IR 05000443/1989009
| ML19325E195 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 10/23/1989 |
| From: | Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19325E193 | List: |
| References | |
| 50-443-89-09, 50-443-89-9, GL-89-10, IEB-85-003, IEB-85-3, IEIN-88-068, IEIN-88-68, NUDOCS 8911020216 | |
| Download: ML19325E195 (21) | |
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U.S. NUCLEAR REGULATORY COMMISSION
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REGION I
f Docket / Report No.:
50-443/89-09 License No.:
Licensee:
Public Service Company of New Hampshire
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1000 Elm Street Manchester, New Hampshire 03105 i
Facility:
Seabrook Station, Unit No.1, Seabrook, New Hampshire
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Dates:
August 18 - October.10, 1989 I
Inspector:
A. Cerne, Senior Resident Inspector l
Project l
Engineer:
A. Chu, NRC Office of Nuclear Reactor Regulation
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Approved By:
S C k N, 3 4-lo/23/mi l
Ebe C. McCabe, Chief, Reactor Projects Section 3B~
Date
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Areas Inspected: Operational safety, maintenance & surveillance, radiological f
controls, security measures & logs, quality verification activities, reportable v
events, and a technical review of certain design modification activities.
i Results: Maintenance and surveillance were well managed, except for one ques-
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tionable operational decision relating to the emergency power supply for the operable diesel generator fuel transfer pump (section 4.2).
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There was~ timely investigation, effective response, and comprehensive correc-l tive action planning for a Radiological Occurrence Report concerning radiography
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access controls (section 5.1) and for evidence of improper handling of a radio-
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logical survey by a health physics technician (section 5.2).
An active quality verification program was evident and included trending and internal reviews, and corrective action initiatives (section 7.1).
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Two licensee reports require further actions and subsequent NRC review. One
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concerns a self-identified Technical Specification violation regarding the Con-
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l densate Storage Tank Enclosure (section 8.1) and the other involves a 10 CFR 21 evaluation of a potential Atmospheric Steam Dump Valve deficiency (section
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8.2).
In addition, licensee actions regarding evaluation of missed monitoring
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instrumentation testing is considered unresolved (section 8.1).
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TABLE OF CONTENTS
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1.
Persons Contacted.....................................................
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2.
Summary of Activities (94600, 30702).........................
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Operational Safety (71707)...........................................
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3.1 Plant Operations.................................................
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3.2 Plant Tours....................................................
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Maintenance / Surveillance (61726,62703,92703).......................
4.1 Maintenance & Design Change Implementation Activities...........
p 4.2. Surveillance Activities.........................................
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5.
Radiological Controls (84750)........................................
i 5.1 Genera1.........................................................
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5.2 Investigation of a Health Physics Technician....................
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Security (81700).....................................................
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Quality ' Veri fication Activities (35502, 92701).......................
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7.1 Audits Trending & Corrective Action Fo11owup...................
r 7.2 Field Assembly Inspection & Records Followup (RI-89-A-112)......
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8.
Review of Licensee Reports (92700, 90712, 93702).....................
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8.1 Licensee Event Reports..........................................
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8.2' Part 21 Reports.................................................
8.3 Reportable Events at Other P1 ants...............................
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9.
Technical Review of Modi fications (37828, 92702).....................
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9.1 Control Bu i l di ng Ai r Sys t em.....................................
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f, 9.2 Safety Parameter Display System.................................
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t 10. Ma n a g eme n t Me e t i ng s ( 30703)..........................................
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DETAILS 1.
Persons contacted - New Hampshire Yankee (NHY)
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E. Brown, President and Chief Executive Officer
'J. DeLoach, Executive Director of Engineering and Licensing
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- B. Drawbridge, Executive Director of Nuclear Production
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T. Feigenbaum, Senior Vice President and Chief Operating Officer
J. Griilo, Operations Manager R. Hanley, Operations Training Manager
- T. Harpster, Director of Licensing Services
- D. Moody, Station Manager
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'J. Peschel, Operational Programs Manager
- N. Pillsbury, Director of Quality Programs J. Vargas, Manager of Engineering J. Warnock, Nuclear Quality Manager
- Attended exit meeting conducted on October 10, 1989.
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Other licensee and contractor personnel were also contacted.
2.
Summary of Activities F
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2.1 Senior Resident Inspector Activities
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One senior resident inspector (SRI) was assigned to the site during the entire inspection period. The SRI conducted routine and un-
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announced backshift, including deep backshift, inspection activities (112 total hours, with coverage provided during 12 backshift hours
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and 2 deep backshift hours).
Certain licensee sponsored training of its employees was monitored, including procedure compliance training
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and a lecture on core values and work ethic presented by the NHY
President and Senior Vice President. The SRI also participated in
the inspection of the licensee's partial-participation annual emer-
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gency prepeiedness exercise on September 27, 1989; attended the
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public meeting on September 6, 1989 and the enforcement conference on September 7, 1989 held in response to the Region I Augmented Inspec-tion Team inspection findings; and participated in the Systematic Assessment of Licensee Performance Board deliberations for Seabrook on August 23, 1989.
- 2.2 Visitino Inspector and Manacement Activities
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An NRC Systematic Assessment of Licensee Performance (SALP) Board convened at the NRC Region I in King of Prussia, Pennsylvania on August 23, 1989 to evaluate the performance of activities at Seabrook
Unit I for the period from August 1,1987 to June 30 1989.
The SALP
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Report (50-443/87-99) was issued on September 26, 1989.
