EFW system was installed

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and operational consistent with the design requirements as described in the UFSAR and the -

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jy** Adesign basis documentrAn unresolved item was opened to track the licensee's ongcing N

._offorts to update the M&TE specification sheets.

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E1.4 Major Plant Modfications l

Ta..

Insnection Scone (37550)

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The' inspector reviewed design modification packages, including items such as, safety

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evaluations, design inputs, calculations and design change notices (DCNs)s related to'the

following plant modifications: (1) DCR 96 016," Primary Component Coolic; Water Heat

Exchanger Replacement";(2) DCR 96 012;" Service Water System Pipe Joint Repair

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LModification"; (3) DCR 96-034, "EFW Flow Control Valve Time Delay"; and (4) MS-V394 Control Modification".' The review also included plant walkdowns to observe the newly

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Observations and Findinos DCR 96 016-Primary Comoonent Coolant Water (PCCW) Heat Exchanoer Reolacement i The' inspector noted that engineering thoroughly considered the various factors regarding tne selection of the new PCCW heat exchangers with titanium tubes replacing the 90-10 copper nickel tube material used in the prior heat exchangers. Titanium tube material was

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chosen on the basis of good heat transfer capability (better thermal conductivity), industry experience with sea water erosion / corrosion resistance, and no need for cathodic protection (with titanium clad heads). All surfaces in contact with sea water were constructed of titanium to minimize erosion / corrosion.

The inspector reviewed the thermal, hydraulic, and stress calculations in the heat exchanger vendor report (Joseph Oat Corporation Order J 2252). The heat transfer

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output, pressure dico calculations, shell side flow induced vibration analysis, and flow

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velocity profile at the inlet nozzle were found to be consistent with good heat e. changer design practice.' These resulted in improved thermal performance with comparable pressure drops used in prior analyses and precluded flow-induced vibration of the tubes from adverse flow velocity distnbution across the tube bundle. The replacoment heat exchanger has a larger heat transfer capacity and increased service water flow. Based on this new performance information, the licensee revised the containment calculations for

= accident conditions and determined that a reduced containment peak pressure and temperature wou!d result during design basis heat removal conditions following a postulated loss of coolant accident. The stress analyses included basic shell and tube stress, shell nozzle loading stress, seismic qualification stress ender safe shut down earthquake, anchor bolt and hole stresses, and bottom ring and gusset stresses. The inspector found the resulting stresses acceptable withia ASME Code requirements.

The inspector verified that the licensee had prepared a change to applicable sections of the UFSAR to reflect this modification. For exaruple, UFSAR Section 9.2.1 requires revision to

' = indicate that the tube material is titanium, and Table 9.2.1 " Service Water System Flows and Heat Loads" requires revision to indicate new service water system flow rates and heat loads.

During the plant walkdown the inspector noted that the heat exchanger, 2-inch drain lines off the bottom heads were not capped. This meant that only one normally closed valve existed to prevent seNice water from discharging to the auxiliary building floor. The inspector noted this condition to the cognizant design engineer who indicated that the

' original design intended for this drain line to be "hard piped" to the auxiliary building drain system. However, the "hard pipe" oesign was not implemented. The licensee issued an engineering work reqaest to evaluate this condition to determine what changes were appropriate to be consistent with service water drain lines serving other safety related heat

- oxchangers.

The licensee's design control manual provides instructions for categorizing the reason for issuing design change notices (DCNs). Engineering tracks these DCN reasons in 10 categories to take appropriate corrective action and improve the overall uodification process. The inspector reviewed the DCNs issued by the licensee during the

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Ilmplementation of this modification and wted that many were attributed to Category 6 (i.e., Ease of Constructiv) to facilitate construction resulting from dimensional deviations

'in matching the as built din ensions of the heat exchanger with the as-four'd pipe and foundation locations. Also, many DCNs were attributed to Categories P.A (Additional Design Details) ano 2B (Drawing Correction) which provide a muasure of the qua!ity of the initial engineering work applicable to the modification, in discussing this DCN data with engineering in ro'ation to other modifications, the inspector noted that engineering was aware that Categories 2A and 2B had the highest number of DCNs from all modifications currently bei g t scked. Engineering recognized that corrective actions were needed to

improve in this area by delivering better engineered work products wnen issuing the onginal mod: fications.

QCR 96-012-Seabrook Service Water Svstem Pios Joint Reosir Modification This modification involved the innovative use of AMEX 10/WEKO seals (a product of Miller

.

Pipeline Corpoistion) to resolve the degradation that had been noted at the 24-inch,

' cement-lined service water piping field weld joints. The choice of joint sealing was based on consideration of the expeditious nature of using the AMEX-10/WEKO seals and the availability of the option to replace the pipe in the future. The cognizant design engineer-

'

had presented this AMEX 10/WEKO seal application in an ASME paper NE Vol. 20 1996 Joint Power Generation Conference Volume 3, "Seabrook Station Service Water Piping l

Rafurbishment Using the Joint Seal Method".

The AMEX 10/WEKO seal is a thin, elastomeric circumferential boot of EPDM material that l

is non-water ebsorbent and is sufficiently flexi' ole to sustain thermal expansion and vibration. It is designed to isolate each field weld joint from the service water environment and prevent electrochemicalinteraction and any further degradation in the weld joint area.

l The seal is attached to the pipe innor diamett r with an application of quicksetting cement

'and is retained by 6% rnolybdenum stainless steel alloy corrosion resistant bands which -

are press fit to the pipe well. Tha design has provision for air leakage testing at each joint after installation.

During refuel outage OR-04, approximately 74 joint seals were installed in the service l

l water B Train underground piping and 10 joint seals were installed in the above-ground piping. During refuel outage OR-05, a similar number of joint seals were installed in the service water A-Train underground piping. The inspetor reviewed the 10 CFR 50.59 l

evaluation for this modification and noted that the AMEX 10/WEKO seal assemblies were l

classified as non-nuclear safety class, seismic category I components. The installation process was classified as safety related, since seal degradrtion could contribute to flo'v -

blockage of safety related equipment. While the seals have no pressure retaining function regarding the service water piping, they were classified as seismic category I because seal degradation during an earthquake could restrict service water flow and reduce cooling of safety related equipment.

