IR 05000456/1987007
| ML20215H831 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 04/28/1987 |
| From: | Kropp W, Little W, Thomas Taylor, Tongue T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20215H818 | List: |
| References | |
| 50-456-87-07, 50-456-87-7, 50-457-87-06, 50-457-87-6, NUDOCS 8705070086 | |
| Download: ML20215H831 (40) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION III
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Reports No. 50-456/87007(DRP); 50-457/87006(DRP)
Docket Nos. 50-456; 50-457 Licenses No. NPF-59; CPPR-133
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' Licensee: Commonwealth Edison Company Post Office Box 767 Chicago, IL 60690 Facility Name: Braidwood Station, Units 1 and 2 Inspection At: Braidwood Site,.Braidwood, Illinois Inspection Conducted: -February 15 through April 11, 1987 Inspectors: Nag
'Jfha T. M. Ton!)ue 9/thk?
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.(2hdhOr W. J. Krop) 4[h7 Date f h, #
T. E. Taylob 4/28 7 Date EG&G Idaho, In J. B. Moncur K. Phillips R. Storck -
Approved By: . i 'e[ '
Brai wood Project or [ 7 yef D
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Inspection Summary Inspection on February 15 through April 11, 1987 (Reports No. 50-456/87007(DRP);
No. 50-457/87606(DRP))
Areas Inspected: Routine,. unannounced safety inspection of activities with regard to licensee action on previously. identified items, licensee event reports review; exceeding technical specification action statements for fire protection; strike by contractor employees; regional request; Chairman Zech and Commissioner Asselstine onsite visits; electrical equipment installation -
Unit 1; electrical cable tray hangers - Unit 2; piping - Unit 2; pipe supports -
Unit 2; operator awareness; out-of-service; auxiliary feedwater; procedure review; maintenance; startup test witnessing and observation; operational safety verification; Unit 2 plant tour; report review; and meetings, training, and other activitie Results: Of the twenty areas inspected, no violations were found in seventeen areas, three violations were found in three areas (AV0s, technical specification action statements, and lubrication - Paragraphs 2.b. , 4,13.c. , and 17),
respectively. In addition, concerns were addressed in four areas (0perator Awareness - Paragraph 13, 0ut of Service Process - Paragraph 14, Auxiliary Feedwater System complying with the FSAR - Paragraph 15, and Procedure Reviews -
Paragraph 16).
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DETAILS 1. Persons Contacted Corporate Personnel 18 . Thomas, Executive Vice President 1C. Reed, Vice President, Nuclear Operations 1T. Maiman, Vice President, Projects 1N. Wandke, Assistant Vice President, Nuclear Services 1L. DelGeorge, Assistant Vice President, Engineering and Licensing 1D. Farrar, Director,' Nuclear Licensing 1W. Shewski, Manager, Quality Assurance 1S. Hunsader, Nuclear Licensing Administrator 1L. Gerner, Superintendent, Regulatory Assurance 18. Stephenson, Manager, Department of Nuclear Safety Braidwood Personnel
M. J. Wallace, Project Manager 1,2E . E. Fitzpatrick, Station Manager W. Vahle, Construction Superintendent 1,2C. W. Schroeder, Station Services Superintendent 2L. E. Davis, Assistant Superintendent, Technical Services 2K. Kofron, Production Superintendent C. Tomashek, Project Startup Superintendent B. Byers, Assistant Construction Superintendent 2M Lohman,ProjectStartupSuperintendent 2P. Cretens, Station Startup Assistant Superintendent 2F . Willaford, Security Administrator 2D Paquette, Maintenance Assistant Superintendent 20. O'Brien, Operations Assistant Superintendent i
E. L. Martin, Quality Assurance Superintendent 1,2 D L. Shamblin, Assistant Project Manager 2R. Benn, Assistant Security Administrator G. E. Groth, Project field Engineering Manager 1,2P. Barnes, Regulatory Assurance Supervisor 2M. Takaki, Quality Control Supervisor 2R. E. Aker, Radiation / Chemistry Supervisor J. Jasnoz, Tech Staff AR/PR Coordinator 2R Lemke, Technical Staff Supervisor G. Nelson, Projects Supervisor T. W. Simpkin, Operating Experience Group 2R. C. Bedford, Regulatory Assurance-R. Mertogul, Systems Supervisor 2R. Kyrouac, Quality Assurance Supervisor 2E. Wendorf, Lead Electrical Supervisor
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2L.'Kline, Regulatory Assurance Industry Grou M R. Dougherty, Project Construction Department 2L.' W. Raney, Nuclear Safety
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2P. Zolan, Construction Quality Assurance 2T. Bobic, Maintenance 2A. J. D' Antonio, Quality Control 2T. J. Lewis, Startup 2J. Gosnell, Quality Control 2J. D. Dierbeck, PCD Mechanical Lead 2G. Nelson, TSS Assistant 2N. Coan, Office Supervisor 2R.'J. Ungeran, Operating Engineer 2R.' Yungk, Operating Engineer 2R. J. Legner, Operations 2S. Hedden, Instrument Maintenance 2G. Schieberl, Training
, 2R. Hoffman, Mechanical Maintenance
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NRR 1S. Varga 1J. Stevens 1 Denotes those attending the SALP Meeting held at the Region III Office in Glen Ellyn, Illinois on March 17, 198 Denotes those attending one or more exit interviews conducted on March 5, 26, April 2 and 9, 1987 and informally at'various times throughout the inspection perio The inspectors also talked with and interviewed several other license employees, including members of the technical and engineering staffs,
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startup engineers, reactor and auxiliary operators, shift engineers and foremen, electrical, mechanical and instrument personnel, contract i
security personnel, and construction personne . Licensee Action on Previously Identified Items Open Items
- (Closed) 456/84009-08; 457/84009-08: The number of_ outstanding i Field Change Requests (FCRs) with only verbal approval and not documented written approval appear to exceed acceptable quality practices, considering the possibility of verbal miscommunication The large number of outstanding FCRs with only verbal approvals were due to a policy in effect at the time. This policy stated that outstanding piping support FCRs would only be incorporated into the i- drawing and the drawing released by Sargent & Lundy (S&L) afterja complete installation verification was performed. This policy was
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changed on March 16, 1987, to require the mechanical contractor to complete the piping support installation within six months from the-specific FCR release date. TheinspectorreviewedS&LProject Instruction, 8B-13, which defines the use of verbal FCRs. The inspector also reviewed the status of FCRs presently identified outstanding. There did not appear to be an inordinate number of verbal FCRs. This matter is considered close (Closed) 456/84042-01; 457/84038-01: Identifying design changes that effect the FSAR and the subsequent method of updating the FSA The inspector determined that the FSAR is updated periodically-through the combined effort of the licensee's Project Engineering Department and Sargent & Lundy (S&L). S&L responsibilities and th methods for updating the FSAR are defined in S&L Project Instruction PI-BD-56, " Review of Design Changes for Effects on Radiological Release Paths FSAR and SRCL." This matter is considered close (Closed) 456/85007-01; 457/85007-01: Avoid Verbal Orders (AV0) were-utilized by L.K. Comstock (LKC) to instruct the electrical craft to remove, replace, or rework electrical components. Also LKC identified deficiencies in their drawing control program which.may have resulted in some installations and inspections being performed to a superseded drawing revision. The portion of this open item pertaining to AV0s is directly related to Unresolved Items 456/84006-01; 457/84006-01, which is subsequently evaluated and closed in this report. In regards to the drawing control progra deficiencies identified, LKC issued Procedure 4.3.2.4, " Rework,"
on March 29, 1984, which described the control and tracking of rework of electrical components caused by a revision to design documents or interference problems. To address the drawing control revisions prior to March 29, 1984, LKC issued Procedure 4.13.3,
" Area Completion / Turnover." The program defined in this procedure was developed and implemented to ensure that safety-related electrical components have been inspected to the latest drawing revision which affected item. The area turnovers have been completed for a major
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portion of Unit 1. To evaluate the effectiveness of LKC Procedure-4.13.3, the inspector selected conduit hangers, cable trays, cable tray hangers, and junction boxes in those Unit 1 areas which had been complete The results of these inspections are documented i in Paragraph 9 of this report. No problems were noted with the hardware or documentation.
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Based on the results of these inspections, it appears the licensee and LKC are effectively implementing LKC Procedure 4.1 Therefore, this matter is considered close (Closed) 456/85053-06; 457/85051-01: The inspector requested identification and evaluation of three items pertaining to the fire seal installations:
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(1) The inspector requested that the licensee develop and specify to craftsmen, appropriate mix proportions of accelerato solutions to be added to the gypsum slurry (particularly-large surface areas).
-(2) The inspector requested the licensee to identify and evaluate the fire resistivity of all fire barrier penetration seal configurations exceeding the dimensions of those configurations supported by the Transco Products, Inc. Fire Test dat Additionally, the licensee was requested to identify and evaluate all fire barrier penetration seals where the fire barrier itself has less depth than the tested fire code CT Gypsum and Thermofiber CT felt penetration seal assembly (floors or walls that have a depth of less than nine. inches thick).
(3) The inspector requested the licensee to evaluate fire protection cable density fill criteria and document onsite-corrective actions taken where necessar CECO has provided the following responses, tests, and details for the above three items:
(1) Transco Products, Inc. has provided to CECO ~ procedure "Special Process - Fire code CT Gypsum Cement," PSQAP 9.1 BRi Revision 5~,
Section 9.3.10(d). It identifies the appropriate mix proportions of accelerator for the gypsum slurry mix. This is the procedure used on all gypsum cement penetration fills. (Approximately 86%
of the penetrations for Unit 1 are filled and 50% for Unit 2.)