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I A management meeting between the NRC and licensee personnel was held on September 6, 1989 to discuss the results of the NRC Region I Aug-mented Inspection Team (AIT) inspection 50-443/89-82.
This open meeting was held at the University of New Hampshire in Durham, New Hampshire and included a public comment period at the conclusion of the discussion between the NRC and the licensee.
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An enforcement conference was held in the Region I Office in King of
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Prussia, Pennsylvania on September 7, 1989. This meeting between NRC i.
and licensee personnel was convened to discuss the assessment of safety significance, root cause(s), and ir.terim and 'inal corrective actions for the potential violations identified in AIT Inspection Report 50-443/89-82. An Enforcement Conference Meeting Summary was issued on September 19, 1989.
Beginning on September 25, 1989, a Region I physical security inspec-tor conducted an unannounced five day inspection of the licensee's security controls, equipment and plan implementation.
A meeting was held with licensee management on September 29, 1989 to discuss the results of this inspection. These results will be documented in
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Region I Inspection Report 50-443/89-12.
On September 26, 1989, the Region I Reactor Project Section 3B sec-tion chief toured the site and held discussions with senior licensee managers and control room operators.
On September 27, 1989, the licensee conducted a partial participation annual emergency preparedness exercise.
NRC observation and inspec-E tion of this exercise was performed by a team of five Region I and headquarters personnel, beginning with an announced entrar,ce meeting g
and briefing with licensee personnel on September 26, 1989.
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licensee conducted a drill self-critique and were informed of pre-l liminary NRC inspection results at an exit meeting on September 28,
1989.
These results will be documented in NRC Region I Inspection Report 50-443/89-10.
On October 4, 1989, a project engineer from the NRC Office of Nuclear l
Reactor Regulation commenced a three day visit at the site to review the status of certain issues documented in supplements to the Seabrook Safety Evaluation Report (SER). The visit results are dis-cussed in section 9 of this report.
2.3 Plant Activities The plant remained in operational mode 5, cold shutdown, with primary coolant temperature between 120F and 140F and the reactor coolant system vented at the top of the pressurizer. Maintenance and modi-fication activities shif ted from train ' A' to train 'B' equipment as
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i the train 'A' residual heat removal system was returned to service.
Major work was conducted on the control building air, containment building spray, service water and component cooling water systems.
E.4 Reorganization of Licensino Services The inspector was briefed on a planned reorganization of the NHY Licensing Services group. The draft charter and expanded functions of.
the licensing group were reviewed and discussed with the Director of Licensing Services.
The licensee indicated that revision to the Final Safety Analysis Report is required as a result of this re-organization. A licensee review is being conducted to submit th>
necessary FSAR changes to the NRC Office of Nuclear Reactor Regula-tion.
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Operational $3fety 3.1 Plant Operations
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The inspector observed plant operations during regular and backshift l
inspections of the control room and during routine tours of the plant.
In the control room, plant logs, night orders, technical specification action statement status, and alarm conditions were re-viewed and operators were interviewed regarding control board indi-cations and system lineups. The critical safety function (CSF)
e status and its applicable status trees were examined during control i
room visits to check operator understanding of plant instrument availability and system status.
Even though the use of the displayed CSF status was not applicable to the current operational mode 5 con-
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ditions of the plant, control rnom operators on different shifts were
all cognizant of the reasons why specific CSF parameters displayed
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unreliable or abnormal indications.
The inspector also discussed with operations personnel a recent re-vision to the Operations Management Manual (OPMM, Revision 20), pro-i viding guidance for mode 5 and 6 operation with respect to equipment
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operability for conditions not covered by the Technical Specifica-
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tions.
For example, "in mode 5 with the RCS loops filled, the re-
quired operable RHR pump shall be the pump associated with the oper-
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able diesel generator." The inspector's observation of shif t activi-
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ties confirmed operator cognizance and compliance with the new OPMM
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provisions.
The inspector reviewed an operations department report of control room phone unavailability, for both the Emergency Notification System (ENS) and Nuclear Alert System (NAS) lines, to the NRC headquarters
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and state notification centers, respectively.
The licensee's in-ternal review of phone unavailability over the last nine months identified no root cause or pervasive long term problem, but deter-mined that further review of unavailability was prudent.
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i review will be accomplished over an additional nine month period, j
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after which an analysis of the need for corrective measures will be
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accomplished. The inspector discussed with the licensee their plan for further root cause identification and review of control room phone problems and had no further questions.
It was noted that, on
both October 2 and 3, 1989, the NAS phone was taken out of service
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The inspector was
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notified of this work in advance, as was the NRC headquarters duty officer in accordance with 10 CFR 50.72. The inspector verified that backup phone lines were available for~ emergency notification during
the time the NAS was out of service.
I In anticipation of potential severe weather conditions associated with the approach of Hurricane Hugo to New England, the inspector
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discussed with operations personnel the licensee's severe weather
preparations. Closure and temporary sealing of the outside hatch plugs to the train 'B' equipment vault were verified. The inspector i
reviewed abnormal operating procedure OS 1200.03 establishing plant provisions to minimize the adverse effects of severe weather condi-
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tions and conducted an unannounced deep back shift inspection of the
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plant on the night of Septamber 22-23, 1989 to ensure plant readiness
.and operator awareness of potential weather problems. No NRC con-
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cerns were identified.
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t With respect to overall plant operations during this inspection period, the inspector identified no violations or unresolved safety questions and confirmed both acceptable operator performance in rou-
tine mode 5 evolutions and knowledgeable response to NRC questions
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regarding system lineup and component status, i
a 3.2 Plant Tours
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The inspector conducted several general inspections throughout the
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plant to witness work activities in progress, check equipment status
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and required valve positions, and spot check housekeeping and overall administrative controls of outage work conditions. The following observations and conclusions were made:
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small work items and debris, particularly within the containment i
building, required additional attention.
