IR 05000456/1986039
| ML20215M986 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 10/17/1986 |
| From: | Gardner R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20215M982 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-1.A.2.1, TASK-1.C.1, TASK-1.C.4, TASK-1.C.5, TASK-2.E.4.1, TASK-2.F.1, TASK-2.F.2, TASK-2.G.1, TASK-TM 50-456-86-39, 50-457-86-30, NUDOCS 8611040130 | |
| Download: ML20215M986 (24) | |
Text
.
U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Report No. 50-456/86039(DRP); 50-457/86030(DRP)
Docket Nos. 50-456; 50-457 Licenses No. CPPR-132; CPPR-133 Licensee:
Commonwealth Edison Company Post Office Box 767 Chicago, IL 60690 Facility Name:
Braidwood Station, Units 1 and 2 Inspection At:
Braidwood Site, Braidwood, IL Inspection Conducted:
August 3 through October 8, 1986 Inspectors: NRC T. M. Tongue W. J. Kropp T. E. Taylor M. J. Farber E.G.&G., Idaho J. Townsend R. Larson B. Barnes
/0 7&
'
Approved By:
R N.
ne, Chief Braidwood Project Section Date Inspection Summary Inspection on August 3 through October 8, 1986 (Report No. 50-456/86039(DRP);
50-457/86030(DRP))
Areas Inspected:
Routine, unannounced safety inspection of licensee action on previously identified items; construction worker concern; allegations;
'
headquarters request; plant procedures; NUREG-0737; implementation of strike plans; events occurring during the inspection; technical specifications; RTO-RTOC; plant tours and independent assessment.
Results:
No violations or deviations were identified.
8611040130 861021 PDR ADOCK 05000456 G
.
r
.
DETAILS
,
1.
Persons Contacted Commonwealth Edison Company (CECO)
Corporate Personnel A. Miosi, Nuclear Licensing Administrator Braiuwood Personnel M. J. Wallace, Project Manager
- C. W. Schroeder, Station Services Superintendent
- D. L. Shamblin, Project Construction Superintendent
- P. L. Barnes, Regulatory Assurance Supervisor
- G. E. Groth, Assistant Construction Superintendent M. E. Lohmann, Assistant Construction Superintendent
- E. E. Fitzpatrick, Station Manager
- L. M. Kline, Regulatory Assurance Group Leader C. J. Tomashek, Project Startup Superintendent
- D. E. Paquette, Maintenance Assistant Superintendent
D. E. C'Brien, Operations Assistant Superintendent R. Legner, Senior Operating Engineer G. Masters, Operating Engineer R. Ungren, Operating Engineer F. Willaford, Security Administrator M. Andrews, Station Chemist R. Lemke, Technical Staff Supervisor G. Nelson, Assistant Technical Staff Supervisor
'
- R. E. Aker, Radiation-Chemistry Supervisor T. Keith, Lead Health Physicist
- T. W. Simpkin, Regulatory Assurance T. E. Quaka, Site Quality Assurance Superintendent
- R. D. Kyrouac, Station Quality Assurance Supervisor T. Meyer, Station Fire Marshall G. F. Marcus, Assistant to Manager Quality Assurance
- L. E. Davis, Assistant Superintendent - Technical Services
- D. L. Cecchett, Regulatory Assurance
- A. J. D' Antonio, Regulatory Assurance Supervisor
- J. K. Jasnosz, Regulatory Assurance S. H. Stapp, Quality Assurance (0perations)
M. Takaki, Quality Control Supervisor K. M. Hall, Nuclear Services Health Physics S. M. Stevenson, Nuclear Services Health Physics A. M. Padleckas, Construction Quality Assurance S. C. Hunsader, Construction Quality Assurance
- R. Yungk, Operating Engineer J. Phelan, Project Field Engineer - Electrical Supervisor D. Tier, Project Construction E. Netzel, Quality Assurance Supervisor
-.-
-
- - -
-
'
.