(2) For seals installed in openings larger than the.overall tested configurations, specific details have been developed for use of (BR-E-20) " Ceramic Board Dividers for Oversized CT Gypsum Seals" Transco Products, Inc. This detail uses ceramic board type dividers to section the large openings into components which are within tested dimensions for sealing material Transco Test Report No. TR-109 supports the use of dividers in penetration seals. (The tested configuration was 32" x 109" in the control room floor. --There are approximately 20 penetrations larger than the test configuration. They are approximately l' x 18' with the ceramic board dividers-every 2'.)
Where the floor or wall depth is less than 9" Transco Detail BR-E-10, "CT Gypsum with Sheet Metal Collar for Fire / Air Seals at Cable Tray / Cable Openings in Floors" is used. The primary use has been in the upper cable spreading rooms. This detail 6-
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has been successfully tested for uses as a three hour rated fire barrier, and is documented by Transco Fire Test Report N TR-22 er cable spreadin room are provided wit 3 this configuration.All openings in the floor =of of the openings in the upper cable spreading room have been filled.)
(3) All penetration seal installation details used for fire seals are reviewed and approved by Sargent & Lundy and M&M Protection Co. as well as data from fire test. reports that are applicable to the fill configuration Transco Test No. TR-198 supports the use of cable fill densities up to 64.3%, and Transco Test No. TR-220 supports cable fill densities up to 46%.
Approximately 20% of the penetrations at Braidwood will have a fill density greater than 40%. The maximum fill in any of the fuller penetrations will be less than 60%.
The above responses, details, and tests are considered adequate to resolve and verify the fire rating of the fire barrier penetration seal configurations at the Braidwood Statio (Closed) 456/85053-07; 457/85051-02: The inspector requested additional hose protection be provided to facilitate manual fire-fighting efforts in the control roo Rather than relocate an existing fire hose station an additional fire hose station has been installed (No. 292) located at Column -
Row L-14. This fire hose station is also included in the pre-fire plan for the control room 1D-75. Fire hose station No. 292, along with fire hose stations No.15 (at L-24) and No. 27 (at L-12),
provide adequate manual fire fighting protection for the control room. This issue is close (Closed) 456/85053-08; 457/85051-03: The inspector informed the licensee that since the valves were designed and constructed to perform the same function as approved valves that are UL listed, this may be an acceptable deviation from NFPA Standard No. 14 if identified and justifie Subsequent review of this item indicated that there are no listed fire protection valve test standards for seismically qualified or end welded water control valve The valves manufactured by Anchor Darling Inc., ITT Grinnell Inc., and Powell Valves Inc. have all been identified as valves used in containment where they must be seismically qualified as well as fire approve They were manufactured and tested to ASTM and ANSI Standards and these valves meet or exceed the requirements for UL listin This issue is-close .
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(Closed) 456/86002-19; 457/86002-19: The inspector discussed the location of the deluge spray valves protecting the charcoal filters-in the control room emergency makeup, post LOCA purge, auxiliary building filtered tank vent exhaust, and the TSC air cleaning systems with the license The concern was'over the proximity of the valves to the protected filters and whether the heat from a fire in the filter would prevent access to the valve to control the fir Another issue was the physical location of the deluge valve for the-TSC syste An inspection of the air cleaning systems was conducted and a review of the licensee's Discussions were held with the licensee' pre-fire plans was s Fire Marshall andmad HVAC engineer regarding the air cleaning systems flow paths and functional operating characteristic The licensee moved several of the valves to a more accessible location. They have also addressed and made provisions for gaining access to the valves for manual operation during an incident involving elevated temperatures in the immediate area. This item is close (0 pen) 456/86045-02; 457/86033-02: The walkdown of the procedure entitled " Control Room Inaccessibility, Revision 1, No.180WA PRI-5, dated July 18, 1985 identified that the portable radios were deficient, specifically between the remote shutdown panel and containment. The licensee indicated that the deficiency of the radios was caused by the temporary removal of power to the repeater station inside containment. In a telephone conversation on October 22, 1986, the licensee indicated to the inspector ti,1at portable radios can only be used during refueling because of noise spiking. The licensee indicated that there are other means of communication such as the paging system and the telephone system. The inspector questioned whether the paging system and the telephone would be available should a loss of offsite power occur. The licensee indicated that the telephone system will be the only form of communication should a loss of offsite power occur because it receives power from the security diesel. During a recent inspection by NRC consultants, verification was made that the phone system is on the security diesel, bus 033WS. A review of the remote shutdown area revealed that no phone or phone jack was available. The licensee installed a phone line in the remote shutdown panel are The licensee was requested to provide an analysis that demonstrates that communication is available when " Control Room Inaccessibility Procedure (1 B0WA PRI-5)" is utilized during a fire and loss of l offsite power. This item will remain open pending review and j acceptance of licensee's submitta r
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(0 pen) 456/86045-03; 457/86033-03: The fire wraps which are required to protect redundant trains were not installe The licensee was able to demonstrate with drawings and engineering change notices that provisions are underway to have the fire wraps installed. The installation of the fire wraps required to comply with Appendix R, III.G.2, is incomplet During a recent inspection by NRC consultants, it was noted that approximately 75% of the required fire wraps have been installe It was discussed with the licensee that until work on the fire wraps is complete that compensatory measures as indicated by Technical Specification would be acceptable. This item will remain open pending verification of the completed work by Region II (0 pen) 456/86045-04; 457/86033-04: In the auxiliary building elevation 346 feet, the inspectors observed that redundant RHR Systems (including pumps, motor, unit coolers and cables) were not separated / protected as required per the requirements of-Section C.5.b.(2) of the staff fire protection guidelines. In Amendment No. 7, it indicates that the RHR System could be repaired within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (which~is the prescribed time to achieve and maintain shutdown, and includes estimated times to complete any repairs). Allowance for repairs is consistent with the guidelines of Section C.S.b.(1) of BTP CMEB 9.5.1. However, the applicant's repair procedure only addresses repair of cables. The cable repair procedure has been reviewed and found acceptable. The licensee has no procedures for repairing other vulnerable components of the RHR system. The RHR System is vulnerable to fire damage. Therefore, the plant may not be able to achieve and maintain cold shutdown within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> as stipulated in Section C.5.b.(1). The Braidwood Safety Evaluation Report (SER) Supplement No. 2, dated October 1986 stated "By letter dated September 22, 1986, the applicant committed to upgrade the wall between the redundant RHR pumps to one having a 1- hour fire rating. Before exceeding 5% of rated power, the applicant will notify the staff if the upgrade of the wall is not complet On the basis of limited combustible content, the staff concludes that the 1- hour rated wall is acceptable protection for the redundant RHR pumps."
During a recent inspection by NRC consultants, it was noted that the wall at 343' level between RHR pumps for Unit 2 along wall line "W" between columns 21-23 was still under construction. It was discussed with the licensee that until work was completed for upgrading the wall that measures as indicated by the Technical Specifications would be acceptable. This item will remain open pending verification of the completed work by Region II (Closed) 456/86045-05; 457/86033-05: The sprinkler protection provided in the auxiliary building open ;1atchway between elevations 401' and 426' did not provide adequate protection for the unprotected opening. The licensee was requested to provide adequate sprinkler protection by the use of sprinklers below the floor and the use of the draft stop .
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Additional water curtain sprinklers have been provided around the open hatchway at the following elevations: 426', 401', 383', 364',
and 346'. The draft curtains have also been extended to 18" at elevations 426', 401', 386', 364', and 346'.
This item is considered close (0 pen) 456/86045-06; 457/86033-06: In response to this item the licensee made several recommendations (technical justifications)
and relocated some of the obstructed fire sprinkler system head An inspection / review of these items was made with the licensee and their NFPA code consultant of M&M Protection Consultants. Listed below are the results of that review:
- Area 1-GG Waste Oil Tank - Head No. 1 and 2 are now satisfactorily located. The response regarding head No. 3 is acceptabl * Area 1-JJ Component Cooling Pumps - For all of the heads identified, except number 40, the action taken or the technical response is acceptable. Head No. 40 was moved, but needs to be repositione * Area 1-LL Pipe Penetration Area - Of the fire sprinkler heads identified Nos. 1, 2, 4, 10, 12, 13, 14, 16, 17, 18, and 19 have been either satisfactorily relocated or
! the technical response is acceptable. Heads No. 6, 11,
- and 15 were moved, but need to be repositioned agai Head No. 20 still needs to be move Regarding the repositioning of the heads mentioned above, representatives of Ceco and M&M Protection Consultants accompanied the reviewer during the inspection and committed to repositioning the remaining five sprinkler head It was discussed with the licensee that until repositioning the
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remaining sprinkler heads, compensatory measures as indicated by I
the Technical Specifications would be acceptable. This will remain an open item )ending verification of repositioning of the five sprinklers ay Region II (Closed) 456/86045-07; 457/86033-07: The inspector requested I adequate hose coverage be provided for the cable tunnel areas l including the valve galleries.
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Hose station No. 129 (364' level turbine building at L-7) has been equipped with 200 feet of 1 1/2 inch fire hose and hydrant hose l house 0FP14S has 200 feet of 2 1/2 inch and 50 feet of 1 1/2 inch
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hose.' Both locations have adequate length to provide coverage for .
the Unit 1 main steam and auxiliary feedwater pipe tunnels and valve galleries. Pre-fire plans 10-14 and 10-15 have been revised to include this information. Surveillance OBWOS 7.10.5.a-1 has been
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revised to include the hose information.