The inspector noted l
improvement in this area as the inspection progressed.
l Control of temporary scaffolding and staging equipment within
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the station were in compliance witn Administrative Procedure MA l
4.10 and appeared adequate to ensure system and component oper-l ability were not adversely affected by temporary equipment use.
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Roving fire watch patrols were being implemented in accordance
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with Technical Requirement No.11 for hourly checks of fire
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seals and other fire rated assemblies renderad nonfunctional by l
authorized work activities.
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Valve positions (e.g., accumulator isolation valves closed),
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valve system lineups (e.g., service water and fuel supply valves open for required diesel generator operability), and locked valve status (e.g., pressurizer vent for the re&ctor coolant
system locked open) -- all were spot checked and found to be
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properly positioned, controlled and the status correctly docu-
mented in accordance with Technical Specification and system
lineup' requirements.
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Certain sticker / tags located on equipment installed within the
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plant (e.g., penetration seals serving as fire barriers) incor-
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rectly referred to controls and removal authorization specified
in old construction procedures which have been superseded by i
station procedures. Discussion with the licensee indicated
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their intent to remove the outdated procedures from the Unit 1 applicable procedure control program and revise the affected
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sticker / tags to reflect the correct references.
With respect to the areas inspected during the plant tours and to the
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questions raised by the inspector on specific issues, the licensee was i
responsive to both the destions and any items requiring additional man-l agement attentior..
No violations were identified and no unresolved safety
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concerns remain open.
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Maintenance / Surveillance i
4.1 Maintenance & Design Change Implementation Activities The inspector witnessed specific maintenance and design change work in progress in the plant, reviewed the associated work request and design coordination report implementatiori packages, and discussed the
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work activities with the cognizant craftsmen and technical support l
engineering personnel.
In selected cases, records of installation, e
maintenance and testing were reviewed to verify compliance with the (
applicable code and operability requirements.
The licensee's nuclear
quality organization's involvement in the observed work activities
was verified, either by documented evidence of hold points and in-l spection or by observing quality control personnel in the performance of their duties. The inspector also reviewed licensee work proce-t dures and related design documents, as necessary, to confirm the
acceptability of the work.
The following list represents tha dif-t ferent activities, equipment and records examined by the inspector.
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Documenta Work MS0514.05 MOVATS testing on residual heat removal l.l (RHR) valve RH-FCV-610.
89 WOO 3902 Damaged seal tight fitting on limit switch
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for service water valve SW-V-4.
DCR 86-081 Installation of steam generator secondary
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manway access platforms.
ECA 99/114572A Valve testing and setpoint data for the pressurizer safety valves.
L DCR 86-032 Rerouting tubing for a steam generator level
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transmitter.
89W001371 Service water valve SW-V-18 flange bolting
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rework.
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DCR 87 311 Work associated with installation of new
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containment building spray (CBS) and RHR
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88W001765 Inspection of service water valve.SW-V-5
worm gear.
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quest, and DCR and ECA stand for a design coordination request and
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engineering change authorization respectively.)
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The inspector evaluated the above work activities against the applic-
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able design specification, code and standard requirements.
Pressur-
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izer safety valve testing was verified to be in compliance with
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NUREG-0737 commitments.
Craftsmen appeared knowledgeable of work controls and standard practices.
The cogniza.nt engineers were able
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to answer all the inspector's questions on design torquing, valve
testing, and visual examination criteria. Coordination of the field
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work with control room activities was evident from the standpoint of
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tagging controls, planning for specific train and subsystem outage
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l times, and operator cognizance of field activities.
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The inspector also witnessed a demonstration of strain gauge testing
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of Limitorque valve operator stroke settings.
This technique is
being developed by the licensee to replace MOVATS testing of motor-i l.-
operated valves and appeared to be in compliance with the guidance
and requirements of NRC Bulletin 85-03, NRC Generic Letter 89-10, and
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ASME Code Section XI. The engineering principles serving as bases l
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for the use of strain gauge testing were explained by technical sup-
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l port personnel and the status of allowable rant,es of thrust ca7tula-tions was discussed. The inspector concluded that the licensee's
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design of this program exemplified initiative in d*veloping a test uniquely fitted to the problems and applications of motor-operated
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valves at Seabrook. Also, this development work was well thec to
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comply with the Generic Letter 89-10 mandated reviews.
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With regard to all of the above maintenance and design change activ't'es discussed above, the inspector identified no violations and had no un-resolved safety concerns. -The licensee's routine program of controls
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appeared effective and properly implemented.
4.2 Surveillance Activities The inspector witnessed a portion of train 'B' diesel generator f
monthly operability surveillance (0X1426.05) in the control room and
inspected the final positions of certain diesel generator jacket
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water and fuel oil and service water valves to ensure operability
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after test and required rework completion. One problem was noted
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during this test: the train 'B' fuel oil transfer pump, DG-P-38B,
failed to operate in the automatic mode and transfer fuel oil from
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the train 'B' storage tank to its respective day tank. Man'ual opera-i tion of the pump was possible, However, in order to repair the auto-
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matic control problem, DG-P-388 was tagged out of service and i
electrical supply breaker opened.
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Technical Specification 3.8.1.2.b requires one operable diesel gene-rator with an operable fuel transfer pump in Mode 5.