B. Betourne, Construction Quality Assurance
.1 n Anware Chift Enninone E 1 U$gbe"PCd'E$gi$e N Mechanical S. Skrentny, Project Construction M. DeBoard, Project Startup
- A. Chomacke, Nuclear Safety
- P. Holland, Regulatory Assurance
- 8. B. Stephenson, Manager Nuclear Safety L. K. Comstock R. Seltman, Quality Assurance Manager C. W. Hart, Licensing Supervisor Sargent & Lundy D. J. Raef, Project Leader Electrical Nova Power T. Lewis, Startup Staff The inspectors also talked with and interviewed other licensee employees, including members of the technical and engineering staffs, startup engineers, reactor and auxiliary operators, shift engineers and foremen, electrical, mechanical and instrument personnel, cantract security perscnnel, and construction personnel.
- Denotes those attending one or more exit interviews conducted on August 14, 21, 28, September 4, and October 8, 1986, and informally at various times throughout the inspection period.
2.
Licensee Action on Previously Identified Items a.
Open Items (Closed) 456/84042-08; 457/84038-08:
CECO had not documented their review of nonconformance reports (NCRs) with regards to 10 CFR 50.55(e) reportability criteria.
Further review by the inspector identified that INPO had earlier reached the same conclusion.
The construction superintendent indicated that CECO had reviewed each NCR for reportability, but had not documented these reviews properly.
To resolve this issue, Ceco reviewed all past NCRs including those handled by the four CECO subcontractors listed below and enhanced their nonconformance review systems to address the NRC concerns.
(1)
G. K. Newberg Co.
(2)
L. K. Comstock Co.
.
_ _ - -
_
.
.
__ --.
__
._ -
, _
_
~
.
(3) Phillips Getschow Co.
(4) Pullman Sheet Metal Co.
CECO conducted the review using the reportability criteria listed in Braidwood Project.9rocedure PM-04 and Procedure PCD-23 describing responsibilities and actions required for CECO's processing of subcontractor NCRs.
CECO's system for reviewing NCRs and those used by CECO subcontractors have been significently improved since this concern was identified.
This item is closed.
(Closed) 456/84042-10; 457/04038-10:
L. K. Comstock (LKC)
electrical cable tray rework reports require component removals, but do not require subsequent component reinstallations after rework is complete.
The inspector felt that the existing system needed correction to assure reinstallation of components removed for rework.
To correct this problem, LKC has implemented the following:
(1) LKC Memorandum E85-02-05 clarified the intent of LKC Procedure 4.3.24 by directing engineers to include both removal and reinstallation of an item on the same rework report.
This procedure also requires that a new rework report be initiated if timely reinstallation is not possible as a part of the first rework report.
(2) LKC Procedure 4.13.3, " Area Completion / Turnover."
(3) LKC Procedure 4.13.2, " System Completion / Turnover."
These LKC procedures adequately address reinstallation of components removed for rework and implementation of these procedures resolves this concern.
This item is closed.
(Closed) 456/85007-04; 457/85007-04:
The inspector requested verification of proper environmental qualification for the three containment spray pump and motor issues listed below:
(1) Temperature and radiation exposure limits.
i (2)
Lower pressure limit of -0.50 inches of water.
(3) Auxiliary components.
CECO has provided the following item by item response to the above three issues:
(1) Regarding temperature and radiation exposure limits, Sargent &
Lundy (S&L) has verified that all components of the containment l
!
t l
1
-
.
spray pumps and motors, as listed in the bill of materials (P:g: 39, T:b G Of the report Oudited), re qu:lified te the
environment listed in FSAR Table 3-11-2, Zone A13C.
Several reports are referenced to substantiate the components qualifications to the required limits.
(2) Regarding the lower pressure limit of -0.50 inches of water, since this pressure level is equivalent to only -0.018 psig, this variation is felt to be within the limits of normal atmospheric pressure variation and is of no consequence.
Hence, this is not considered a valid or realistic testing requirement.
(3) Regarding auxiliary components, the RTD, heater, and associated terminal blocks are not required for the motor to perform its safety function.
These items are classified as non-class 1E; therefore, environmental qualification is not required.
The seismic qualification report for the motor is a proprietary Westinghouse document; however, S&L's review of this report revealed that it included qualification of the junction box and was determined to be acceptable.
This is documented in S&L CQD File CQD-003432, dated August 16, 1982.
The above three responses are considered adequate to resolve the questions raised by the inspector; this item is closed.