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The above response is considered adequate to resolve the question raised by the inspectors. This issue is closed.
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(0 pen) 456/86045-08; 457/86033-08: The inspector observed in the auxiliary building, elevation 426', that gaps and holes existed in steel diamond plate at the hatchway, which under fire conditions could allow heat and products of combustion to rise into the vertical adjoining area During a recent inspection by NRC consultants it was noted that the licensee has generated Engineering Change No. 36033 to provide removable concrete panels and caulk with Dow RT 96-081 sealant hatchways at elevation 426' Column Q and 12, and elevation'364'
Column N and 18. The hatchway at elevation 426' Column U and 13 has been sealed at gaps and opening with Dow RT 96-081 sealan The hatchway at elevation 426' Column U and 23 will be filled as construction progresses on Unit 2. It was discussed with the licensee that until work was completed on the hatchway at elevation 426' Column U and 23,-compensatory measures as required by the Technical Specifications would be acceptable. This item will remain open pending verification of the completed work by Region II (Closed) 456/86045-09; 457/86033-09: The licensee was reguested to insure the integrity of the fire barriers by sealing the inside of the spare sleeve or provide test results from a recognized testing laboratory to justify that seals are not necessary. The licensee committed to seal spare conduits with fire resistant material to form a three-hour fire rated barrie The spare conduit sleeves have been sealed to the Transco Products-Inc. Detail BR-E-26. This detail has metal-caps or plugs on both ends of the sleeve, and a five inch minimum ceramic blanket fill in the center of the sleeve. The detail assembly was tested, "Transco Products Inc. Test Report," No. TR-207 verifying the three-hour fireratIngoftheassembly. All spare sleeves have now been seale This issue is close (Closed) 456/86045-12; 457/86033-12: The inspector noted that the
" position" indicator on a PIV was not in it's proper position. The licensee stated that the preoperational tests were not complet The licensee has completed a system demonstration test which
identified and corrected several minor deficiencie They have also established a surveillance procedure, OBWOS 7.10.1.1.e-3, that requires a quarterly inspection of all PIV ;
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The system demonstration test and results were reviewed, the surveillance procedure was reviewed and an inspection was made of-randomly selected PIVs (No. 0FP580, 802,'386,-and 588). All were operated through their complete cycle and the " position" indicators functioned properl The licensee's actions and responses are considered adequate to resolve the items raised by the inspectnr. This-issue is close (0 pen) 456/86045-13; 457/86033-13: The inspector discussed with the licensee action taken to address the 10 CFR Part 21, report submitted by Ruskin Manufacturing Company regarding the potential problem with fire dampers failing to close under normal. duct pressur The licensee was requested to provide a valid engineering analysis which should have included but not be limited to parameters such as size of damper, air flow, identification number /lecation and if -
installed in fire walls utilized for Appendix.R consideration It was recommended'that (regardless it may)be of manufacturer . beneficial to review all fire dampers An option to providing the engineering analysis for the dampers was to shut off inlet and exhaust air upon initiating the fire brigad Should the licensee have elected not to provide an engineering analysis then the licensee should have provided the plant operating procedures and/or pre-fire plans for revie In a letter dated January 27, 1987, the licensee stated that "At: '
Braidwood all safe shutdown fire dampers will be tested in the preop testing program at the site. The testing is carried out using one of the following method (1) For fire dampers with fusible links, the test engineer manually removes the fusible link from the damper and then verifies the closure of the fire damper. This testing is done normally with the ventilation system out of service. Therefore, no air flow is present when these dampers are teste Fire dampers that have only fusible links for activation are not part of the boundaries for areas with gaseous suppression system (2) For those fire dampers with electro-thermal links (electrically operated fusible links) that are not part of gaseous suppression system boundaries, the testing is accomplished by remote electrical activation of the fire damper. During this type of testing the fire damper's electro-thermal link (ETL) is activated by a signal from an ionization duct detector in the ductwork for the syste The ionization duct detector and the ETL of the fire damper are interlocked through a relay so that when the duct detector is l activated a signal is sent to the ETL at the fire damper 'which I causes the ETL to melt and results in flow drop' ping of the dampe This testing is done without air flow through the. syste l L
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(3) For those. fire dampers.with electro-thermal links that are part of gaseous suppression system (carbon dioxide and halon)
boundaries (see Attachment A), the-testing is accomplished by remote electrical activation of the ETL at the fire-dampe These ETL's are interlocked with the electrical controls of the gaseous suppression systems. During this type of testing the fire damper's ETL is activated by a signal from the controls -in'
the gaseous suppression system-logic. Activation of the ETL results in the closure of the fire damer. This testing _is done with normal ~ air flow through the syste In addition to.the above, some of the fans serving.the areas noted in Attachment A can also be tripped due to interlocks with ionization duct detectors on the systems (that'information'is also noted in Attachment A for your reference). It is apparent therefore that in the event of a fire in many_ areas of the plant the fans may be shutdown automatically without any fire brigade or unit operators actio It is not considered necessary to incorporate any operator action to
, shutoff inlet and exhaust air upon initiating the fire brigade."
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On March 6, 1987, the NRC inspector met with the licensee to discuss the Part 21 on Ruskin Fire Dampers. The licensee indicated that there are only 18 Duct Mounted Ruskin Fire Dampers. And these dampers could be enveloped in the test of the four section 58". x 36" ganged Ruskin Fire Damper with 45,000 CFM Air Flo The licensee-indicated that the test consisted of tripping the dampers, hearing the four dampers close and opening the access panel (without hearing
any further damper falling).
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The inspector indicated to the lice'nse that this test is
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- The fan was tripped after activation of the damper due to the high pressure interlock cutoff which reduced the flow to less than the specified flow of 45,000 CFM.-
a The license indicated that the test engineer heard the dampers drop and after removing the access panel heard no further dampers drop (at the time of the meeting the test engineer was not available).
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- The inspector reviewed the information provided by the licensee and additional information was needed to adequately address this concer The licensee was reqested to provide the inspector with the following:
(1) Test objective for conducting "The Ruskin Damper Under Air Flow Test" (Damper 1VD23YA)
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-(2) -Description o'f event regarding '.'The Ruskin' Damper Air -
Flow: Test"J(Damper IVD23YA) '
-(3) Actual ~ sign off sheet for enveloping the fire damper test (Damper IVD23YA)- '
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.'(4) -Engineering analysis - additional information regarding~ .
the. enveloping approach used by the' licensee was requested, which wasito include but not be limited.to velocity.-(FPM),
size of. damper, air. flow' pressure, type;of spring, vertical or. horizontal,.and type of link.-
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The licensee agreed to conduct a-test utilizing a viewing port
- for visually observing. the Ruskin -fire' damper trip under air -
flow conditions. -The licensee is requested to forward the test-objectives and notify Region-III'approximately two weeks prio .to testing the Ruskin fire damper under air flow condition This item will remain open pending review and-acceptance by Region-II It was recommended to the licensee that it may be beneficial to: review all fire dampers regardless of manufacturer.' 'It is-the inspector's understanding from the licensee that no further action will be taken at- '
this time-regarding fire dampers.other than Ruskin fire' dampers failing under air flo '
In the' Byron /Braidwood Stations Fire ProtectionLReport Volume'3,-
Braidwood Section 3.0, " Guidelines of BTP- CMEB 9.5-l',',' _ Paragraph 3.5.a(4)'
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states " Ventilation system penetrations in rated; barriers?are protected by fire dampers having~.a rating equivalent.to that of the. barrier." In the Underwriters Laboratory standard for. safety entitled " Fire-dampers
ceiling dampers is evaluated on the basis that air-conditioning:and ventilating systems are automatically shut down when'a. fire occurs a described in the various provisions of the standard for the installationL -
of air-conditioning and ventilations system,1 NFPA No. 90 A. , Therefore,'
the ratings are applicable to fire dampers ~~and ceiling dampers-installed'
L in systems:where air movement is~ effectively stopped at the start of.a i fire." -
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' A January 27, 1987 letter to J. % aler,-NRC from D. Farrar, CECO listed
- _ several American Warming and Ucati ating Dampers (AWV); In addition,2the latter states "As you can ;& r r the attachment, the manufacturer of'
I the dampers being tested m5a ( flow conditions at Braidwood isinot l limited to Ruskin Company. Thereiore, we'are certain that our testing program also demonstrates the capabilities;of other vendors' dampers to close under air-flow." ,
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2 i It is:.not clear froni the Januaty. 27,1987 -letter that all the' fire' damper manufactures.:have been identified and'that'the dampers submitted. envelope 0 the fire dampers for'AWV. In addition, the testing methodology was notL
~ described-in detail-for these test '
- The licensee is requested to-provide. adequate justi.fication that~all'
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i Ldampers: required to close under air flow conditions'(other than the-
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~previously identified Ruskin fire dampers) will close to ensure the integrity of the1 fire barrier.~.Thistis considered an unresolved ites L(456/87007-07;.457/87006-04) pending Region III' review and acceptance-Lof the licensee's submitta ,
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~(Closed):456/84006-01; 457/84006-01: The NRC inspector requested .
the licensee to review the area of electrical component removals and
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reinstallations to ' assure that all removed items were properly'
reinstalled and reinspected. In response to the~NRC inspector's: 1 request, the licensee initiated an Avoid Verbal Order-(AV0) Review '
Program and an Area Completion / Turnover Program and-issued an Comstock (LKC) procedure for. controlling rework of electrical components. In regard to these actions, the inspector determined the following:
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(1) Prior to March 29, 1984, AV0s were_used by LKC to instruct-the electrical craft to remove, replace or rework electrical- d components. The purpose or method of using AV0s was not- '
proceduralized. The use of this' document resulted.in the
- removal, reinstallation, or rework-of electrical ~ components without the required installation or~ inspection controls. To l resolve this matter of AV0s, the licensee discontinued the use of AV0s in March 1984,1for installation activities and documented the inappropriate use~of AV0s on LKC Nonconformance Report (NCR) 1996 issued on March 30, 1984. The' disposition of this NCR required a. review of approximately 4,000 AV0s to ascertain the need for an inspection.-
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(2) To determine the need for inspections based.on the AVO data-collected during the disposition of NCR~1996, LKC utilized their program for area turnovers which was described-in LKC Procedure 4.13.3, " Area Completion / Turnover." .The area -
turnover program required LKC engineering to scope an area
- using~ tabulation drawings,1 specification lists, etc. on item
. checklists. These checklists supplemented the existing equipment installation tabulation drawings. Inspection requirements for' items were then identified by LKC' engineering'
on those checklists. The item' checklists ~were developed by- ;
- area and were known;as Area Scoping Packages'(ASP). -After 1 completing the ASP, LKC Engineering transmitted the: ASP to the !