The train 'A'
diesel generator was operable at this time. The operators then lined up the train 'A' fuel transfer pump (OG-P-38A) to supply the train
'B' day tank through cross-connected piping, as allowed and discussed in FSAR section 9.5.4.3.
Approximately one-half hour after setting i
up the cross connection, the train ' A' diesel generator was taken out
.i of service for maintinance on its electrical bus.
However, the
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safety evaluation in FSAR section 9.5.4.3 states: "all the motor-
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driven pumps are powered from the bus on which the diesel generator it serves is connected." The system lineup directed by the operators
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was acceptable only to the extent that offsite power was available
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since, with the train 'A' electrical bus 5 out of service, no emer-
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gency power supply was available to DG-P-38A to keep the train 'B'
diesel generator day tank supplied with fuel.
Thus, even with the cross-connected fuel oil lineup, the train
'B' diesel generator was inoperable in accordance with the provisions of Technical Specifica-
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tion 3.8.1.2.b and the Action Statement should have been entered.
All the Action Statement requirements required by the subject lit. N ing condition for operation were in fact being met by existing plant
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conditions.
Additionally, a quality assurance surveillance of this operational activity was in progress and concurrently documented the
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h NRC inspector's concern regarding diesel generator operability in i
surveillance report QASR 89-00733. A subsequent QA engineering L
evaluat fon of this situation confirmed that the subject Action State-
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ment shovid have been entered.
The inspector discussed this evolution with Operations management who corroborated the position that the operators' action had not bean
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prudent.
No adverse safety impact was evident since electrical bus 5, and thus the trein
'A' diesel generator, could have been restored to operable status in a short period of time, certainly before the train 'B' day tank would have appruached low level conditions upon L
The licensee is currently developing a Tech-l nical Clarification to be issued as guidance to the operators to pre-
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clude' recurrence of similar situations. The inspector raised some additional questions regarding assumptions made about fuel oil stor-age tank level and fuel filling / settling times in the safety evalu-ation of FSAR section 9.5.4.3.
The licensee responded with appro-
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priate revisions to the train 'A'
and 'B' diesel generator fuel oil l
system operating procedures, 0S1026.05 and 051026.13.
l The inspector had no further questions and considered both the opera-tions department actions appropriate to the concerns and the QA de-
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partment involvement as evidence of an active quality program, i
i Other surveillance activities checked by the inspector include a re-
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view of the weekly diesel generator train 'A'
& 'B' surveillance report
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(procedure 051426.12), and discussion with operators or, shift re-
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garding the cond>ct and independent inspection of the affected valves; a review of the most recent monthly surveillance (procedure 0X1401.11) of the pressurizer vent valve, RC-V-468, and independent
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inspections of valve position and locking device by the inspector
during both August and September, 1989; and examination of the pres-
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surizer safety valve surveillance procedure (EX1804.044) with respect t
to both Technical Specification and ASME Section XI in-service test-
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ing requirements.
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With respect to the latter inspection item, the inspector reviewed a
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i QA surveillance report (QASR 89-00216) documenting review of proce-l
dure EX1804.044 to the guidance and criteria provided by NRC Infor-
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mation Notice (IN) 88-68 on setpoint testing of the pressurizer
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safety valves. No problems were identified by this QA surveillance.
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The inspector noted that EX1804.044 allows the pressurizer safety valves to be setpoint tested in place with pneumatic assir.t equip-u I,
ment. However, discussion with licensee technical support personnel L
revealed their plans not to use this option, but rather to send the
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l valves off-site to have setpoint pressure tests perforraed. The in-
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spector confirmed that, in addition to the three intt011ed pressurizer
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safety valves, five spare valves are available for Use, indicating
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ing to adjust setpoints in place using the pneumatic as:ist equip-
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ment.
Thus, the concerns raised by IN 88-68 appehr to have been
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satisfactortly addressed by the licensee's setpoint testing program.
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The inspector has no concern regarding the current operability of the three pressurizer safety valves, since' records indicated that they were properly steam set at the factory by the Crosby Valve & Gage Company.
- The inspector has no questions regarding the scope, scheduling or conduct of the licensee's surveillance activities. No violations were identified and the overall conduct of the surveillances that were examined appear to be well controlled.
5.
Radiological Controls 5.1 General The inspector observed routine access controls into the radiologic-ally controlled area (RCA) of the plant and noted proper temporary radiological posting and controls at areas where the RCA had been opened for maintenance and outage work. The inspector spot-checked the monitoring controls for tools and major equipment egressing the RCA by health physics personnel.
Personal item scanning techniques were discussed with health physics personnel at the HP checkpoint.
The inspector reviewed a radiological occurrence report (ROR) 89-17 documenting the inadvertent presence and possible exposure of one licensee employee during radicgraphy within the plant. Analysis of the individual's dosimetry and calculation of the maximum dose to which the individual could have been exposed (i.e., approximately 2 mR) revealed no overexposure conditions.
This was verified by duplicating the conditions in question with a dosimeter placed where
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the individual was located.
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The root cause of this incident was identified to be a failure on the I
part of the radiographers to conduct a three-dimensional search of the area where the radiography was to take place. The individual was positioned on a platform above the shot prior to the radiographers'
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L establishment and posting of the high radiation area.
The contract
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L radiographers were counseled by the Health Physics Department Super-
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visor and three dimensional searches of radiography zones, to include
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verbal attempts to identify personnel in the area, were emphasized.
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The inspector discussed with HP personnel the posting of radiography l,
areas in general, and specifically checked how the RHR vault would be controlled, given the unshielded height, and access ladders and
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doors into that area. No problems were identified.
The corrective l
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action to this incident appeared appropriate to both root cause and i
radiological concerns, i
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The inspector had no further questions on this issue.