The kspector reviewed the piping seismic qualification analysis and wall thicknoss computation for the apolicable field joints and found these calculations adequate to justify

. operability of the service water system. The licensee also evaluated the pressure drop in the system due to the presence of the sects,and found it to be inconsequential. The

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licensee is evaluating what tests and inspections are warranted in the future to assure that the seals perform properly as they age.

"

i Electrical Modifigatigng

The inspectors reviewed two EFW system electrical l&C system design changes

. Implemented recently at Seabrook.

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RCR 96-034,"EFW Row Control Valve Time Delav": This DCR added a 0.5 second time ~

delay by adding a new pickup relay into the control circuit for each emergency feedwater flow control valve to prevent occasional spurious valve isolations due to transients experienced by valve reset operations or perturbations within the control circuit supply.

- DCR 97 004 "MS \\ a4 Control Modificction": This DCR was implemented to resolve a

. single failure vulnerability identified in the EFW system design during the review of design

- basis dcaumentation in response to the NRC 10 CFR 50.54(f) request. The licensee

'

e determined that the automatic initiation of the EFW system may not occur if a design basis feed water line break (FWLB) or rnain steam line break (MSLB) occurred on the "A" steam generator and.an assumed single f ailure results in a loss of the "B" train solid state

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protection system (SSPS). A MSLB or FWLB on the "A" SG would result in insufficient -

= pressure at MS V393 to operate the turbine driven EFW pump (P 37A) from the "B" stsam gene ator.- in addition, an automati staa of the motor driven EFW pump (P-378) would also be lost due to the assumed single failure of the "B" train SSPS. This modification added redundant logic for MS V394 to ensure the capability for automatic operation of the turbine driven EFW pump on a loss of the "B" train SSPS in conjunction with the assumed single failure.

d The inspectors reviewed the design change modification packages consisting of design -

inputs,10 CFR 50.59 evaluation, specialty group technical review, calculations, and the post modification test results and found them acceptable. The inspectors noted that the applicable UFSAR Sections were appropriately identified in these design packages to be updated upon closure of this design change packages as p6r the established engineering

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department procedures.

The inspectors also noted that several design change notices (DCNs) were issued regarding the DCR 97 004 design package. Deficiencies ws:a identified in the original DCR such as s'

specifying an inconect switch. Also, DCNs were mt ded to correct problems in the circuit wiring connection design, field installation design, material selection and the testing package. -The inspector noted the licensee had appropriately documented these identified deficiencies in several adverse condition reports (ACRs 97-1698,1598, and 1700).

Several DCNs may have been attributed to implementing this modification on a " fast track" basis since the EFW design vulnerability was discovered during the last refueling outage

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and the licensee was committed to design and implement this modification prior to unit restart. These problems further evidenced the need for the I!censee to deliver better

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engineered work products when psuing original modifications.

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o An inspector follow item (IFl 50-443/96 01101)had been identified to track completion of

'this modification. This item is now closed.

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Conclusions

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Enc, neering gave thorough considerations to the various technical factors regarding the

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" selection'of the new PCCW heat exchangers.;The inspector verified the adequacy of the J

.;

- heat exchanger thermal,' hydraulic, and stiess calculations.9The licensee provided an *

innovative application of tne AMEX 10/WEKO seals to resolve the degradec' field weld;

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' joints in the 24-inch service wstor piping.

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Overall, the electrical modifications were well supported with a souad. technical design

t basis and in conformance with the established design control procedures. However, the.

inspectors noted that changes were required to correct several design deficiencies concerning modification DCR 97 004 which evidenced the need to deliver better

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engineered work products when issuing the original modification. This observation was

similar to the PCCW heat exchanger modification where many of the DCNs wore attributed

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. to design engineering. -

a cE1.5-Techrscal Resolution of Plant Problems

L-a.

. Inar>ection Scone (37650i

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The_ inspectors reviewed several examples regarding the licensee's resolutic7 of plant -

g problems including operability determinations made and corrective actions taken.

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b.: : Observations and Findinas l

Steam Generator Feed Pumn Turbine Steam Line Vibration.

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( The steam generator feed pump turbine (SGFPT) receives steam from a high pressure (HPP

p-or low pressure (LP) source as a function of plant load. The HP control valve is open at y'

plant loads up to 40%. At plant loads greater than 40%, the HP control valve is closed and steam is supplied to the SGFPT from the LP control valve.

. The inspector reviewed the engineering evaluations and associated calcukations conducted L'

in response to excessive vibration of th6 SGFPT HP main steam line. The vibration L

problem was observed by operations personnel and caused by a failure (i.e., repetitive

' '

opening and closing) of the HP control valve at about 1-second intervals. Shortly-

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1 afterwards, the cognizant mechanical engineer and two engineering managers performed a -

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field walkdown at 1:30 a.m. on September 16,1997, of_ the affected piping (an-

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approximate 115-foot length of main steam piping loosely supported by vertical spring -

' hangers and two,'40 foot branch lines with rigid guides ccNinuing to the SGFPTs 1 A snd 18). Based on this walkdown and further evaluation upon recurrence of the vibration

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problem _a day lateriengineering formally evaluated the vibration problem in l Engineering

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Evaluation SS EV-97 024 with the following observations and conclusiores;

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The HP pipe, vibration had deflections of * 2 inches axially and i18 inches laterally.

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4-p" i Vibration was noted also in the'12 inch LP piping to a lesser extent.

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After the plant walkdown on September 16,1997, the cognizant mechanical

engineer considered the piping in question to be operable based on prior stress calculations performed to justify 1993 modifications of the supports for this piping.

The liennseo porformed a visual examination at three high stress locations as

identified oy the stress analysis and found no indications of plastic flow or fatigun cracks.

The licenseo's analysis (Engineering Evaluation SS EV 97 024 dated September 29,

1907) demonstrated a maximum alterneting stress value that resulted in t. Jhtly exceeding the piping yield stress which meant that components of the piping system could potartially go plastic. Also, at this altomating stress level and using a simplified fatigue evaluation approach from ASME Section lil, Subsection NP 3650, the components in the piping system could have a short (l.o., hours as opposed to days) design f atigue life. However, tSe absence of any negative data, such es cracks or pipe indications, indicated that the analytical conclusions reached in the engineering evaluation rogarding plastic flow and f atigue were conservative.

The l'consoo indicated that improvements to the SGFPT piping supports would be pursued i

in future outages alor7 with the performance of surf ace and volumetric examination of the p!pe at the high st'ess locations to confirm those conclusions and assure the long term adoquacy of this piping.