LKC QA. department'for record packaging and inspection statu '
-The LKC QA department reviewed all documentation associated with the' ASP for the following attributes:
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h -_m .-,w..e .,-e-w .m- r r w -.mt ,,-.e se m w--r-*-. ++w4 4eh,,,,- .=,.,.w , h er y, y,y,,*-,r -,w+ .--yrw, W-a-,,+-gw--,e% -,-, ,-e, -e w yew
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(a) verified that required and supplemental installation / inspection records have been included in the document packag (b) verified that inspection / installation documents contained in the package had been properly reviewe (c) verified that the documentation package formed a logical history of the installation and inspection of the item (s).
(d) research all NCR/ICR numbers which were referenced on the inspection records as to their statu After the ASP had been completed by the LKC QA department, LKC engineering reviewed the installation status of all items in the ASP. This was accomplished by reviewing all engineering documents generated after the identified inspection dates to assure that items were installed / inspected in accordance with current engineering documents. LKC engineering utilized the information compiled from the AVO reviews performed in accordance with the disposition to LKC NCR 1996 to determine if any additional inspections were required for an item whose installation had been affected by an AVO.
, (3) LKC issued Procedure 4.3.24, " Rework," on March 29, 1984, which delineated the methods to be utilized for controlling and tracking the rework of electrical components. These controls included methods to ensure LKC QC involvement in rework and/or reinstallation Based on the inspectors review of this unresolved item, the licensee was informed that the practice prior to March 1984 of using unproceduralized AV0s to control installation activities affecting quality is considered a violation of Criteria II of 10 CFR 50, Appendix B (456/87007-01(DRP); 457/87006-01(DRP)). The licensee has implemented the corrective action to preclude recurrence. These actions are described in Paragraphs (1), (2), and (3) above, and are summarized as follows:
Corrective Action - In March 1984 the licensee instructed LKC to discontinue the use of AV0s for installation activities and documented the inappropriate use of AV0s in LKC NCR 199 Also, to evaluate the need for inspections on those items affected by AV0s, LKC utilized the area turnover process described in LKC Procedure 4.1 To determine the effectiveness of this corrective action, the inspector selected ;
60 conduit hangers, cable trays, cable tray hangers, and junction boxes in those Unit 1 areas which had been completed ,
using LKC Procedure 4.13.3, " Area Completion / Turnover." No i problems were noted with the hardware or documentation. The l l
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results of these inspections are documented in Paragraphs 9 and 10 of this_ report. Even though not all areas of-Unit 1 and Unit 2 have been completed under the auspices of LKC-Procedure 4.13.3, Lit appears from the results of these inspections that the Area Completion / Turnover Program has been effectively implemente Corrective Action to Prevent Recurrence - LKC issued Procedure 4.3.24, " Rework," on March 29, 1984 to control the rework of electrical components. To determine the effectiveness.of the corrective action to preclude recurrence, the inspector selected a sample of Unit 2 electrical hangers in addition to those items inspected in Unit.1. No problems with the hardware or documentation were noted. Based on the results of these inspections, it appears that the LKC 4.3.24 has been effective
, in controlling rework activitie Based on the issuance of the violation identified above, the unresolved. item is considered close (Closed) 456/85038-05: The electrical contractor was unable to produce quality control welding inspection records for cable tray sup This item was subsequently addressed in Inspection Report 456/8603 ports. The four hangers which were missin records were inspected by L.K. Comstock (LKC) g weld These inspector inspection inspections were documented on the appropriate forms. However,.this item remained open until the completion of LKC area walkdown activities. The inspector reviewed LKC's area walkdown activities defined in LKC's Procedure 4.13.3, " Area Completion / Turnover." To determine the effectiveness of these walkdowns, the. inspector selected conduit. hangers-and cable tray hangers.in those Unit1 1 areas which had been completed using LKC Procedure 4.13.3. The-results of these inspections are documented in Paragraphs 9 and 10 of this report. No missing weld inspection records were noted during this inspection. This matter is considered close (Closed) 456/86045-11; 457/86033-11: The inspector identified that the licensee should have hydrostatically tested a-section of' fire-main at 280 psi rather than 235 psi and that'a procedure should be used to limit " test" pressures in the underground during fire pump test procedures by using two isolation valves in series and monitor the pressure downstream of the first valv The licensee has retested the affected sections of the fire main to -)
325 psi and established a testing procedure using two isolation l valves in series during the. churn portion of the fire pump test procedur ,
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The inspector reviewed the results of the hydrostatic test for the section of fire main from the two fire pumps to the two isolation valves, Test No. 694 for the motor driven fire pump discharge piping, dated December 3, 1986 and Test No. 695 for the diesel driven fire pump discharge piping, dated December 9,1986, and found them acceptable. Test Procedure, " Fire Protection Pump Flow and Pressure Test," was also reviewed and found acceptabl Based on the licensee's corrective actions, this item is considered close c. Safety Evaluation Report (SER)
(Closed) 456/8600005: Upgrade Essential Service Water (SX) System to and from containment fan coolers to Quality Group B; add debris screen to miniflow purge system supply duct and upgrade the Reactor Containment Fan Coolers (RCFC) to Quality Group B. The inspector reviewed the following drawings to verify that these SER items have been incorporated into the design:
Drawing Revision Sheet Items M-42 V 5 SX Upgrade M-106 Z 1 Debris Screen Drawing M-42 required magnetic particle examination on RCFC coil nozzle joint at the water box header and radiographic examination of the weld at the flange connection to the nozzle. This drawing also identified that the SX piping was to be installed as Quality Group B inside the containment. The inspector had a regional specialist review radiographs for some of the flange welds to verify that the drawing requirements were met. The following radiographs were reviewed with no problems noted:
1C-1VP01 AA-5-FW-1 1C-1VP01 AA-5-FW-2 1C-1VP01 AB-1-FW-1 1C-1VP01 AB-1-FW-2 1C-1VP01 AB-5-FW-1 and R-1 1C-1VP01 AB-5-FW-2 1C-1VP01 AA-1-FW-1 1C-1VP01 AA-1-FW-2 1C-1VP01 AC-FW-1 and R-1 1C-1VP01 AC-1-FW-2 1C-1VP01 AC-5-FW-2 1C-1VP01 AC-5-FW-1 and R-1 1C-1VP01 AD-1-FW-1 1C-1VP01 AD-1-FW-2 1C-1VP01 AD-5-FW-1 1C-1VP01 AD-5-FW-2 and R-1 This matter is considered close , _ .. .
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(Closed) 456/8600024;.457/8600024: Procedures to Depressurize the-Reactor Coolant System Should Include a Precaution to Protect-the Pressurizer Relief Tank (PRT).
In the Byron SER, a commitment ~is made that states, "To reduce the potential for a higher than normal. containment temperature and humidity, the applicant has stated that the procedures'for depressurization using the PORVs will include precautions to ensure integrity of the pressurizer relief tank (PRT)
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The inspector reviewed the applicable procedures ~and verified that~a precautionary staterent is provided stating, " Caution - When -
Depressurizin Maintained." gThe the following RCS Using the.PZRwere procedures PORV, Ensure PRT Integrity is reviewed:
18wEP ES- " Natural Circulation Cooldown"
" Natural Circulation Cooldown With Steam Void in
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18wEP ES- Vessel (With RVLIS)"
1BwEP ES- " Natural Circulation Cooldown With Steam Void in Vessel (Without RVLIS)"
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In each case, the caution was appropriate, was made a permanent modification, and was reviewed and implemented as committed. This SER item is considered close . Licensee Event Reports (LER) Review
- Through direct observations, discussions with licensee personnel, and i review of records, the following event reports were reviewed to determine that reportability requirements were fulfilled, immediate corrective action was accomplished, and corrective action to prevent recurrence had
been or would be accomplished in accordance with technical specifications:
- (Closed) 456/87005-00-LL: Incorrect Wiring of Digital Rod Position Indication System causes Unreliable Rod Position Indication and
. Subsequent Manual Reactor Trip. The licensee corrected the; .
manufacturer / construction discrepancy by modifying the' associated wiring and changing appropriate electrical drawings through their Action Item Tracking (AIT)' syste ,
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- (Closed) 456/87006-00-LL: Violated fire Watches Attributed to Personnel Error. The licensee has re-instructed those personnel involved or taken other appropriate disciplinary action such as
dismissal. This is in violation of Technical Specification 3.7.11 and is further addressed in Paragraph'4,'" Exceeding" Technical Specification Action Statements for Fire Protection (456/87007-02(DRP)).