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5.2 Investigation of a Health physics Technician
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I On September 13, 1989 the inspector was infqrmed by licensee manage-
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ment of the investigation and subsequent resignation of a health
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physics technician, based upon evidence that the individual submitted l
a documented report of a radiological survey that actually had not
been performed.
The licensee's investigation included a review of l
all work accomplished by the subject employee since he began work
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with New Hampshire Yankee at Seabrook Station.
No Technical Spect-i fication surveillance errors, unexplained radiation monitor setpoint
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anomalies, or high radiation door control problems were identified.
The licensee's review concluded that this inoividual's questionable
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performance raised no Technical Specification compliance questions s
and had no adverse impact on public health or worker safety, i
Since this represents the second recent incident (reference: inspec-i tion report 50-443/89-08, section 6.2) of this nature, the licensee l
conducted a root cause analysis of both cases. Other investigative
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and corrective action undertaken by the licensee included interviews with the station health physics technicians and evaluation of other
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health physics work activities, briefings of the station managers, l
conduct of an independent cause and effect analysis, and development
of a performance monitoring program to assure the detection of simi-
lar occurrences.
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The inspector reviewed the licensee's action plan for implementing
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l the above items.
The overall plan appeared well directed. While no
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adverse safety impact or violations of license conditions or licensee
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management inadequacy were identified in these cases, both incidents i
emphasize the need for ongoing independent verification and manage-
I ment monitoring activities commensurate with the safetyrelated im-
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I portance of the work being performed.
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The inspector had no further questions on this issue.
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Security I
i The inspector conducted a review of the licensee's Security Log for the second and third quarters of calendar year 1989.
Entries were evaluated against the criteria specified in NUREG-1304 and Regulatory Guide 5.62.
Evidence of an active chemical screening program in accordance with the
NHY Fitness for Duty policy was noted.
In t5e latter regard, the inspec-tor also observed drug detection dog search activities within the pro-tected are _
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The inspector noted adequate guard posting for compensation of inoperable
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or open vital area security doors.
In one case of guard response, where personnel exited a specific security zone and the door failed to close be-l cause of differential pressuros, the inspector observed a guard respond c
quickly.
Signs were posted at that door to alert personnel to door clos-
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ure problems, a work request was issued for evaluation and repair, and a guard was subsequently assigned to attend to the door as a compensatory
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measure.
In another example, the inspector witnessed licensee personnel action in regard to an inoperable card reader. While two personnel
ignored the failed light, a third licensee individual brought the problem
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to the attention of health physics and security personnel and a sign was posted until the card reader was repaired.
In this case, the subject door
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represented neither a HP nor a security door problem, but the card reader had been installed for personnel accountability and was intended to be
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relocated in accordance with a design change.
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Additionally, the handling and physical security of safeguards information
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was discussed with security department personnel. One questionable inci-t dent of control was investigated, logged and properly handled. The in-
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spector raised another question of a safeguards information nature regard-ing a purtion of the protected area fencing.
This issue is currently
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being researched by the licensee and NRC headquarters personnel and does i
not currently represent a deviation from Security Plant requirements.
- The inspector found the licensee security staff responsive to any ques-
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tions or areas of concern.
Overall, the security program was effectively implemented.
7.
Quality Verification Activities
7.1 Audits. Trending & Corrective Action Followup (RI-89-A-112)
The inspector reviewed the Joint Utility Management Audit (JUMA) re-
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port documenting the results of the August 1989 review and evaluation
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of QA/QC activities at Seabrook Station by an audit team comprised of members from other utilities. The Quality Trend Analysis Report, covering the first half of 1989, was also reviewed and evaluated in I
light of the recent NRC assessment of Seabrook activities (reference:
SALP Report, 50-443/87-99).
Additionally, internal licensee audit reports and compliance reviews (e.g., GE HFA and Agastat relay in-
vestigations) were spot-checked for the scope and completeness of QA
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inspection coverage.
Overall, an active QA program with involvement in several technical issues and all plant disciplines was in evi-dence. QA trending of problem areas correlated well with NRC find-ings, and attempts to implement corrective measures had been initi-ated by licensee management through procedural compliance training,
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procedure consistency reviews, and organizational goals for personnel commitments to excellence.
The effectiveness of QA corrective ac-tions and management initiatives for self-identified problems will remain an area for future routine NRC inspection in accordance with inspection program provisions.
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i In regard to a specific Corrective Action Request, CAR 89-0008 (RI-89-A-112), the inspector conducted a follow-up inspection of the
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identified deficiency.
In this particular case, a personal computer (PC) pro.*am was being utilized to sum the individual reactor con-s
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tainment penetration leakage results, as measured and calculated tv l
the Type 'B'
and 'C' leak rate testing performed in accordance with I
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engineering surveillance procedure, EX1803.003.
The inspector re-viewed and evaluated the procedure to the criteria of 10 CFR 50, L
Appendix J.
Technical support personnel were interviewed regarding the conduct of individual penetration test measurements.
The result-
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ing calculations and review against acceptance criteria were done by i
hand and verified. Only the summation of the individual results and
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comparison to total allowable leakage criteria was performed utiliz-
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L ing the PC program, which had not been independently verified in ac-cordance'with 10 CFR 50, Appendix B.
At the request of the inspec-
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tor, the licensee performed hand calculations to validate the ac-l curacy of the subject PC program.
The results of these calculations, to include consideration of standard deviations associated with the
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individual measurement errors, revealed an accuracy of the software
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in question to two decimal points.