Hiah PCCW System Tomoeratures under Post Accident Conditions The licensea identified during their 10 CFR 50.54(f) review a concern regarding the possibility that all PCCW pumps could trip due to high temperature under post accident conditions. The pump high temperature trip function is discussed in Section 9.2.2.5 of the Seabrook UFSAR. Specifically,in Calculation MSVCS FAG 16, Revision 1 the licensee determined a PCCW system post accident temperature of 134.5'F (PCCW heat exchangor dischargo) which was based on the PCCW heat exchanger plugging allowance and the allowable pump degradation. With a PCCW pump trip setpoint of 1350F and including an instrument uncertainty of 5.4'F, there was a potential that all PCCW pumps could trip under post accident conditions. This information was included in ACR 961351 which was issued on December 9,1996.

The inspector reviewed the licensee's eutuation of this problem which culminated in an operability determination which was approved by the onsito safoty review group on January 22,1997. A main cause of the problem was due to the throttled positions of the PCCW heat exchangor outlet valves (SW.V19 and SW V20) which ensured a positive heat exchanger outlet pressure in accordance with a prior design change but also resulted in reduced servico water flows. The system performance that could result in a peak PCCW temperature of 134.5 F was based on (1) a single service water and PCCW pump each

. fully degradcd; (2) service water temperature was 65'F with flow through the vont pipe (candy cano) since the return flow to the transition structure was assumed to be blocked; and (3) the PCCW heat exchangor was assumed to heve 200 tubes plugged. The licensee ovaluation for operability used a servico water supply temperature of 65 F, took credit for current performance of the servico + ar and PCCW pumps as demonstrated by inservice l

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testing information, and assumed that 200 tubes were plugged although actual tubes plugged in each heat exchanger was less than 100. The licensee's results ir.dicated a peak PCCW heat exchanger outlet temperature of 125.6'F which was less than the pump trip setpoint including instrument uncettainty.

The inspector concluded that the licensee had performed a bounding analysis to support operability of the PCCW system. The inspector noted that valves SW V19 and SW V20 are now full open with the new PCCW heat exchangers installed (DCR 96 016in Section E 1.4).

c.

Conclusions The licorisee has taken appropriate steps to evaluate the vibration effects on the SGFPT HP and LP piping regarding maximum stress levels and the piping design fatigue life. Future licensee efforts wore indicated to examine welds at high stress locations and to improve the piping support system in resolving tne problem concerning the riotential for high PCCW temperatures during post accident conditions, the licensee performed a bounding analysis to support system operability.

E2 Engineering Support of Facilities and Equipment E2.1 Control of Temporary Equipment a.

Insoection Scoos (71707)

The inspectors performed several plant tours to evaluate the controls of temporary equipment. The review included personnelinterviews and review of applicable l

documentation.

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b.

Observations snd Findines The inspectors reviewed the administrative procedmes that govern the control of temporary equipment including: MA 4.8, " Control of Temporary Equipment," MA 4.3,

" Temporary Modifications", and MA 4.5, " Configuration Control During Maintenance &

Troubleshooting." The inspectors noted examples wher0 the above procedures did not require an evaluation or documentation for temporary plant equipment either connected to or installed near permanent plant equipment. These examples included:

An alternate sample pump was connected to the Plant Vent - Wide Range Gas Monitor

(WRGM) located inside the Primary Auxiliary Building (PAB). This pump had been installed since April 1991, however no formal docuraentation existed to indicate that the installation had been reviewed and found to be teceptable to leave this pump installed. The inspector noted that procedure MA 4.8, did not require the evaluation to be documented for temporary equipment installed in a Seismic Category I building,.

The SE indicated that the pump was not conaidered to be temporary equipment or a temporary modification, as defined by proceduro MA 4.5 since an approved operating procedure existed for this piece of equipment.

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The SE determined that this pump was not required to be connected to the system.

Chemistry procedure CX901.13, was revised to require disconnecting of the pump

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when not in use, and Engineering Work Request (EWR 97 0558) was initiated to

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evaluate addition of this pump on applicable drawings and to allow the permanent connection in a valved out condition.

The inspectors noted a small metal test manifold in the 10 foot elevation of the PAB (mechanical penetration area), located in the vicinity of several safety related safety i

injection system components. The manifold was installed to periodically vent the SI,'

and RHR systems. The inspector was concerned that no formal controls existed that authorized this manifold to remain in place and that if it became unrestrained it could

have impacted safety related equipment during a seismic event,

,

The inspectors also identified several exampas where temperature monitoring

instruments were installed external to plant components without any formal controls or

-evaluation. The station policy for non intrusive monitoring equipment was to hang a

- - caution tag on the temporary monitoring equipment.

The licensee initiated ACR 97 2514to evaluate Seabrook's policies and methods for-control of temporary equipment. The inspectors did not identify any specific operability concerns with the currently installed equipment but noted that the program controls could result in a non appropriate application of temporary equipment. The inspector discussed this observation with station management and they agreed to

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reWew the issue.

c. Conclusions The inspector determined that Seabrook's program for control of temporary equipment was weak in that it allowed non permanent components to be connected to or located near plant systems without requiring a formal evaluation. The inspector did not identify any immediate oper;bility concerns with the temporary equipment, and noted that the licensee initiated an ACR to review the program.

E2.2 Engineering Support For Degraded Pisnt Conditions a. Insoection Scone (37751)

The inspector reviewed the effectiveness of engineering support to promptly and effect ely resolve degraded equipment conditions, b. Observations and Findinas Positive Displacement Charging Pump Oil Leak

-The inspector discussed the PDP oil leak described in Section 04.1 with the system engineer (SE). The SE indicated that this oil leak had been a recurrent problem due to the location of the pressure switch, and that the plan was to defer repair of this leak pending completion of a modificatloa to relocate the pressure switch (DCR 96-024). The inspector

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was concerned that the leak would not be promptly repaired and discussed this issue with

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the Technical Support Manager who indicated that the station would promptly implement DCR 96 024 and repair the oil leak.