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- (Closed) 456/87007-00-LL: Missed Technical Specification Surveillance Due to Improper Scheduling. .The licensee determined that the cause was due to improper computer scheduling for the staggered surveillance. The computer program has been corrected for the reactor trip breaker and other appropriate equipmen * (Closed) 456/87008-00-LL: Reactor Trip Caused by Deenergizing the Nuclear Instrumentation System Due to Water Intrusion Into Control Room as a Result of a Plugged Drain. The licensee has initiated several corrective actions; such as clearing the plugged drain line, informing contractor personnel by letter of the importance of floor drain cleanliness, placing mesh catch devices in floor drains with regular cleaning and inspection, reevaluating the need for additional hydrolazing of floor drains, investigating and evaluating the affect of the mild etching acid on the affected panels, and initiating a program to require shift supervisor approval of any floor etchin A similar event also occurred in Unit 2 on April 2, 1987, and will be evaluated during the next inspection perio * (Closed) 456/87009-00-LL: Reactor Trip From Pulling Intermediate Range Control Power Fuses While Trouble Shooting as a Result of Cognitive Personnel Error. The personnel involved were counseled and an operator aid sheet on the trips associated with pulling specific fuses will be generate * (Closed) 456/87010-00-LL: Inadvertent Loss of Power to Instrument Bus 111 Resulting in a Reactor Trip Due to Personnel Error. An electrical lineup was verified and construction personnel in the area of the affected breaker were counsele * (Closed) 456/87011-00-LL: Loss of Residual Heat Removal Due to Loss of Component Cooling as Result of Leaking Component Cooling Inlet Valve. The licensee corrective actions were: repair of the leaking valve, adjusting the limits on the motor operator of the leaking valve, troubleshoot the alarm annunciator and an action item to check the scale and calibration of the surge tank instrument loo * (Closed) 456/87012-00-LL: Fire Detection Zone 10-8 Inoperable due to Unauthorized Removal of a Detector. This is considered in violation of Technical Specification 3.7.11 and is further addressed in Paragraph 4, " Exceeding Technical Specification Action Statement for Fire Protection" (456/87007-02(DRP)).
- (Closed) 456/87013-00-LL: Manual Reactor Trip Due to Inadequate l Procedur This resulted in a control rod misalignment when the l licensee chose to drive the affected bank (A) in until the first rod l indicated a fully inserted position. A manual trip was initiate l The rod timing mechanism has been corrected and a procedure change i was initiated to prevent recurrenc '
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- (Closed) 456/87014-00-LL: Violation of Tech Spec Action Step 3.6.1.3(a). The licensee's leak rate test verified that containment had not been breached. The corrective actions were to complete the door repairs, instruct operating and technical staffs, and compile a list of all missed Technical Specifications since license issuance for shift engineers. In order to prevent recurrence, the licensee has or is providing an electrical outlet in the airlock to retain the operable door shut, evaluating the possibility of providing test features from outside the airlock, and revising Procedure BwVS 6.1.3.a-2,
"LLRT of Containment Airlock Doors." In recognition of the potential seriousness of this event, the licensee developed a Potentially Significant Event (PSE) Report, dated February 28, 1987, on this subject in accordance with their Nuclear Station Division Directive NSDD-A07. This is discussed further in Paragraph * (Closed) 456/87015-00-LL: Manual Reactor Trip Due to Deficient Procedure as a Result of Rod Misalignment During Testing. This is essentially the same as LER 456/87013 and the corrective action is the sam With the exception of LERs 456/87006 and 456/87012, the preceding LERs have been reviewed against the criteria of 10 CFR 2, Appendix C, and when the incidents described meet all of the following requirement, no Notice of Violation is normally issued for that item: The event was identified by the licensee, The event was an incident that, according to the current enforcement policy, met the criteria for Severity Levels IV or V violations, The event was appropriately reported, The event was or will be corrected (including measures to prevent recurrence within a reasonable amount of time), and The event was not a violation that could have been prevented by the licensee's corrective actions for a previous violatio In addition to the foregoing, the inspector reviewed the licensee's Deviation Reports (DVRs) generated during the inspection perio This was done in an effort to monitor the conditions related to plant or personnel performance, potential trends, etc. DVRs were also reviewed for assurance that they were generated appropriately and dispositioned in a manner consistent with the applicable procedures and the QA manua Concerns with a number of DVRs are addressed in Paragraph 13, " Operator i Awareness" and Paragraph 16, " Procedure Reviews."
One violation was identifie l l
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4. Exceeding Technical Specification Action Statements for Fire Protection While reviewing a number of Licensee Event Reports (LERs), the inspectors noted several examples of required fire watches not meeting the required time limits of the Technical Specification (Tech Spec) 3.7.11 action statemen Reference Paragraph 3, " Licensee Event Reports." These-multiple examples of failure to meet the requirements of the Tech Spec 3.7.11 action statement are as follows:
Licensee Event Report 456/87012 is summarized: On February 17, 1987, at approximately 1035 hours0.012 days <br />0.288 hours <br />0.00171 weeks <br />3.938175e-4 months <br />, the Operating Department was informed that detection zone 10-8, located in the 401 elevation of was in TROUBLE alar Per Technical the Auxiliary Building, Specification 3.7.11, F ire Rated Assemblies, a continuous fire watch ( was required to be established in the area within one hour after the detection zone became inoperable, as a result of inoperable fire detection existing on two sides of fire walls in the area which contained inoperable fire seals. A review of the available Main Control Room alarm history indicated that the alarm was received on February 13 at 1606 hours0.0186 days <br />0.446 hours <br />0.00266 weeks <br />6.11083e-4 months <br />. Ninety-one hours and thirty-nine minutes passed before a continuous fire watch was established in the are Cause of the TROUBLE alarm was due to unauthorized removal of a detector in detection zone 10-8 by unidentified personne Inadequate Control Room response to the detection zone TROUBLE alarm was due to cognitive personnel error. The 10-8 detection zone was '
repaired and returned to servic Licensee Event Report 456/87006 is summarized: At 0400 on March 18, 1987, the Route 4 hourly fire watch patrolman was beginning her route when she was detained. Radiation chemistry personnel informed her j that she could not enter the auxiliary building since she had not
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signed the latest Radiation Work Permit. The patrolman immediately signed the applicable permit, but was 24 minutes late before she patrolled the 146 elevation in the auxiliary buildin At 0600 on March 31, 1987, the Route 4 hourly fire watch patrolman was waiting to be relieved at the starting point of the rout Contrary to procedure requirements, the patrolman did not start the route at 0605. Due to a lack of personnel availability, the scheduled relief was approximately 33 minutes late. As a result the general areas of the 416 elevation and the 451 elevation in the auxiliary building were patrolled approximately 50 and 22 minutes late, respectivel The failure to post continuous and hourly fire watches as required by the Tech Spec 3.7.11 action statement for fire rated assemblies is considered a violation of the Limiting conditions for Operation of Tech Spec 3.7.11 (456/87007-02(DRP)).
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5. Strike by Contractor Employees On March 13, 1987, construction laborers working for a licensee contractor, G.K. Newberg Co., went out on an unsanctioned (wild cat)
strike. The issues were over unresolved grievances for shift premium pay and pay for wearing radiation monitors (film badges or TLDs). The incident received some media interest locally and in Chicago, Illinoi The matter was resolved on March 27, 198 During the strike, the
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resident inspectors monitored the activities and verified that normal
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access to the plant was not significantly affected, that plant security was unaffected, and normal operations were unperturbe . Regional Request
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Unqualified AMP Splices - Request for Information By memo, dated March 3, 1987, Region III requested information regarding l the use of unqualified AMP electrical splices. The following questions were asked by Region III: Identify any kind of AMP splice or terminal lug used in EQ applications at the plant. Include model type and insulation.
! Describe the installed configuration. Include information on enclosure (Nema 3, weep hole), separation of AMP splices in same t enclosure, and location of enclosure.
l l Identify any use of AMP splices in an EQ instrumentation circui Submit a copy of the licensee's justification for continued I operation for any unqualified AMP splices, and the licensee's j appropriate corrective action (such as taping over with 3M tape).
Through discussions with licensee personnel, it was found that AMP splices are used in safety-related EQ applications at Braidwoo However, no EQ credit is taken for the insulation on the AMP equipmen The licensee has applied additional qualified insulation over the AMP insulation. This is either Ray Chem heat shrinkable tubing or Okinite tap This information was provided to Region III via memo on March 31, 198 No violations or deviations were identifie . NRC Chairman Lando Zech On Site l
On February 19, 1987, NRC Chairman Lando Zech, Jr. was at the Braidwood site. During the visit, he met with the resident inspectors and members of the Region III staff, licensee corporate, and station personnel; toured the plant and the production training center; and observed a
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demonstration on the Braidwood/ Byron simulator. At the conclusion of his visit, the Chairman made a number of comments. In summary, the Chairman was impressed with management involvement and the model spaces. He stressed the importance of maintaining plant cleanliness and that meeting the aggressive schedules is commendable. He'went on to compliment the-chemistry program, the BCAP results, the " Error Free" plan, control room operators attitudes, and the management educational program. The Chairman cautioned about the importance of the shift from construction to operationsphases,outageplanning,"andattentiontodetail He emphasized the importance of the A and "B" operators and their
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experience, and the importance of doing the job right and well. In summarizing, he stated that Braidwood has shown significant improvement since 1984 and he encouraged the licensee to continue the momentum.