The inspector also checked the adequacy of licensee access and control of the software package and
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identified no technical problems.
In the case of specific example
cited by CAR 89-0008, no technical deficiencies, errors, or safety-related concerns were identified.
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However, the question of generic software controls for PC and/or
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station computer usage in the performance of safety-related activi-
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ties is a valid QA issue. A memorandum from the Nuclear Quality Manager to the Information Resource Group and Computer Systems Man-l i
agers documents the need for additional computer software quality a
configuration management and controls. The inspector has been in-formed of the licensee's intention to address this self-identified
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concern from an overall NHY programmatic approach. The inspector had
no questions regarding licensee actions on this issue to date and,
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as prescribed by the NRC inspection program, will follow future ac-
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tivities during routine inspections.
7.2 Field Assembly Inspection & Records Followup J
The inspector examined structural supports located on the refueling
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deck inside containment.
These structural saddles serve as the sup-s port for the containment equipment hatch wnen it is removed from position in the containment wall.
1he support assemblies are cur-rently anchored to the containment liner wall by high-strength (ASTM A325) bolts.
The inspector noted, however, that although long
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slotted holes had been utilized in the bolted assembly, no plate l
washers had been installed in accordance with the AISC Specification
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for structural jo w s sing ASTM A525 bolts. While the subject sup-I
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m port assembly does not have a safety-related function, USNRC Regu-Es, latory Guide 1.3 alls for it to be se!smically designed and for the
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appropriate 10 CFR 50, Appendix B criteria to be applied to its
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. fabrication and installation.
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However, upon the request.of the inspector, the licensee could. pro-
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E duce no quality assurance records documenting inspection of this sup-
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W-port. Discussion with licensee engineers indicated that the lack of
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the required plate washers appeared to be'an' isolated oversight...No
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other assemblies were found with a.similar deficiency and, in, fact,
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the inspector noted ongoing field work which required slotted holes
@o in structural bolted connections and for which plate washers had been L
suppl $ed. A work request (89X004667) was issued to replace the bolts
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and install the missing washers. An additional work request
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(89W004850) requires that an inspection of'the welds and dimensional
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checks of the equipment hatch saddles be performed per engineering i
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recommendations.
This latter inspection is intended to confirm
acceptable fabrication and provide quality documentation of the con-y^
struction work in accordance with 10 CFR 50, Appendix B.
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Ya The inspector reviewed the noted work requests and discussed licensee
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F actions on this issue with both QA and engirmering personnel.
He had
no further questions and considered the licensee corrective measures
appropriate to the safety significance of the item.
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8.
Re 4ew of Licensee Reports 8.1 Licensee Event Reports (LER)
(Closed) LER No. 89-010: ESF Actuation - Diesc1 Generator Start, s
This spurious event was discussed in section 6.4 of Region'I Inspec-K tion Report 50-443/89-08. The inspector reviewed the LER which con-firmed the previous determination of the root cause as personnel-
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I error.
The undervoltage signal generated for the emergency bus, re-K sulting from the compartment door opening, caused the spurious diesel generator starts. Emergency diesel ger.erator IB started as required
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G and all safety systems functioned as designed.
Corrective action
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D included additional labeling on the subject fuse compartments, warn-T ing that opening of the potential transformer fuse cubicle doors re-f sults in deenergization of the associated electrical bus. Additional IL training of all auxiliary operators regarding proper potential trans-l0 former fuse operations is also plannea.
L, The inspector had no questions regarding the cause or significance of
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this event and no concern regarding the adequacy of licensee correc-
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tive actions.
This issue is closed.
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U (Closed) LER No. 89-011: Unsealed Penetrations in the Condensate
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Storage Tank Enclosure. On September 5, 1989, a four hour notifi-
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cation was made-in accordance with 10 CFR 50.72(b)(2)(1) to inform the NRC that evidence was not available to confirm the-fact that all j
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penetrations through the condensate storagt tank (CST) enclosure were
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watertight. This enclosure is a concrete structure around the CST
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'which, as discussed in FSAR sections 3.8.4.1.g and 9.2.6.3, is in-
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. tended'to be " capable of retaining the contents of the tank should.
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the tank rupture."
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On October 5, 1989, the licensee submitted an LER discussing this
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1 deficiency and confirming three unsealed piping. penetrations in the
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CST enclosure.
Thus, during two previous periods of time, when the
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plant was at or above mode 3 for testing, the limiting condition for
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operation of Technical Specification 3.7.1.3 requiring "a concrete CST enclosure that is capable of retaining 212,000 ge11ons of water" i
had been violated..The root cause of this condition is being in-
vestigated and a supplemental LER is to be submitted by the lic6nsee.
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?g Corrective actions include sealing the penetrations prior to future entry into mode 3.
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t Although a supplerental report is due on this event, this LER is con-sidered administratively closed.
Licensee root cause analysis and corrective actions are considered an unresolved item (89-09-01).
Licensing issues which relate to this event will also be tracked with
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this item.
Pending completion of corrective actions by the licensee,
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further assessment by the NRC, and uaderstanding of the complete de-
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sign and operability considerations, this issue remains unresolved.
- s (0 pen) Potential LER: Failure to perform Technical Specification Sur-l veillances. On October 3,1989, the licensee identified the fact that certain radioactive liquid effluent and gaseous effluent moni-
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toring instrumentation surveillances had not been performed as re-quired by Technical Specification 3/4.3.3.9 and 3/4.3.3.10 in the time intervals prescribed.