(Closed) 456/85008-11(DRS):
Verification of proper installation of containment spray pump impellers.
The inspector reviewed Field Change Orders (FCOs) 1CS-20891 and 1C5-20024 where measurements showed that the A pump impeller was smaller than the B pump as required.
The inspector also reviewed a November 5, 1985, memorandum which outlined three action items to ensure that the impellers are not switched in the future.
The items are as follows:
(1) A letter, identifying the A pump as having a smaller impeller and cautioning against switching impellers, was to be entered in tne pump instruction manuals.
(2) A similar caution was to be entered in the pumps' Maintenance History files.
(3) Identification tags were to be affixed to each pump.
In Inspection Report 456/86016(DRP); 457/86014(DRP) the inspector determined that Item (3) had not been completed.
During this inspection, the inspector verified the identification tags specifying the pump number and impeller diameter were affixed to each containment spray pump.
This item is closed.
(Closed) 456/85015-04; 457/85016-04:
A previous inspection identified a concern with several nonconforming items that were accepted based on the disposition "USE-AS-IS" with no justification for ameeptance.
.- -
_
..
.
.
Copies of the nonconformance reports referenced in the original in:pecticn were reviewed at the Sargent & Lundy field effice.
.
Justification for all the nonconformance reports reviewed were filled out in the block entitled " Explain Justification," or by attachment to the nonconformance reports. This item is considered closed.
(Closed) 456/85029-01: This item was left open pending further review of site QA participation in flushing activities.
Project Startup Procedure PSU-200, Revision 3, was revised to include a step to notify QA on a procedure hold point.
Since the previous inspection, approximately 20% of the hold points in the flush procedures were verified by QA.
During this review, it was found that several of the procedure hold points were being waived because the flushes were being performed on the backshift when QA coverage was not available.
The licensee has since identified QA personnel available to witness hold points on the backshift.
The inspector verified that the deficiencies which were identified have been corrected.
This item is considered clcsed.
(Closed) 456/85032-03; 457/85031-03:
L. K. Comstock (LKC) Quality Control Supervisors were not certified as Level II inspectors in the
'
specific areas they were supervising.
An allegation had previously been received (March 13,1985) which also pertained to LKC QC supervisors not being certified as Level II inspectors in the areas they were supervising.
This allegation (RIII-85-A-0062) was closed in Inspection Reports No. 456/85021; No. 457/85022.
Therefore, this item is considered closed based on the results of the NRC review of that allegation.
(Closed) 456/85038-03; 457/85037-03:
The licensee was requested to evaluate their justification for not including the high pressure emergency core cooling systems (ECCS) in their requirements for predetermined torque for flange bolts that is to be specified on the Mechanical Joint Checklist.
The licensee reviewed the high pressure ECCS and determined that the flanged joints are not subjected to widely varying pressure /
temperature conditions during normal operation. The licensee does not consider that the normal operational loads would dictate the need for the measured verification of actual bolt stresses and that " standard mechanical practices," along with hydrostatic testing, would be adequate.
This item is closed.
(Closed) 456/85038-06; 457/85037-04:
L. K. Comstock (LKC) issued Nonconformance Report (NCR) 4302 which identified concrete expansion anchors (CEAs) installed in several locations without quality control inspections.
The final disposition of this NCR was identified as an open item pending licensee corrective action and subsequent NRC
,
-
.
.
-
-
.
_
___
._
._.-_..
. _ -
-.
-
.
review.
The inspector reviewed the corrective action for NCR
.
- n n. A f a nn.d. i. t.
.n e.c.a. n. t. s h l. a..
T. h.a. e.n. e. v. a.r.t. i. u. e.me t. i. n..n.e.n. n.e. i. e. t.o.d a. f.
A'In?
.
. -.
.
-
.
-
No further action is required for CEA travellers numbered 0001 through 17143.
These CEA travellers were included in the CEA reinspection program which is included in the scope of 10 CFR 50.55(e) No. 84-17.
This 50.55(e) is currently open and will be evaluated by a regional specialist.
CEAs with' travellers sequentially numbered higher than 17143, which were not inspected by Quality control, will have a revision to the CEA traveller prepared by LKC's engineering department.