The inspector performed a walkdown of the PDP to locate the oil leak and noted that a deficiency evaluation tag for numerous minor oil leaks indicated that the PDP was not

- expected to be run for the remainder of the cycle and that station personnel should not initiate any addhional deficiency tags on the PDP. The inspector.was concemed that this 1 statement could inhibit station personnel from reporting equipment deficiencies and

' discussed this observation with the Technical Support Manager who agreed that the_ tag was worded improperly. The licensee subsequently removed the evaluation tag and

. canceled the deficiency evaluation tagging program, and initiated an ACR to determine why this program had not been successfully implemented.

.The inspector subsequently learned that, priar to this review, DCR 96 024 had been j

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. canceled about one month earlier by the Station Modification Resource Committee (SMRC)

since the modification was viewed only aa en enhancement. The modification was then

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approved by the SMRC on November 13 and installed by November 18. During the post-

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n.odification testing the inspector noted that the PDP pump was unable to deliver any flow due to a loss of priming suction to the fluid coupler oil pump. The licensee corrected this problem, and attributed this problem due to not running the PDP for an extended period of time. 'An ACR was initiated to review this event and to develop a period maintenance

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activity to run the PDP.

Roaldual Heat Removal System Pipe '.eak h

On December 3, tho' system engineer (SE) identified boron residuo external to the insulation on the inlet pipe for the suction relief valve (RC V 89) to the 8 RHR pump suction. The inlet piping to the RC V 89 valve is AMSE Class ll piping. On November 4, this issue was

~ discussed at the Station Manager's Meeting. The inspector reviewed this issue with the SE on November 4; and learned that the boron residue had been noted on the pipe v

insulation as early as November 1996, and that a work request had been initiated to remove the insulation and inspect the piping during the outage. This activity apparently had not been performed and the station started up without inspecting this pipe for leakage.

Maintenance and engineering personnel entered the containment on December 5,1997, to

' remove the insulation and inspect the pipe. The inspector also inspected this pipe and noted 4 minor pressure boundary leaks (one just below the upper weld, and three on the "

~ lower weld area). The licensee promptly declared the affected system inoperable as discussed h Section 04.1. The Scensee formed an _ event team to determ_ine the_ro_ot causels) for this event. This ret iew was on-going at the conclusion of the inspection, and

. the issue of whether the licen'se responded promptly to investigate the indications of the

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degraded RC V 89. inlet pipe wndition will remain unresolved (URI 97 07 01).

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Control Building Air (CBA) Air Conditioning System n

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eThe control building air system air conditioning compressors (CBA AC-5A and CBA AC 5B)

ioxperienced' multiple f ailures during the period. Specifically, the CBA AC 5A compressor-l

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f ailed on October 15 (after having been replaced on October 2), and the CBA AC 5B f-

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on November 25. The compressora are part of the Cc rol Room Emergency Makev.4 And Filtration System and the f ailures required the operators to enter TS 3.7.6. TN inspector learned that over the past few years those compressors have f ailed frequeth.

particularly during the winter months when the compressors are lightly loaded. The licensee developed a modification to improve the reliability of this system. The inspector noted that this system had not been categorized as an A 1 system por the maintenance rule and questioned whether this was appropriate given the apparent system operating deficiencies. This issue will remain unresolved pending review of the maintenance history on the CBA system. (URI 97 07 03),

c.

C.onclusions The licensee did not appear to promptly investigate and resolve degraded RHR and PDP conditions. The licensee had a plan to address frequent CBA compressor failures, however, the inspector questioned whether previous failures hd been properly categorized per the molntenance rule program, inspection of these issues was ongoing at the

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conclusion of the in.npection.

E8 Miscel!aneous Engineering issues E8.1 (Closed) LER 94 019 00: Electrical Relay Failures The licensee reported that between 1989 and 1994, twenty high spood lockout HEA relays installed in safety related circuits and four HEA relays in non safety circuits at Seabrook Station failed periodic preventive molntenance testing. During the testing the relays usually tripped in tho "as found" condition but after rosetting, f ailed to trip during the following l

functional test. The licensee replaced allinstalled GE HEA relays with Electroswitch LOR 1

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relays. There were no adverse safety consequences as a result of this problem. The licensee determiriod that the worst case failure of a lockout relay would not result in the plant being outside !ts design basis or in an unanalyzed condition, it was also determined t

l that this condition did not constitute a substantial safety hazard pursuant to 10 CFR 21.

l The LER met the requirements of 10 CFR 50.73, and the inspector had no further questions regarding the event.

l E8.2 (Closed) LER 50 443 / 97 006 00:Non Conservative Fuel Handling Accident Analysis Assumptions

The licensee reported that the transport time assumptions to the containment air purge l

(CAP) system following a postulated fuel drop accident were non-conservative. The licensee re analyzed the fuel drop scenario using the actual spent fuel building design parameters and determined that the as-built system met the UFSAR, General Design Criteria (GDC) 19, and Exclusion Area Boundary (EAB) limits. Additionally the licensee implemented a modification to reconfigure the CAP inlet ducting to increase the transport time. The inspector determined that the licensee's corrective actions to resolve this issue were adequate. This LER is close.

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IV. Plant Suncort R1 Radiological Protection and Chemistry (RP&C) Controls R1.1 Implementation of the Solid Radioactive Waste Program a.

insoection Scooe (IP 86750-01)

The inspector reviewed the licensee's solid radioactive waste program. Information was-gathered through observation of activities, tours of the radiologically controlled areas

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including the waste processing building, discussions with cognizant personnel, er.d review and evaluation of procedures and documents, b.

Observations and Findinas

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Radioactive liquid waste streams, not directly releasible, were processed through a vendor-supplied filter /demineralizer processing equipment skid until they could be released directly.

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Spent resins from the vendor supplied skid have been transferred to high integrity containers (HICs) stored in the shielded storage area of the waste processing building (WPB) and also to the two spent resin sluice tanks in the WPB. At this time, there were eight HICs in the shielded storage area containing spent resin from the skid operation and three empty HICs to be used for resin which was to be sluiced from the spent resin sluice tanks in the near future. The spent resin sluice tanks also contained all of the other spent resins generated since initial plant operation. Spent cartridge filters have been transferred to HICs stored in the boron thermal recovery dominoralizer cubicles (those domineralizers have not been in use) below the floor of the 25 ft elevation of the WPB.