As documented in Station Information Re-
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port (SIR)89-061, as of September 27, 1989, the Technical Specif1-
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cation operability requirements for the subject instruments to be in M
service "at all times" were not satisfied. The licensee is reviewing j
whether the Action Statement requirements were being complied with P'
coincidentally when the limiting conditions for operation were p
violated.
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The licensee conducted and completed satisfactorily the required sur-D ve111ance tests on October 3, 1989.
Reporting in accordance 10 CFR L
50.73 is planned. This item remains unresolved until the licensee's p
evaluation is complete and is evaluated by the NRC (89-09-03).
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8.2. Part 21 Reports (Closed) 10 CFR 21 Report 88-00-01:.Limotorque Motor Operator' Worm
, Gear Defects.. On. March 18, 1988, Limitorque Corporation filed a 10 (
-CFR 21 notification concerning potential defects in the worm gear-
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castings of their Type-H3BC valve actuators.
The licensee identified
~two. valves in the service water system, SW-V-4 & 5, as having H3BC actuators.. Work requests were issued.to. disassemble the two affected valves and inspect'the subject worm gears.
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The inspector reviewed the completed work request'(88W001764) for valve SW-V-4 and inspected the valve after reinsta11ation in the plant.
The.' disassembly of SW-V-5 was inspected in progress and the-
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worm gear was visually examined by the inspector.
The-work request (88W001765) for SW-V-5 was reviewed as was the referenced maintenance
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. procedure MS0519.66. The inspector discussed work controls with the
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maintenance: technicians, observed.QC' inspector availability and hold point assignments, and was provided engineering response'to questions regarding.the correct' acceptance criteria to be used.in the worm ~ gear
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examination.
In the-case'of both valves, no defects were found.
The inspector has no further questions regarding licensee response and action to the 10 CFR 21 report. This item is closed.
(0 pen) Engineering Evaluation No. 89-026: Atmospheric Steam Dump Valve-(ASDV)' Failure Evaluation for Reportability in accordance with
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On April 5, 1989, USNRC Information Notice No. 89-38 was issued to describe the failure of'ASDVs at the Palo Verde units.
Subsequently. on June' 21, 1989, Control Components Inc. (CCI), the
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ASDV manufacturer for both the Palo Verde and Seabrook ASDVs, issued a letter to Seabrook indicating a potential problem with the Seabrook ASDVs based upon the Palo Verde incidents.
Although the specific
cause of the Palo Verde failures was not determined, the connection
to the Seabrook valves was the fact that significant piston ring leakage would cause the ASDVs to fail to open upon demand.
The inspector reviewed NHY Engineering Evaluation No.89-026, which
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concluded that the identified ASOV problems were not a substantial safety hazard reportable under 10 CFR 21. Although it was agreed
!
that significant differences existed between the Palo Verde site, wFere the problems occurred, and the conditions and performance his-tory of the Seabrook ASDVs, the inspector questioned the licensee
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analysis in four specific areas:
(1) assumption that the failures are random ano therefore will not occur to more than one valve at the same time.
(2)
reference to a probabilistic analysis in evaluating a deter-ministic problem.
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(3); reliance upon operator action to manually open a failed ASDV.
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'(4) the decision not.to implement the CCI recommendations for ASDV I
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The licensee indicated that a monitoring program would be implemented J
to evaluate the ASDV performance over time and that' additional test-l
.ing would be conducted to verify operability.under mode 3 conditions.
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While no evidence: exists to suggest that the Seabrook ASDVs have a
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design defect that requires repair, the notification from the ASDV
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vendor suggests enough~ doubt that the proposed licensee actions on
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this issue should be formally cubmitted to the NRC.
Pending sub-
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mittal by the~ licensee of their plan'to address the issues raised by
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Information Notice No. 89-38 and the notification of potential. prob-
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lems by CCI, this issue is ' unresolved (89-09-02). ' The correctness of
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licensee nonreportability under 10 CFR 21 is not in question here, t
but the acceptability of the proposal to address this potential prob-
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lem is an issue that merits formal licensee documentation of.its position and review by the NRC.
l 8.3 Reportable Events at Other plants
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The inspector examined licensee engineering reviews of two reportable
events at other nuclear sites. The first involved electrical faults
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propagating back through the supply system and causing a spurious
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trip of the electrical feeder breaker. At Seabrook Station, elec-
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trical breaker calculations have been: performed to ensure proper co-t ordination between the breakers, fuses.and loads in the electrical
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supply system. Thus, the load protection is designed to' prevent one
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fault from improperly taking down an entire supply circuit.
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In the second case, another plant reported that its pressurizer auxiliary spray line had not been analyzed for temperatures it may be subjected to under conditions of maximum letdown, minimum charging
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and no reactor coolant pump operation. At Seabrook Station, the piping downstream of the regenerative heat exchanger, to include the pressurizer auxiliary spray line, is designed to withstand tempera-
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tures in excess of the most adverse operating conditions, including those associated with maximum letdown and minimum cooling.
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P The inspector reviewed the licensee follow-up on the design questions
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raised by the two noted events at other plants.
The Seabrook-
.7 specific component design criteria were spot-checked.
No unresolved safety issues were identified and the inspector had no further ques-tions.
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9.
Technical Review of Modifications E
9.1 Contro1' Building Air (CBA) System.
As'a result of-a deviation from FSAR commitments in the design'of the-CBA system identified in 1986, and a subsequent violation with re-spect to maintaining the control room at a positive differential pressure, the licensee committed to provide'for NRC-staff review the details of modifications to the CBA system.
In a letter dated-January 22, 1988, the licensee described the proposed CBA modifica-o n
tions.and in another letter, dated March 30, 1989, stated that the j
j-work would be comp.leted by September 30, 1989.