This revision will require a quality inspection of the CEAs as identified on the revised traveller.
CEAs installed with no traveller initiated will have travellers initiated by LKC's Engineering Department and will be inspected by Quality Control.
The area turnover process will identify those CEAs installed with no documentation.
(Closed) 456/85051-01; 457/85049-01:
During a previous inspection, the inspector reviewed LKC procedures and identified three concerns.
(1) LKC Procedure 4.1.1, Revision C, " Site Organization Chart," did
.
not document the actual Quality Control / Quality Assurance organizatioa in place.
Also, Procedure 4.2.1, Rev. C,
" Position Delineation," required updating to define the actual responsibilities of personnel.
LKC memorandum, QA85-12-35, response stated that the procedures were being revised to reflect LKC's current organization.
This was done in Procedure 4.1.1, Revision D, and 4.1.2, Revision D, dated January 15, 1986.
This inspector's review of the open item identified differences in the site organization chart (Procedure 4.1.1, Rev. D) and the position delineation (Procedure 4.1.2, Rev. D).
In discussing these differences with LKC personnel, it was learned that Revision E had just been released.
A review of Procedures 4.1.1, Revision E, and 4.1.2, Revision E, dated July 17, 1986, determined that the site organization chart and position delineation both agree.
(2) LKC Procedure 4.14.1, " Internal Audit Program," Paragraph 3.3.1, allowed limited activity procedures to be audited on a 24 month schedule, while Regulatory Guide 1.144 states at least annually.
Procedure 4.14.1, Paragraph 3.3.1, was revised (Rev. C, dated February 13,1986) to perform a complete audit annually.
(3) A difference in audit frequency was noted between LKC Procedure 3.1.4, Paragraphs 2.3 and 2.4, and LKC Procedure 1.0.1, Paragraph 4.14.
-
.
Procedure 1.0.1, Paragraph 4.14, was revised (Rev. A, dated Octnhor 1.19AM to "at least once in a 12 month nariod." to comply with Procedure 3.1.4.
~
~
The above concerns were addressed by revisions in the procedures.
These revisions are satisfactory to close this item.
(Closed) 456/85053-05:
An inspector determined that the diesel AFW pump battery bank voltage was being checked with the system battery charger operating.
The voltage measureaent taken with the battery charger operating does not properly diagnose the operability status of the battery bank.
The current revision of Procedure Bw05 7.10.1.3.1-1 has corrected this problem by only permitting measurement of the battery bank voltage without the battery charger in the circuit.
This item is closed.
(Closed) 456/85053-09; 457/85051-04:
Discrepancies with safety injection drawing and procedures.
The original inspection revgaled that accumulator vent control valve 15I0943 was not shown on the drawing and was not installed in the system. A followup inspection verified that the valve was installed and that the drawing was revised to show the valve.
This item remained open pending closure of deficiencies SI-12-227 and SI-12-228 and issuance of the revised prestart mechanical lineup.
Deficiencies SI-12-227 and SI-12-228 were closed on August 21, 1986, and February 13, 1986, respectively.
Bw0P-SI-M1, Revision 1, was approved on July 29, 1986, listing the missing valves.
This item is considered closed.
(Closed) 456/86007-02:
Safety Injection Check Valve Operability and leakage test, BwPT-SI-13, resulted in the 1750 psig Safety Injection Header Relief Valve, 15I8851, lifting.
Since the SI pump discharge pressure at the time should have been no greater than 1600 psig, the source of pressure necessary to lift that relief appeared to have been the Reactor Coolant System (RCS) which was at approximately 1840 psig.
The licensee's findingt indicated significant back leakage (RCS into the SI system) through several check valves and the SI pump operating in a recirculation (pump bypass) condition apparently resulted in the SI system pressure increasing to the SI header relief valve (1SI8851) set point.
Test engineers stated that increasing the RCS pressure to 1900 psig (within test specification)
eliminated the lifting of the relief valve 1SI8851.
The increased RCS pressure increased the differential across the check valves, thereby decreasing the back leakage through the check valves and hence alleviating the SI system pressurization.
The 1SI8851 relief valve was serviced and recalibrated and the ISI8853 A and B valves were checked to ascertain that they had not lifted.