Dry radioactive waste (DAW) volume has been kept low through a ' green is clean' program and by a radioactive waste sponsor nrogram which has been in effect for three years. The sponsor program involved a represen, 'ive from the radioactive waste group interf acing with and attending meetings with each group in the plant which influenced the generation of radioactive waste. Their contribution was to help each group work on ways to minimize radioactive waste generation. Good performance in this area has continued, and the generation rate of DAW has remained low. Only sixteen boxes of compacted DAW have been generated since January of 1996. All of those boxes were currently stored on site.

No radioactive waste, other than DAW, has ever been shipped for disposal by this licensee.

All of the previous DAW shipments occurred in December of 1995, Wasro was currently stored in the shielded storage area of the WPB, the asphalt storage building, and the Unit 2 cooling tower. Waste inventory records for these three storage areas were up to-date.

Revision 01 (10-0197)of the Seabrook Station Process Control Program (PCP) was reviewed. This document addressed the dewatering of depleted bead resin and process cartridge filters and a natural drying process for process cartridge filters. The body of this-document mentioned or discussed one vendor report, two radioactive waste department instructions, and one study (Health Physics Siudy/ Technical Information Document

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(HPSTID) 96 015," Natural Drying Times for Saturated Process Filters"). This PCP document referenced additionally two vender procedures and one radioactive waste group procedure. All of these vendor reports, licensee procedures, and the HPSTID dealt with

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the dewatering process except for one procedure which dealt with spent resin cask operation. This PCP document did not address the processing of DAW. The Process Control Program document, as described in Section 1.25 of the Technical Specifications,

was meant to contain or reference current information covering all aspects of radioactive waste disposal. While all the detailed current information was availeNe in the licensee's other documents and procedures, the licensee egreed that the PCP,itself, could be improved and stated that the PCP would be revised to include more complete information describing or referencing the waste processing and packaging methods used, the applicable regulatory requirements, and the burial ground requirements. This issue will be reviewed during a subsequent inspection (IFl 50-443/97 07 04).

A study (Health Physics Study / Technical Information Document (HPSTID) 96 015," Natural Drying Times for Saturated Process Filters")(12 0196) was used to justify a change to the PCP to include the natural alt drying option. The inspector reviewed this document and observed that the actual filter / container configuration to be used for drying, in the case that this option were to be utilized, was not clearly described. Also, calculations and comparisons of volumes of waste containers and waste cartridge filters were not clearly

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presented. The licensee agreed that a clearer description of the final filter / container configuration to be used for drying and a clearer presentation of the calculations and comparisons of volumes of waste containers and waste cartridge filters would be helpfulin more clearly supporting the rationale for this change to the PCP. The licensee stated that clarifications would be added to the HPSTID. This issue will be reviewed during a subsequent inspect!on (IFl 50-443/97 07 05).

The inspector reviewed the generation and use of scaling factors or correlation factors to quantify the concentration of difficult to measure radionuclides in materials or for classification of wastes. This review included Chemistry Procedure (CP) 5.1 (Rev.15),

Isotopic Characterization of Radioactive Waste. Paragraph 4.3.1.5 of CP 5.1 recommended but did not require that correlation ratios be updated on a regular, periodic basis (une year for Class B and C wastes; two years for Class A waste) and did not

- address updating correlation ratios for m in the sh!pment of radioactive material (non-waste). This paragraph recommended but did not require that correlation ratios be updated when the ' trigger' rat!o of selected fission products to activation products change in the reactor coolant system (RCS) by a factor of ton. This paragraph gave examples of ' trigger'

radionuclide ratios and did not identify specific ones. Per this procedure, smears (to characterize DAW) and laboratory filters used to filter the liquid process streem (to characterize high activity process filters) are analyzed annually for the difficult to measure 10 CFR 61 radionuclides. The preferred method for spent resia characterization is direct'

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sampling of the resin. This procedure did not describa what mechanisms and frequency were used to check on a regular and periodic basis on the appropriateness of the current correlation ratios in use in between the annual sampling results; or to check in the event of a major upset likely to significantly change the isotopic spectrum of radioactive shipn.ents l

or radioactive waste. The licensee stated that this procedure would be revised to address

- the aforementioned items. This issue will be reviewed during a subsequent insnection (IFl l

50-443/07 07 06).

l The inspector verified that samples were sent out annually for analysis of the difficult to-

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measure 10 CFR 61 radionuclides and that the results were used to update the correlation

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ratios for DAW, for RCS waste / material, and for spent fuel pool (SFP) waste / material. The inspector reviewed results for multiple samples sent out for analysis during the first half of 1997 and a documented evaluation of those results, "10 CFR 61 Isotopic Correlation Ratio Evaluation During Cycle 5," dated October 10,1997.

Radiological housekeeping conditions in the WPB were good in that radioactive materials were properly labeled and stored. Alsleways were clear, and all materials and equipment were stored in an orderly f ashion.

c.

Conclusions The management of solid radioactive waste and of transportation of radioactive materials was generally effective. The volume of low level radioactive dry active waste (DAW)

which was being generated continued to be low as the result of effective management.

Weak attributes were noted in the scope of the Process Control Program (PCP) document, in the documentation of the technical rationale for changes to the PCP, and in the procedure for updating scaling factors for radioactive waste streams. Radiological F

housekeeping conditions in the WPB were good.

R1.2 Compliance with NRC and DOT Regulations for Shipping of Low Level Radosctive Waste (LLRW) for Disposal and Transportation of Other Radoactive Materials a.

innosction Scoos (IP 86750-01 and Tl 2515/133)

The inspector reviewed the licensee's transportation of radioactive materials. Information was gathered through discussions with cognizant personnel and review and evaluation of procedures and documents, b.

Observations and Findinas The documentation for eight recent shipments was reviewed. As stated earlier in this report, the licensee had not shipped any radioactive waste since December of 1995.

These recent shipments involved cartridge filters for analysis (Department of Transportation (DOT) Type A Quantity), chemistry samples (DOT Limited Quantity)(3), robotics equipment (DOT Type A Quantity), fuelinspection equipment (DOT SCO (Surface Contaminated Object) l). material specimen (DOT (Low Specific Activity) LSA II), and laundry (DOT LSA-II). The shipping paperwork was in compliance with the NRC and DOT regulations for transportation of radioactive materia!.. Up to date copies of licenses for all facilities to which the licensee shipped radioactive materials were on file and available for inspection, c.

Conclusions The licensee's control of the transportation of radioactive materials was effective.

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R5 Staff Training and Qualification in RP&C

a.