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During this inspection, a project engineer from the NRC Office of s
Nuclear Reactor Regulation (NRR).eumined the modified CBA system.
The NRR engineer found that the proposed additiorial HEPA filter F-8038 and the proposed bypass piping with two back draft darapers were
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installed. The original two purge lines were capped off and their
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associated purge valves removed.
In the report of the " Control Room Area Ventilatior System 18-Month Surveillance" dated September 29, 1989, NRC review found that the flow balance of the tuodified CBA cystem is within the acceptance criteria of the Seabrook Technical
_
Specifications, 1100CFM +/- 10%. ' Based on the September 29, 1989
'
surveillance report, emergency filter train ' A',
CBA-F-38 had a total flow of 1193 CFM which consisted of 573 CFM make up air and 620 CFM recirculation air. Emergency filter train
'B', CBA-F-8038 had a total flow of 1173 CFM which consisted of 579 CFM makeup air and 594 CFM
. recirculation air. These test results indicated that the makeup air was within the design value of less than or equal to 600 CFM.
The control room maintains a positive differential pressure (dp) to l
its adjacent areas during normal and emergency operation modes. The l
J surveillance test indicated that the Contrcl Room Area to outside dp
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was 0.15 inches water gage (WG) and the Control Room Area to Cable Spreading Room DP was 0.16 inches WG.
These are greater than the i
Technical Specification requirement of 0.125 inches WG.
I On October 3, 1989, the control room was in the emergency mode of operation due to an electrical bus E6 outage for maintenance. The NRR engineer observed that tne emergency filtration train operation maintained positive control room dp.
Based on the above findings, this inspection concluded that the im-piemented modifications on the control room resulted in an acceptable
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L surveillance test and, therefore, these modifications are acceptable with respect to licensee commitments for the CBA design modification.
Additionally, because of the ongoing CBA modifications, the Action l'
Statement requirements of Technical Specification 3.7.6 were imposed
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upon plant operations. The senior resident inspector (SRI) reviewed l
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licensee adherence to the action requirements, to include the licen-
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see's controls to ensure no positive reactivity changes. Technical
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.
Clarification TS-098 was issued on August 1,1989 to provide inter-pretation of boron concentration limits.and temperature controls in line with the intent of the' Action Statement to maintain the plant in
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a-suitable reactivity condition.
The inspector had no questions re-garding the implementation of reactivity controls.
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Based upon both SRI inspection and NRR project engineer review, the
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modification to the CBA system appears to have' been controlled and
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implemented in accordance with design commitments. Operations and licensing staff cognizance of the. impact upon routine operational activities and the requisite work controls was~ evident and'no con-
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cerns or unresolved safety questions were identified.
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9.2-Safety Parameter Display System (SPDS)
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Supplement No. 7 to NUREG-0896, the Seabrook Safety Evaluation Report
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.(SSER 7), was issued in October of 1987.
Paragraph 18.2 of SSER 7
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describes the March 25, 1987, Partial Initial Decision of the Atomic i
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Safety and' Licensing Board (ASLB) as it related to the Seabrook
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safety' parameter display system (SPDS).
Several of the ASLB issues involve staff verificatior prior to Seabrook operation above five i
. percent power or before startup following the first refueling outage.
NRC Region I Inspection Reports 50-443/87-16 & 88-15 describe pre-vious NRC inspection of certain SPDS issues.
Below are listed addi-
tional SPDS issues reviewed by an NRR Project Engineer during this inspection.
(1) Board Order 2(b).
For this SPDS item, the licensee created an
additional Emergency Coolant Recirculation Status Tree Logic in
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its SPDS program.
The NRR engineer observed this Status Tree Logic on a plant computer. screen display which showed the de-
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signed and actual residual heat removal flow in GPM.
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Hydrogen concentration can be called out onto the computer screen display on the containment status tree when the hydrogen monitor is functional. When the monitor is not functional, the display will show a black indication of the terminus exiting the hydrogen concentration logic box.
(2) Board Finding 35.
Summary status of containment isolation can
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be called out on the computer screen display for the Critical
Safety Function (CSF) Summary.
Isolation valves are grouped
,
into phase A and phase B.
Individual valve position of each I
containment isolation valve is indicated on the screen in red for the open position, and in green for the closed position.
Operators can easily determine, by pattern recognition from the assigned SPDS location, the overall status of containment isola-tion.
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'(3). Boa'rd Findings'40/41/42. The data validation algorithm was im-
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Vj This new range checking is more representative of the operating
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A range of the instrument.
Data not within the expected range.
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will not'be included in the final average.' Also,'a bandwidth check is added'to the software to prevent.the SPDS from making a
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.dctermination unless at least two reliablejredundant measure-f
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ments agree within a preselected tolerance..
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For hydrogen concentration instrument readings, licensee con-
$1 trols provide a programmatic check of whether: (a) the hydrogen monitor is "on"; (b) the monitor is' not failed;.and (c) the con-
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centration is' reliable. When the above three conditions are
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being. met, then the reading is an acceptable concentration'
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1-The NRR project engineer.had no further questions regarding the de-sign and implementation of the above SPDS items.
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Management Meetings
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At periodic intervals during the course of this inspection, meetings were
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' nspection. An exit meeting was conducted on October 10, 1989,. to discuss
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- the inspection findings during the period.
During this' inspection, the
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NRC inspector received no comments from the licensee that any of their
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inspection items or issues contained proprietary information. No written
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material, except for publicly available NRC Headquarters Daily Reports of
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.the types discussed in section 8.3 of this report, was provided to the
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