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During his discussions with the licensee's staff, Chairman Zech provided
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a number of comments that the licensee responded to by letter dated March 31, 1987. In general, the response addressed utilization of the simulator prior to initial criticality, employee education program for non-management employees, equipment operator training and experience, l containment status with respect to the model space program, outage l planning, and maintenance interface with operations.
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8. NRC Commissioner James Asselstine On Site l On March 24, 1987, NRC Commissioner James Asselstine was at the Braidwood l
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site for meetings with the resident inspectors and Regional staff and; meetings with licensee station and Corporate staffs. During the visit, the Commissioner made a plant tour and observed several drills on the Braidwood/ Byron simulator at the Production Training Center (PTC).
- Commissioner Asselstine commented that the visit was interesting and
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useful and that he was favorably impressed. He stated that Braidwood looks like it is working as a team and they should continue with the Model Spaces. The Commissioner cautioned about the upcoming work load during the next year with Unit 1 starting up and Unit 2 in preoperational testing. He cautioned about the challenge of the transition from construction to operations and paying particular attention to detail. He added a comment by stating, "How you start is how you will do for years,"
which is substantiated by a recent NRC report, e.g. a poor start leads to poor performance and visa versa. Finally, he stated with this favorable impression, he would like to see Braidwood at least as good as Byro . Electrical Equipment Installation - Unit 1 l
- The inspector verified that 60 welded conduit hangers, welded cable tray
- hangersandweldedjunctionboxeswereinstalledinaccordancewith l
specified requirements. Drawings were utilized to verify compliance with the following attributes:
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- locations
- configuration
- required welds
- weld acceptance TheitemsselectedwerelocatedintheUnit1safetyinjection(SI) pump rooms, the chemical and volume control (CV) pump rooms, the containment spray (CS) and residual heat removal (RHR) pump rooms, the containment building, ard other various auxiliary building area The weld inspection packages were also reviewed for the 50 hanger The following documents were reviewed:
- installation reports
- configuration reports
- final weld inspection reports
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- closed Inspection Correction Reports No problems were identified with the installation or documentatio The following items were inspected:
(H)= hanger,(T)=cabletray,(J)=junctionbox Area Elevation Title Drawing N Hanger N ' SI&CV Pump Room 0-3315A CP-016 H)
CP-020 CP-041 0-3314A CC-088 H)
CC-092 H)
CC-096 H)
0-3314 IJB093A(J)
5 346' Aux. CS/RH Room 1-3002 1515A T 1515B T 8HV2C H 8HV2B H 8HV2A 8HV2D 8HV2E 8HV2F (H)
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Area' ' Elevation Title Drawing N Hanger No'.
- 8HV38-7HVl(H).
8HV4A'(H)
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8HV4B.(H)
8HVSA (H).
8HV5B1(H)
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1-3302A 'CP-019(H)
l WH-001 (H)
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- 2 330' Aux. Building .0-3010 2H2 (H)
l- 2H1A-(H)
i 2H1B,(H
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2 HIC (H CC104()
0-3309A CC-CP-501(H)
CC-509 H)
CP-507 CP-505 CP-504 C0A09F4 C0A0995-C0A0919 C0A0909 C0A0903 Area Elevation Drawing N Hanger N Unit 1 426' 1-3254 Containment H012(H)(J)
IJB731R
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j 426' 1-3251 H090(H)
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412' 1-3343 1JB690R(J)
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412' 1-3578 CC3(H)
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412' 1-3244 H70(H)
L H038 (H)
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412' 1-3241 H011 H041
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401'. 1-3533 1JB998R(J)
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! 401' 1-3534 CP002(H). " 401' 1-3513 l T52(H).
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Area Elevation Drawing N Hanger N l
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401' 1-3511 IJB954R (J)
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377' 1-3513A IJB192R(J)
TS4-501(1-2)(H)
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377' 1-3571 WCP1-1 (H)
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377' 1-3514A TS2-501(H)
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401' 1-3531A WV30(H)
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401' 1-3532 1JB172R(J)
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390' 1-3222H H022 (H)
1322E(T)
No violations or deviations were identifie . Electrical Cable Tray Hangers - Unit 2 The following cable tray hangers were verified for compliance to specified requirements by the inspector:
Location Elevation Drawing N Hanger N Unit 2 377' 2-3521A TS4-003 Containment 377' 2-3521A TS2-004 417' 2-3241H H039 417' 2-3241H H007 426' 2-3254H H064 The following attributes were verified for compliance to the requirements specified in the applicable drawing:
- location
- configuration
- required welds
- weld acceptance No violations or deviations were identifie . _ _ - _ _ - _ _ _ _ - _ _ _ _
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11. Piping - Unit 2 Theinspectorselected35weldedpipejointstoverifycompliancewith specified requirements. The verification consisted of reviewing the following documents and activities:
- inspection reports
- nondestructive test reports
- nonconformance report closures weld acceptability
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- welder certifications and qualifications
- weld procedure approval
- inspector certifications e inspection procedure approval
- marking of approved material
- weld rod issue and documentation l
- base metal documentation
Title or Location 150 N Weld N SI Pump (28) Room 2544A-128 18, 19, 20, 21 CV Pump (28) Room 2546A-246 8, 9, 10-1, 11-1
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2544A-130 6,7,8,9 l 2A CV-19 6, 10, 11 2A CV-19 21, 22
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CV Pump (2A) Room M-2544A-113 3,4,5,6 l
2A-SX-20 12C, 12D, 12E, 12F SI Pump (2A) Room 2A-SI-23 7, 7A, 8, 9 2539A-105 1 l 2A-SI-38 3d,23B, 3C, 3D No violations or deviations were identified.
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e 12. Pipe Supports - Unit 2 The inspector selected 30 pipe hangers for compliance with the applicable drawings. These pipe supports were located in the 2A diesel generator room, containment, safety injection (SI) pump (2A and 28) rooms and the chemical and volume control (CV) pump rooms (2A and 28). The pipe supports were verified for compliance with the following attributes utilizing the applicable drawings:
- location
- configuration
- required welds
- weld acceptance The inspector also reviewed weld inspection packages. The following documents were reviewed to verify compliance with specified requirements:
- inspection reports
- nondestructive test reports
- nonconformance report closures
- final inspection and installation data
- torque reports Following is a list of the pipe supports which were inspected:
Elevation Title or Location Hanger N ' 2A Diesel Room 2SX46001R 2SX47006R 2SX47005R 2SX47003R 377' Unit 2 Containment 2RH02031V 2RH02100X 2SI14031V 2RH02075R i 2RY060155 l 2RY06008X 2SIO9009X 2RH020075
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Elevation Title or Location Hanger N ' SI Pump (2A) Room SVFF40008T SVFF40007T 2SI15009R 2pi-SIO44-H012 2SI12002G 364' CV Pump (2A) Room 2SX17034G 2SX36030X 2FT0922-H007 2FT0922-H011 2CVF39001T 364' CV Pump (28) Room 2CVK61018T 2CV01A001R 2SX30058G 2SX15082G 2CVK61016T 364' SI Pump (28) Room 2SX56111T 2VFF40038T 2VFF40039T 2VFF40040T 2VFF40041T No violations or deviations were identifie . Operator Awareness During this inspection period, several incidents have been identified by the inspectors and the licensee pertaining to operator awareness. In a previous Inspection Report, 456/86063; 457/86044, a concern was also identified pertaining to operator awareness. This concern identified that on several occasions operators or Shift Control Room Engineers (SCREs) were not aware of the status of specific control panel items relative to ongoing plant activities. Several items noted during this inspection period pertaining to operator awareness were: On January 12 1987,thelicenseeissuedDeviationReport(DVR)
87-015whichIdentifiedthattheoperatorswereunawarethatthe Volume Control Tank (VCT) outlet valves, 1CV1128 and ICV 112C, had shut and the suction valves, ICV 1120 and ICV 112E, from the Reactor Water Storage Tank (RWST) to the charging pumas had opened. The operators became aware of this valve lineup w1en a VCT high pressure annunciator alarm was received. This resulted in the operator's discovery of the high VCT level and the ICV 112D and ICV 112E valves being open. The operators immediately implemented action to restore
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the plant to a normal valve lineup. The closing of valves ICV 1128 and ICV 112C and the opening of valves 1CV1120 and ICV 112E occurred as a result of a planned ?lant evolution. The evolution which caused the switchino of tie wction of the charging pumps from the
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VCT to the RWST was the deenergizing and reenergizing of instrument bus 112 caused when the source of power of. instrument bus 112 was switched from the constant voltage transformer to the inverter. The operators were aware that the Boron Dilution Protection System (BDPS) could be actuated. A BDPS actuation would result in closing valves 1CV1128 and ICV 112C and the opening of valves ICF1120 and ICV 112 However, after switching the source of power for the instrument bus 112 the operators cleared the annunciators which had alarmed and reset. The BDPS alarm did not remain lit. Therefore, the operators believed the BDPS actuation had not occurred. The inspectors stressed to the licensee's plant management that operators should not rely solely on annunciator alarms for plant status. The operators should also utilize the control board indications, especially in those situations where plant evolutions were expected to change system lineups. DVR 87-015 was closed on February 12, 1987. The corrective action identified consisted of the shift personnel involved being made aware of the proper annunciator response that is ex)ected when instrument power is momentarily)