Insoection Scoce (IP 86750-01)

The inspector reviewed the qualifications and training of selected radioactive waste personnel, information was gathered through discussions with cognizant personnel, and review and evaluation of documents, b.

Observations and Findinas The inspector reviewed the Training Program Description for Rad Waste Technicians (assistant radioactive waste (RW) technician, RW technician, and sonfor RW technician),

the Training Program Description for Hazardous Materials, and course descriptions for RW1010C (Management of Radioactive Material), RW1034C (Shipping Refresher), and RW1010l(Management of Radioactive Matedal). The training programs provided for initial and annual continuing training and the recurrent training as required by 49 CFR 172.704 overy three years. The inspector also reviewed the training records for the radioactive waste technicians and for the individuals authorized to certify low level radioactive waste or radioactive material shipments and authorized to use the computerized program for determining waste classification and shipping classification.

The training program materials were extensive, detailed, well documented, and covered the required information. The training records were readily available and well organized and showed that the current qualification and training status of radioactive waste personnel met requirements.

c.

Conclusions The training program for radioactive waste personnel was well managed, and their current training status was satisfactory.

R7 Quality Assurance in RP&C Activities a.

lnsoection Scoce (IP 86750 01)

The inspector reviewed the licensee's quality assurance (OA) activities for solid radioactive waste management and transportation of radioactive materials. Information was gathered through discussions with cognizant personnel and review and evaluation of documents, b.

Observations and Findinas The Nuclear Safety'and Assessment Audit Report No. 96 A07 02/A25123, Process Control Program /Radwaste, was conducted from July 29 through August 2,1996. This i

audit evaluated the Process Control Program, hand calculations / computer analysis, l

records / document control, handling / storage of radioactive waste, bill of lading / manifests, receipt of radioactive material, packaging / radiological limits / communications, administrative control OA program / assurance of quality /self assessment, training and qualification, and waste characterization. This audit resulted in two adverse condition reports (ACRs) and

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five recommendations. The ACRs addressed the lack of incorporation of Environmental Compliance Manual requirements into radioactive waste procedures and the lack of performance of self assessment commitments by the Radioactive Waste group. This audit noted that the PCP only addressed dewatering resins and filters while the FSAR describes

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processes such as solidification and encapsulation, that the PCP did not include processing methods for dry active waste (DAW), and that there was a discrepancy between the PCP document and a procedural filter dewatering worksheet for the acceptance criteria for dewatering filters. Based on this audit's documentation and findings, the audit covered the full scope of the program under audit and was a good assessment of program quality.

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f A Quality Assurance Surveillance Report (QASR 97 0061) was conducted from September 8 through October 1.1997 and covered the Procer,3 Control Program and radioactive waste activities. The draft report included three draft ACRs for a lack of detailin the rationels inr changes to Revision 0 of the PCP, for lack of revision to Szction 11.4 of the UFSAR to reflect a change to the PCP covering the processing of spent filter cartridges by

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the dowatering methods in place for spent bead resins, and for the potential for blockage I~

" of a design basis floor drain by a bag of radioactive waste in the WPB. This surveillance had good depth of review and produced several findings requiring the evaluation of need for corrective action, c.

Conclusions l

The licensee's Nuclear Safety and Assessment audit and surveillance reports in the Process Control Program and Radioactive Waste areas were good assessments of program quality.

R8 Miscellaneous RP&C lasues A recent discovery of a licensee operating their f acility in a manner contrary to the Updated

' Final Safety Analysis Report (UFSAR) description highlighted the need for a special focused review that compares plant practices, procedures, and/or parameters to the UFSAR descriptions.

!

l While performing the inspections discussed in this report, the inspector reviewed the

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applicable portions of the UFSAR that related to the areas inspected. The inspector found

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that Section 11.4 of the UFSAR, Solid Waste Manegement System, did not appropriately

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reflect observed plant practices, procedures, and/or parameters. A very large portion of l

Section 11.4 was devoted to equipment and processes which were not currently in use j

and not contemplated for use in the near future (ex. asphalt solidification / encapsulation l-system, boron recovery system, and evaporators), while processing methods and equipment, which were currently being utilized and/or were contemplated for near term use

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(vendor filter / resin skid and spent resin and filter cartridge dewatering), were addressed l

= only briefly. Also, Section 11.4 was not updated in a timely manner for a change in the PCP (Section R7).

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in a licensee letter dated February 7,1997,in response to the NRC's request for information pursuant to 10 CFR 50.54(f) regarding adequacy and availability of design bases information, the license committed to perform a comprehensive review of the

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-- 30 UFSAM, During this inspection, the licensee stated that the rev!r w of Chapter 11 of the

. I UFSAR had already been scheduled.

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Conduct of Security and Safeguards Activities

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The inspectors reviewed the security program during the period of November 17 20,1997.

l Areas inspected included: management support; protected area (PA) detection equipment; j

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training and qualification; audits and the vthicle barrier system. The purpose of this =

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"1 inspection was to determine whether the licensee's security program, as implemented, met

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the licensee's commitments and NRC regulatory requirements..

j b.1 Obaarvations and Findinas a

m w. %.; Management support >was evident based on the effective implementation of the security j

- program as documented in this report.

' Alarm station operators were knowledgeable of their duties and responsibilities and-l

security training was being performed in accordance with the NRC appioved training and

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qualification (T&O) plan..

The PA detection equipment satisfied the NRC approved physical security plan (the Plan)

commitments and security equipment testing was being performed as required by.the Plan.

Maintenance of security equipment was being performed in a timily manner as evidenced

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- by minimal compensatory posting associated with non functioning security equipment..

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Conclusions

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"The inspectors determined that the licensee was conducting its security and safeguards

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activities in a manner that protected public health and safety.

Status of. Security Foollities and Equipment

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82.1 Protected Area (PA) Detection Aids

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lnanection Scons I

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1The inspectors conducted a physical inspection of the PA intrusion detection systems i-(IDSs) tol verify that the systems are functional, effective, and met the licensee's Plan l

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Observations and Findinas On November 18,1997, the inspectors determined by observation and selected testing

.iat the IDSs were functional and effective, and were installed and maintained as described in the Plan.

c.

ConclusiOD The PA IDSs met the licensee's Plan commitments.

52.2 Alarm Stations and Communications

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a.