The inspector has reviewed and concurs with the licensee's findings.
The inspector notes that the check valves involved are to be or have been serviced and during normal operation will have the full RCS pressure across the valves.
.
.
_
.
_ _ _
_-
.. _ _
_
-
..
The licensee's fin'ings and actions are considered sufficient and d
y this item is closed.
b.
Unresolved Items j
l (Closed) 456/78006-03; 457/78006-03:
Chloride Content for Category I Concrete Structures.
Chloride content was found to be slightly
above the level recommended in the applicable ASME/ACI standards.
This item'was submitted to NRR for review and the staff deleted the earlier imposed requirements of yearly inspections to establish:
4=
.(1) accumulation of chloride deposits; (2) discoloration of j
concrete;-(3) cracks; and-(4) changes from previous observations.
This was transmitted to the licensee via NRR letter dated September 11, 1986.
This issue was previously addressed in Inspection Report 456/84019; 457/84018 in August 1984.
The
,
inspectors concur with the NRR assessment and this issue is considered closed.
(Closed) 456/84009-10; 457/84009-10:
It was previously identified that LKC did not have a final walkdown procedure for electrical hardware. A final walkdown QC inspection prior to licensee turnover is helpful in assuring,that all electrical installations are acceptable and that previously installed items have not been damaged
.
or removed.
Uncontrolled removal has been a problem for LKC as
,
previously identified in Inspection Report 456/84006; 457/84006.
'
To resolve _this issue, LKC has generated Procedure 4.13.3, Revision B, " Area Completion / Turnover," which supplements related LKC Procedure 4.13.2, Revision A, " System Completion / Turnover."
- LKC will implement these two procedures.
This item is closed.
,
l.
(Closed) 456/84042-07; 457/84038-07:
Prior to February 21, 1984, I
L. K. Comstock personnel closed Inspection Correction Reports (ICRs)
without documenting on the ICR the action (s) performed to resolve
'
the identified deficiency.
LKC determined that there were 956 ICRs
.
which had been closed without documenting the corrective action.
!
Eighty of these ICRs were reinspected to determine if the deficiencies were corrected.
The results of these inspections
-
indicated that the deficiencies were corrected even though the corrective action was not documented on the ICRs.
To determine the l-effectiveness of these reinspections, the inspector selected six of
!
these 80 ICRs and performed independent reinspections.
The results l
of the inspectors reinspections substantiated LKC's conclusions.
I The inspector also reviewed 10 ICRs issued after February 21, 1984 and verified that the corrective actions are being documented on the ICRs.
This item is considered closed.
(Closed) 456/85029-02:
Based upon a review of selected test deficiency reports, the inspector was concerned whether quality
'
assurance was appropriately involved in the test deficiency process.
Review of 15 such deficiency reports (Form BwSM 4-1) indicated
,
I
-
i
-
.
-.
.-
.
.
.
-
-.
'
..
that quality assurance was involved and rigorously implementing L
the ranniromantc - nf 9er tinn 8; O nf ROT-U and cunnlementinn I
GuideIines for QA Review During. Deficiency Clos'e-Out." Seven of
~
the 15 deficiency reports reviewed.had been rejected by QA as needing additional work before QA would sign.them off as being
,
resolved.
This appeared to be ample evidence that QA was appropriately involved and acting in accordance with approved procedures.
This item is considered closed.
(Closed) 456/85031-01:
During a previous inspection, a drain wire
,
termination was not shown on the drawing.
Also, LKC ICR No. 7000 along with the cable de-termination and re-termination cards
-
referenced the wrong drawings.
LKC wrote Nonconformance Report 4307, Rev. O, to correct this item.
The inspector reviewed the nonconformance report, the revised LKC ICR No. 7000, and the cable de-termination and re-termination cards reflecting the correct
drawings.
Drawing 20E-1-4105M, Revision J, was changed to show the
cable termination.
This item is considered closed.
!
[
(Closed) 456/85038-04:
Cable IVV092 was not routed thru raceway
!
12026C as required by the cable pull card.
This cable pull had been
'
- accepted by a QC inspector as being correct on November 20, 1981.