not lock-in . interrupted (i.e.action The corrective tie BDPS switchover did not actuation address the would importance that operators should not totally rely on annunciator alarms, especially when plant evolutions were expected to result in changes to plant status, b. While observing control room activities, the inspectors noted that the extra on-shift Nuclear Station Operator (NS0) was acknowledging and silencing Unit 1 annunciator alarms without notifying the Unit 1 on-shift NSO. The inspectors observed that the on-shift NS0 did not appear to routinely acknowledge the alarms silenced by the extra on-shift NSO. The inspectors expressed their concern to the station management that the extra operators are not verbally announcing alarms in a manner to assure that operators responsible for plant operation are cognizant of the alarms, c. On February 19, 1987, complying with Technical Specification (TSa actionconcern pertaining)to statements was the plan discussed with the licensee's plant management. The inspectors noted, while reviewing Licensee Event Reports (LERs), that there were four events identified by the licensee where action statements in the TS were not implemented by the operators. These LERs were 86-002(11/17/86),86-006(12/4/86),86-008(12/12/86),and87-001 (1/2/87). Since February 19 1987, several more instances were identifiedconcerningTSactionstatements. These instances were identified in LER 87-012 (2-13-87), LER 87-014 (2/23/87), DVR 87-060 (2/21/87), and DVR 87-101 (3/11/87). DVR 87-101 is discussed in further detail in the following paragrap l l
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Upon identification of the event described in LER 87-014 (2/23/87)
the licensee classified it as a Potential Significant Event (PSE) in accordance with Nuclear Station Division Directive A.07. LER 87-014 identified an event where the operable outer door to the emergency air lock was opened to allow leak rate testing of the inoperable inner door. TS Action Statement A.3 associated with Limiting Condition of Operation 3.6.1.3 was not entered as required when the outer door was opened. One of the short term corrective actions identified in the PSE required a listing of all LERs since license issuance that pertained to missed TS action statements. These LERs were then to be discussed with all Shift Engineers. This discussion was conducted on February 25, 198 The concern is that if this activity continues, the potential exists for the licensee to exceed time limits for TS action statement Violation (456/87007-02(DRP)) is an example of exceeding TS action statement time limit DVR 87-101 was issued on March 17, 1987 which described an event which occurred on March 11, 1987. This DVR was issued at the request of the inspectors. While reviewing the Unit 1 operating log, the resident inspector noted that the IB Diesel Generator (DG)
was declared inoperable for maintenance at 0359 on March 11, 198 Subsequently, a log entry was made at 0517 noting that Unit 1 was entering Limiting Condition of Operation Action Requirement (LOCAR)
5.2-la due to the 1A Residual Heat Removal (RHR) Pump being taken out of service (OSS). Discussion with control room personnel revealed that the 1A RHR pump was being taken OSS for lubrication maintenanc Based on the log entries, it appears the shift personnel were not aware that placing the 1A RHR pump OOS rendered both trains of Emergency Core Cooling (ECCS) technically inoperabl Shift personnel realized this condition within five minutes of
" pulling to lock" on the 1A RHR pump control switch on the main control board. At 0524, LOCAR 5.2-la was exited and the 1A RHR pump was returned to service. Since the 1A RHR pump was returned to service within five minutes, no TS action statements were violate This event of placing the 1A RHR pump 005 is an example where operators initiated an 005 without proper consideration of plant systems interactions (1B DG inoperable) where the 005 rendered both trains of a ECCS technically inoperable. The review of the 00S for the 1A RHR pump for Technical Specification application was performed and documented by the SR0 Shift Foreman on the proceeding shift. Discussion of the 00S program is further discussed in Paragraph 14 of this repor . Out-of-Service While reviewing the event described in DVR 87-101 (see Paragraph 13.d of this report), the inspectors noted a concern with Procedure BwAP 330-1, Revision 52, " Station Equipment Out-0f-Service Procedure." This !
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procedure descr'ibes the methods and process for placing'a' component or system out of service. In the main body of the procedure, Section 4, the preparation of the Equipment Outage Form (EOF) is described. The preparation process of the EOF is briefly described as-follows: The requestor fills in the following information_in Section 1 of the EOF:
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- -- Work Request Number (if applicable)
- Equipment Part Number
- * Equipment Description
- Nature of the Work
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- Special Instructions TheSeniorReactorOperator(SRO)ShiftForemancompletesthe appropriate blocks of Section 2 of the EOF, answering the.following questions:
- Does this outage have Technical Specification applications?
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- Does this 00S require verification by a.second qualified-
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- Has the surveillance schedule been reviewed to ensure this outage will not prohibit completion of a scheduled surveillance during the estimated duratio '
' The Shift Engineer then assigns a sequential 00S number and delivers
- the EOF to the Control Room Superviso The Control Room Supervisor reviews the 005 and determines any actions necessary to satisfy Technical Specifications or
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surveillances. He then forwards the 00S to the center desk Nuclear StationOperator(NS0). The NSO reviews Section 1 and 2 of the E0F for completeness and determines the related card location (isolation) Section 3 of the
, E0F. TheNSOalsoindicatesthe005 position (Section3ofEOF). The operator (s) performing the 00S fills out the related 00S cards for each isolation point listed in Section 3 of the E0F. The operator (s)thenproceedstoplacetheequipmentintheplant005.
- and hang the 00S tags. If a verification is necessary an assigned
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on the EOF,
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. The center desk NSO informs the Station Control Room Supervisor that the 00S is complete and signs, dates, and time the EO Control Room Supervisor returns the E0F to the Shift Engineer's-office after reviewing it for completenes The Shift Engineer notifies the requestor that the equipment is 00 The process of initiating an E0F as described above and in Procedure BwAP 330-1 is a concern in that: The reviews by the SR0-Shift Foreman (Paragraph (3) above) could be performed without the related card location (isolation) in Section 3 of the E0F being identified. Therefore, it is conceivable that a component identified on the E0F as part of the 00S could affect a
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system not apparent in the description (Paragraph (1) above) part of l Section 1 of E0F. This could result in the SR0 and Control Room i Supervisor (CRS) not identifying a Technical Specification l application, and Since at the time of the SR0 and Control Room Supervisor reviews (Paragraphs (3) and (5) above) the 005 boundary (Section 3 of the E0F) and the 005 cards might not be either identified or completed, it is conceivable that the actual placing of the affected equipment 00S could lag the SR0 and Control Room Supervisor review by one shift or more. Plant status, especially outages, can change quickly which could affect Technical Saecifications in a way that was not previously identified during t1e reviews by the SR0 and the Control Room Supervisor. In addition, there are no provisions in the Administration Procedure, BwAP 330-1, " Station Equipment Out of Service Procedure," to document the SCRE or CRS review and granting permission to hang the 00S as recuired by ANSI-18.7,1976. (See Paragraph 13 of this report for cetails on how a change in plant status can effect an 005.)
The two concerns related to the potential delay between review and hanging the 00S and documentation of the SCRE or CRS review-and granting permission identified above are considered an Open Item (456/87007-03(DRP); 457/87006-02(DRP)) pending further review by the licensee and the NRC, No violations or deviations were identified.
l 15. Auxiliary Feedwater
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On March 6, 1987, the diesel driven auxiliary feedwater (AFW) pump was started in accordance with Procedure BwPT AF-10, retest 143. Ten seconds after the pump was started, it tripped due to low diesel lubricating oil
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pressure. Immediately after the trip the instrument air isolation valve l (01A623)fortheinstrumentairtovalvesISX173and1SX178wasfound
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close Valves 1SX173 and ISX178 failed open upon loss of instrument ai The function of valves 1SX173 and ISX178 is to supply cooling water to the AFW diesel jacket water cooler, gearbox lube oil cooler, pump lube oil cooler, right angle gear lube oil cooler, and the AFW diesel room cubicle cooler. The low lube oil pressure condition which cr.used the AFW diesel trip, appeared to be the result of the diesel jacket water bei over-cooled by the excessive cooling water flow (ISX173 and 1SX178 failing open) which was present while the diesel was shutdown. The l over-cooling of the jacket water caused a subsequent over-cooling of the I diesel lubricating oil which in turn caused a high differential pressure l-drop across the lube pressure condition. lube oil The filters upon engine licensee startupIon Report (DVF.)resulting in issued Deviat 87-073 to document this event.