Insoection Scona Determine whether the Central Alarm Station (CAS) and Secondary Alarm Station (SAS)

are: il equipped with eppropriate alarm, surveillance e id communication capability,2)

continuously manned by operators, and that 3) the systems are independent and diverse so

- that no single act can remove the capability of detecting a threat and calling for assistance, or otherwise responding to the threat, as required by NRC regulations, b.

Observations and Findinns Observations of CAS and SAS operations verified that the alarm stations were equipped with the appropriate alarm, surveillance and communications capabilities.

Interviews with CAS and SAS operators found them knowledgeable of their duties and responsibilities. The inspectors also verified through observations and interviews that the CAS and SAS operators were not required to engage in activities that would interfere with their assessment and response functions, and that the licensee had exercised communications methods with the local law enforcement agencies as committed to in the Plan, c.

Conclusion The alarm stations and communications met the licensee's Plan commitments and NRC requirements.

82.3 Testing, Maintenance and Compensatory Measures a.

IDanection Scone Determind whether programs were implemented that will ensure the reliability of security-

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related equipment, including proper installation, testing and maintenance to replace

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' defective or marginally effective equipments Additionally, determine that when security-related equipment failed, the compensatory measures put in place were comparable to the g

effectiveness of the security system that existed prior to the failure.

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Observations and Findinas

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Review of testing and maintenance records for security related equipment confirmed that

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I the records y m on file, and that the licensee was testing and maintaining systems and equipment as - 'mitted to in the Plan A priority status was assigned to each work

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request and rep. were being completed in a timely manner for all work necessitating

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compensatory measures.

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c.

Conclusions Security equipment repairs were timely. The use of compensatory measures was found to

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be appropriate and minimal.

Security and Safeguards Staff Training and Qualification (T&O)

a.

Insoection Scone Determine whether members of the security organization were trained and qualified 'o i

> perform each assigned security-related job task or duty in accordance with the T&O plan.

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b.

Observations and FindiDER On November 18,1997, the inspectors met with the security training staff and discussed the training program including response capabilities and its effectiveness and reviewed training records for security officers and re:ponse drill scenarios and critiques. The inspectors also interviewed a number of supervisors and officers to determine if they possessed the requisite knowledge and ability to carry out their assigned duties, c.

Conclusions The inspectors determined that training had been conducted in accordance with the T&Q

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plan. Based on the supervisors' and officers' responses to the inspectors' questions, the training provided by the security training staff was considered effective.

S6 Security Organizaticn and Administration a.

insoection Sqggg Conduct a review of the level of management support for the licensee's physical security program.

b.

Observations and Findinas

+The inspectors reviewed various program enhancements made since the last program inspection, which was conducted la March 1997, and discussed them with security management. These enhancements included evaluation of the bids received for a new security computer system, vital area access control system, and upgrades to both alarm stations.

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Conclusions Management support for the physical security program was determined to be good.

Quality Assurance in doourity and Safeguards Activities 87.1 Audts i

s.

Insoection Scone

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Review the licensee's Quality Assurance (QA) report of the NRC required security program

audit to determine if the licensee's commitments as contained in the Plan were being

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satisfied.

b.

Observations and Findinna

. The inspectors reviewed the 1997 OA audit of the security program, conducted April 21-

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'May 2,'1997,(Audit No,' 97 A04 02). The audit was found to have been conducted in

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accordance with the Plan. To enhonce the effectiveness of the audit, the audit team included several independent security specialists.

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The audit report identified 6 findings. Two of the findings were related to safety issues at

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the firing range, three were related to documentatior' inconsistencies and one finding addressed human error relative to proper display of security identifiestion badges. The inspectors determined that the findings were not indicative of programmatic weaknesses.

The inspectors determined, based on discussions with security management and a review of the responses to the findings, that the corrective actions were effective, c.

Conclusions A review concluded that the audit was comprehensive in scope and depth, that the findings were reported to the ap;,ropriate levels of management, and that the audit

~l program was being properly administered.

.

Miscellaneous Security and Safety issues

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88.1 Vehicle Barrier System (V88!

General On August 1,1994, the Commission amended 10 CFR Part 73, " Physical Protection of Plants and Materials," to modify the design basis threat for radiological sabotage to include i

the use of a land vehicle by adversaries for transporting personnel and their hand carried

' equipment to the proximity of vital areas and to include the use of a land vehicle bomb.

The amendments require reactor licensees to install vehicle control measures, including vehicle barrier systems (VBSs), to protect against the malevolent use of a land vehicle.

Regulatory Guide 5.68 and NUREG/CR 6190 were issued in August 1994 to provide

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guidance acceptable to the NRC by which the licensees could meet the requirements of the

amended regulations.

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17,1996, letter from the licenses to the NRC forwarded Revision 20 to its An April physical security plan. The lotter stated, in part, that vehicle control measures meet the criteria of Regulatory Guide 6,68. An NRC January 31,1997, letter advised the licensee that the changes submitted had been reviewed and were determined to be consistent with the provisions of 10 CFR 50.54(p) and viere acceptable for inclusion in the NRC approved security plan.

This inspection, conducted in accordance with NRC Inspection Manual Temporary Instruction 2515/132," Malevolent Use of Vehicles at Nuclear Power Plants," dated 18,1996, essessed the implementation of the licensee's vehicle control measures, l

including vehicle barrier systems, to determine if they were commensurate with regulatory January requirements and the licensee's physical security plan.

S8.2 Vehicle Barrier System o.

Insoection Scone The inspectors reviewed documentation that described the VBS and physically inspected the as-built VBS to verify it was consistent with the licensee's summary description submitted to the NRC and was in accordance with the provisions of NUREG/CR 6190, b.

Observations and Findirlan The inspectors' walkdown of the VBS and review of the VBS summary description disclosed that the as built VBS was consistent with the summary description and met the specifications in NUREGfCR 6190, c.

Conclusion The inspectors determined that there were no discrepancies in the as built VBS or the VBS summary description.

S8.3 Bomb Blast Analysis a.

Insoection Scone The inspectors reviewed the licensee's documentation of the bomb blast analysis and verified actual standoff distances provided by tho as built VBS.

b.

Observations and Findinas Tha inspectors' reviow of the licensee's documentation of the bomb blast analysis determined that it was consistent with the summary description submitted to the NRC.

The inspectors also verified that the actual standoff distances provided by their as-built VBS were consistent with the minimum standoff distances calculated using NUREGICR-6190. -The standoff distances were verified by actual field measurements and review of scaled drawings.

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Conclusion i

No discrepancies were noted in the documentation of bomb blast analysis or actual

standoff distences provided by the as built VBS.

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l 88.4 Procedural Controls

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Inanection ScoLa

- The inspectors reviewed applicable procedures to ensure that they had been revised to -

include the VBS. -

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Ohaarvations and Findinns j

t The inspectors reviewed the licensee's procedures for VBS access control measures, i

. surveillance and compensatory measures. The procedures contained effective controls to l

provide passage through the VBS, provide adequate surveillance and inspection of the W

VBS, and provide adequate compensation for any degradation of the VBS.

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Conclualon The inspectors' review of the procedures, applicable to the VBS, disclosed no discrepancies, j

V. Management Meetings t

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X1 Exit Meeting Summary

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The resident inspectors presented the inspection results to members of licent.se management, following the conclusion of the inspection period, on December 22,1997.

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' The licensee acknowledged the findings presented,

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a The specialist inspectors presented the inspection findings to members of the licensee management at the conclusion of their inspection activities on October 3, October 10, November 19, and November 20,1997. The licensee acknowledged the findings.

The inspectors asked the licensee whether any materiala examined during the inspection

should be considered proprietary. No proprietary information was identified.

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,X2-Review of Updated Final Safety Analyals Report (UFSAR)

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' A.recent discovery of a licensee operating its facility in a manner contrary to the UFSAR

description highlighted the need for a special focused review that compares plant practices,

procedures, and parameters to the UFSAR description. Since the UFSAR does not.

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activities to the NRC approved physical security plan, which is the applicable document.

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r-While performing the inspection discussed in this report, the inspectors reviewed Section

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4.2 of the Plan titled, ' Physical Barriers." Based on discussions with security supervision, procedural reviews, and direct observations, the inspectors determined that barriors were installed and maintained as described in the Man and applicable procedures.

X3-Other NRC Activities

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Region i and NRR managers visited the site on December 1, and December 2,1997. The -

f visit consisted of a plant tour, and personnelinterviews.

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PA9TIAL LIST OF PERSONS CONTACTED North Atlantic Enerav Service Corocration W. DiProflo, Station Director R. Faix, Engineering Supervibor G. Kline, Technical Support Manager T. Pucko, Licensing Engineer s

R. White, Design Engineering Manager

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H. Anderson, Waste Services Dept. Manager C. Berry, Training Supervisor

M. DeBay, Assistant Operations Manager

' + W. DiProfio, Station Manager B. Drawbridge, Director of Services

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G. Gram, D} rector, Support Serv!ces

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F. Haniffy, Radweste Ops & Shipping Supervisor R. Hickok, Trainir.g Manager

G. House, Precessing Supervisor

S. Kulback, Security Supervisor

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J. Kwasnik, Senior Red Scientist

R. Litman, Chemistry Dept. Manager

R. Messina, Security Supervisor

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M. Ossing, NRC Coordinator

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T. Pucko, Regulatory Compliance

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D. Perkins, Nuclear Oversight J. Rafalowski, C&HP Project Supervisor

D. Robinson, Senior Chemist

D. Roy, Operations Training Supervisor

B. Seymour, Security and Safety Manager

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J. Sobotka, Regulatory Compliance Manager G. St. Pierre, Operations Manager

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R. Thurlow, HP Technical Supervisor D, White, Nuclear Oversight

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Contractor C. Goodnow, Chief of Security, Green Mountain Security Services

NaC R. Lorson.- Senior Resident inspector

+

J. Brand, Hasctor inspector

'+

J.~ McFadden, Radiation Specialist

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W. Olsen, Resident inspector

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T. Fish, Operations Engineer L. Prividy, Senior Reactor Engineer J. McFadden, Radiation Specialist G. Smith, Senior Physical Security inspector l

P. Frechette, Physical Security Inspector

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+ Denotes those present at entrance meeting on October 6,1997 Denotes those present at exit meeting on October 10,1997

" Denotes those present at exit meet;r's on November 20,1997

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P/NTIAL LIST OF INSPECTION PROCEDURES USED iP 88750 01 Solid Radioactive Weste Mana9ement and Transportation of Radioactive Materials Tl 2515/133 Implementation of Revised 49 CFR Parts'100179 and 10 CFR 71 i

. IP 37550, En94neering.

.l lP 93809,

- Safety System Engineering inspection i

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ITEMS OPENED, CLCSED, AND DISCUSSED Onanad

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. URI 97 07 01, Evaluate whethat the prompt corrective actions were

. Implemented for evidence of residual heat removal system h

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' piping, and a positive displacement charging pump oil leak,

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IFl 97 07 02, Update measuring and test equipment specifiestlun sheets URI 97 07 03,.

Evaluate Whether The Control Building Air Conditioning System Was Properly Categorized Per Maintenance Rule Requirements.

IFl 5 443/97 07 04i increase scope of PCP IFl 5 443/97 07 05, Additionalinformation to be added to HPSTID LIFl 5 443/97 07 06, Revise procedure CP 5.1 f.

Cleasd IFl 9&1101 Complete follow up of MS V394 control modification LER 97 007i,

Environmental Protection Program Non Compi;ance LER 97 014,-

Non Conservative Residual Heat Removal System Setpoint

- LER 94 019i.

Electrical Relay Failures

LER 97 0061-

-Non-Conservative Fuel Handling Accident Analysis

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j PARTIAL LIST OF ACRONYMS USED ACR Adverse Condition Report CFR Code of Federal Regulations j

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DAW Dry radioactive Waste DOT Department of Transportation l

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HIC High Integrity Container i

HPSTID -

Health Physics Study /TechnicalInformation Document e

PCP Process Control Program OA Quality Assurance

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RCA Radiologically Controlled Area i

RCS Reactor Coolant System RF&C Radiological Protection and Chemistry i

rtW Radioactive Waste

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SFP Spent Fuel Pool J;

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UNAR-Updated Final Safety Analysis Report l'

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v t e Inspection Report - Seabrook - 1997007
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 \|
1997 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 7(0) 8(0) 9(0) 80(0) 99(0) Security Operator Exams Emergency Planning Other

IR 05000443/1997007

Text

REGION I

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