l The inspectors reviewed the Final Safety Analysis Report (FSAR)
l pertaining to instrument air. Section 9.3.1.3 of the FSAR stater that l that the failure of the service air and instrument air systems wauld not prevent a safety-related component from mitigating the consequer.ces of
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any design-basis accident or performing its intended function u.1 der emergency cooldown condition Since the loss of instrument air would have the same effect on valves ISX173 and ISX178 (failing open) as the l
closing of their instrument air isolation valve (01A623), the inspectors havedeterminedthatanUnresolvedItem(456/87007-04(DRP))w]uldbe issued to document and track the resolution of the affects of loss of instrument air on the diesel Auxiliary Feedwater Syste No violations or deviations were identifie . Procedure Reviews During the inspection period October 24 through December 10, 1986 documented in Inspection Reports 456/86063; 457/86055, a weakness was identified by the resident inspectors in the area of procedure review During this inspection, a review of Deviation Reports (DVRs) by the inspector has revealed a continued pattern of weakness in procedure reviews. The following DVRs are examples of this pattern: ThelicenseeissuedDeviationReport(DVR)87-067onMarch2 1987, whichidentifiedthattheShiftEngineerwasnotaware,until completion of a surveillance, that an Engineered Safety Feature (ESF) was disabled during the performance of the surveillanc The surveillance 3rocedure BwVS 3.2.1-1, was being implemented to verify that tie actuation device which starts the 1A Auxiliary Feedwater pump (AFW) upon loss of power was operable. Procedure BwVS 3.2.1-1 did not have a caution statement that the 1A AFW pump would be disabled during the surveillance. The need for a criterion statement was due to the 1A AFW pump being disabled by lifting a lead in the auxiliary electric room instead of the 1A AFW pump control switch in the " pull to lock" position on the main control l
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boar The lifting of the lead was in accordance with the steps delineated in the surveillance procedure. The 1A AFW pump was disabled during the surveillance for approximately 15 minutes. The lack of a caution statement in the surveillance procedure and the procedure sp"ecifying lifting a lead to disable the 1A AFW pump rather than pulling to lock" on the pump's control switch were contributing factors in the shift engineer not being aware of the disabling of the 1A AFW pump during the surveillanc b. DVR 87-107 was issued on March 19, 1987, to identify that the shiftly channel check of the Pressurizer Pressure Channels 455A, 456/457, and 458 had been missed. These shiftly readings were required by the Technical Specifications when Unit 1 reached Mode 3 above P-11. This shiftly surveillance had been missed due to the Surveillance Procedure 18w05 0.1-1,2,3 not recuiring the shiftly channel check until the unit was in Mode 1 anc 2. Step F.1 required the verification of pressurizer pressure and level instruments in Modes 1 and 2 onl c. On March 31, 1987, the licensee issued DVR 87-123 to identify that while performing 18w05 3.1.1-20 (Train A Solid State Protection System Bi-Monthly Surveillance), Step F.13 was not performed as written. The Permissive Test Switch was left in position 11 instead of being returned to off. When Step 39.5 was performed, placing the input error inhibit switch to normal, a reactor trip signal was generated due to a turbine trip signal already being present (above P-7). The licensee preliminary investigation identified the cause to be personnel error. The inspector reviewed Procedure 18w05 3.1.1-20 and noted that Step F.13 required the individual performing the surveillance to " return the Logic A and the Permissive Test Switches to the 0FF position" with only one initial block for documenting both switches were returned to "0FF." The lack of having an initial block for each switch in Step F.13 was a contributing factor in the individual failing to return the Permissive Test Switch to "0FF."
d. DVR 86-101 was issued on December 20 1 which identified that theprocesscomputerbecameinoperable.986,isresultedinalossof Th the Safety Parameter Display System (SPDS) and accident assessment capabilit Procedure BwAP 1250-6A4 allowed a two hour down time prior to declaring the process computer inoperable. Several attempts to restart the process computer using Procedure BwVP 300-1,
" Process Computer Bootstrap Procedure " were made by the operating staff. These attem)ts failed. TheIIcenseedeterminedduringtheir investigation that )rocedure BwVP 300-1 was inadequate and needed to be revise e. On January 28, 1987, DVR 87-047 was issued to identify the tripping of the Process Computer Inverter, resulting in the loss of emergency assessment capability. The licensee's investigation determined that the contributing factor to delay in recovery was missing steps in Procedure BwVP 300-3, " System Recovery From Magnetic Tape Procedure."
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. DVR 87-055 was issued February 18, 1987, to identify that there was a minus (-) sign missing at Step F.11 in Procedure 18w05 4.6.2.1.d-1 for calculating density change. This procedure is utilized to perform the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> surveillance on the Reactor Coolant System leakag Based on the above DVRs, the inspectors have continued concerns with the licensee's procedure review process. The procedure review process will be evaluated during a future "On-Site-Review" (OSR) inspection. Until this OSR inspection is completed, the considered an Open Item (456/87007-05(procedure review process is
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No violations or deviations were identifie . Maintenance Lubrication Program While performing a review of the licensee's lubrication program for the SI, CV, RH, AFW, and CC pumps, the following discrepancies were identified: BwAP 370-1, " Station Lubrication Program," requires that lubrication reports be completed by the 25th of each month to facilitate updating of the computer program. Review of the computer listing showed several entries had not been updated since issuance of 10 CFR 50.57(c) license. The inspector requested a past due listing for the pumps mentioned in item d. below. All of the pumps (SI, CV, AFW, RH, and CC) showed past due lubrications. Further investigation by the licensee showed that the computer had not been updated. The lubricaticns had been tracked by a 3x5 card record system which was not addressed by any administrative program, Section g. of the BwAP 370-1 requires the Operating Engineer to be notified of any lubrications which are two months past due. Due to the fact that the computer schedule has not been updated, the past due listing was not an accurate assessment of past due lubrication Discussions with Operating Engineers and Lubrication Coordinator shuwed that a past due listing had never been supplied to the Operating Engineer Section a. and b. of DwAP 370-1 require that the SCRE and Shift Engineer are to receive a copy of the monthly lubrication schedul Discussions with a shift foreman and SCRE showed that they had not received such a schedule. They referred the inspector to the lubrication coordinator for such informatio Lubrication records for the SI and AFW pumps were not available at the beginning of this inspectio Upon further investigation, the licensee determined, by other methods than prescribed in BwAP 370-1, that the lubrication for the SI and AFW pumps were curren *
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4 The lubrication program does not include provisions for documentation of the quantity of lubricant used, performance of post lubrication check (i.e., visual inspection for leaks, verify proper
! oil level, running of equipment as appropriate), authorization by the shift engineer to perform this preventive maintenance lubrication activity and retention of quality assurance records providing documentation of lubrication activitie Due to the extensive disregard for the lubrication program administrative i requirements the licensee should reevaluate the effectiveness of this progra ThIsfailuretofollowtherequirementsofBwAP370-1andto provide adequate controls for lubrication activities is considered a violation of 10 CFR 50, Appendix B, Criterion V (456/87007-06(DRP)).
l 1 Startup Test Witnessing and Observation The inspectors witnessed performance of portions of the following Unit 1 l startup test procedure in order to verify that testing was conducted in l accordance with the operating license and procedural requirements, test l data was pro)erly recorded, and performance of licensee personnel l conducting t1e tests demonstrated an understanding of assigned duties and l
responsibilities:
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RD-33 Rod Drop Measurement No violations or deviations were identified.
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19. Operational Safety Verification The inspectors conducted routine plant tours during the inspection period to make an independent assessment of equipment conditions, riant
! conditions, construction activities, security, fire protection, general personnel safety, housekeeping, and adherence to applicable regulatory requirements. During the tours, the inspectors reviewed various logs, daily orders int thedaymeetIngs,erviewedpersonnel,attendedshiftbriefingsandplanof witnessed various construction work activities, and i independently determined equipment status. During the shift changes, the l inspector observed operator and shift engineer turnovers and panel l walkdown During the inspection period, Unit 1 1roceeded from Mode 5 (cold shutdown) to Mode 3 (hot standby). T1e inspectors verified that all applicable requirements for Unit 1 were met during this period which
, included periodic checks of the locked valves for boron dilution l prevention, i
l These reviews and observations were conducted to verify that facility l operations were in conformance with the requirements established under i technical specifications, 10 CFR, and administrative procedure No violations or deviations were identified.
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20. Unit 2 Plant Tour The inspector observed work activities in progress, completed work and plant conditions during general inspections in Unit 2 work areas.
l Observation of work included cable trays, junction boxes, pipe support welding and mechanical equipment. Particular attention was given to l material identification, nonconforming material identification and I
housekeeping. The inspector reviewed work activities by reviewing
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travelers while touring the plant. These travelers pertained to l electrical, piping, and hanger installation.
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! No violations or deviations were identifie . Report Review During the inspection period, the inspector reviewed the licensee's Monthly Operating Reports for January and February 1987. The inspector confirmed that the information provided met the requirements of Technical Specification 6.9.1.8 and Regulatory Guide 1.1 The inspector also reviewed the licensee's Monthly Plant Status Reports for Janrary and February 198 No violations or deviations were identifie . Meeting, Training, and Other Activities Systematic Appraisal of Licensee Performance (SALP) Meeting On March 17, 1987, the Systematic Appraisal of Licensee Performance (SALP) meeting was held at the Region III Office in Glen Ellyn, Illinoi The purpose of the meeting was an opportunity for NRC representatives to discuss the licensee's performance for the period of December 1, 1985 through November 30, 198 It was also an opportunity for the licensee representatives to provide comments. The SALP 6 report is documented as Inspection Report 50-456/87001; 50-457/87001.
l Plant Status Meeting
! A meeting was held on April 6, 1987 between the licensee's Project l Manager,theRegionIIIProjectDirector,andmembersofeachoftheir
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staffs. The purpose of the meeting was for the licensee to provide an update on the status of Units 1 and 2. The meeting was also an opportunity to discuss the licensee's list of items that must be dispositioned prior to Unit 1 Startu . Open Items Open items are matters which have been discussed with the licensee, which will be reviewed by the inspector and which involve some action on the part of the NRC or licensee or both. Open items disclosed during the inspection are discussed in Paragraphs 14 and 1 .
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24. Unresolved Item i
Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, violations, or deviations. Unresolved items disclosed during the inspection are discussed in Paragraphs 2 and 1 . Exit Interview The inspector met with licensee and contractor representatives denoted in Paragraph 1 during and at the conclusion of the inspection on April 9, 1987. -The inspector summarized the scope and results of the inspection and discussed the likely content of this inspection report. The licensee acknowledged the information and did not indicate that any of the information disclosed during the inspection could be considered proprietary in nature.
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