'
The electrical contractor, L. K. Comstock (LKC) issued Nonconformance Report (NCR) 4837 to address the incorrect routing of
'
cable IVV092.
This NCR was closed on March 5, 1986.
To evaluate the performance of the QC inspector involved in the cable pull inspection of cable 1VV092, LKC reinspected 30 cables for correct cable routing out of the 681 cable pulls he inspected.
These
,
reinspections did not identify any other cables which were not i-routed correctly.
However, there were six cable pulls that were reinspected which had deficiencies other than incorrect routing.
There deficiencies were documented on LKC Inspection Correction
.
Reports (ICRs) for resolution.
There was no conclusive evidence that these deficiencies were missed during the original inspection
j.
or were a result of subsequent construction activities.
These six deficiencies were satisfactorily resolved as documented on the ICRs.
To determine the effectiveness of LKC's evaluation, the
inspector selected three cable pulls that were part of the 80 cable
!
pulls reinspected by LKC and performed an independent verification for correct routing. The inspector also selected two other cable
'
pulls inspected by the QC inspector that were not part of the LKC reinspections and verified they were properly routed.
These
'
reinspections by the NRC identified no problems and it. appears that LKC's evaluation of the QC inspector's performance on cable pull
-
inspections was effective.
This item is considered closed.
!
(0 pen) 456/85038-05: The weld inspections records for four cable
tray supports could not be located.
L. K. Comstock (LKC) performed
additional searches for these weld inspection records, but could not locate any documentation that would substantiate these four cable
tray supports were inspected for weld acceptability.
Therefore, LKC l
Quality Control inspectors performed weld inspections for these
!
!
,,
-, - _ _. _ _ _ _. _ _ - _, - _ _. _
_
_ _. _. _ _ _ _.., _ _ _ _. ~ _. _ _ _... _. _
.
supports and documented the results of these inspections on the
a,n,nrnnriato fnemc.
Hnwover, cinco tho fnpr hangorg with missing weld inspection documentation represented approximately 17% of the hangers originally inspected, this item will remain open until the completion of LKC's area walkdown activities.
These activities will then be evaluated to determine if these area walkdowns were effective in identifying any other missing weld inspections.
(Closed) 456/85057-01:
An unresolved item was issued after the inspector reviewed BwPT-SI-02, Revision 0, " Safety Injection Flow Balance," and could not determine that the test method actually verified the ability of the orifices by themselves to prevent pump runout.
The test method ran the pumps with the throttle valves in their throttled position such that flow was limited, not by the orifices themselves, but in combination with the throttle valves.
Since this did not appear to verify the design of the orifices, the adequacy of the test method and acceptance criteria in BwPT-SI-12 was unresolved pending further information from the licensee.
The inspector reviewed the SI and charging pump runout orifice test specifications, setup, and data (Test Report BwPT-SI-012),
all of which properly spec.ify the throttle valves to be wide open during the test.
These specifications are satisfactory and the data indicates that the tests were run as specified and that the runout orifices are satisfactory. Therefore, the original issue is effectively resolved.
During the review, the inspector found that the 1A charging pump and 1A SI pumps were tested (BwPT-SI-012) with different impellers than provided and as tested by the manufacturer.
The replacement impellers were, in each case, taken from the respective 28 pumps which have identical specifications, and virtually identical clearances as shown on the manufacturers pump test curves. The pump performance curves for the original and replacement 1A and 28 pumps have been compared and are sufficiently identical which indicates the impellers are sufficiently identical.
Tnerefore, any differences in the 1A SI and 1A charging pump performance relative to their specified performance characteristics is insignificant.
In view of the acceptable runout orifice tests and the acceptable performance of the 1A high head SI and the 1A charging pumps with the replacement impellers, this item is coasidered closed.
(Closed) 456/86025-02:
The inspectors concern was that the system voltage may not maintained at + 5% of the motor nameplate voltage as stated in the vendor motor manual, without impacting the
-
system voltage test results of the NUREG-0876 Item 8.2.4 required test.
This item was considered to be unresolved pending further review of a Vendor Motor Manual revision by Westinghouse Electric
__ __
_
_
-
.
.
.
-
Company. Westinghouse Electric performed calculations for the essential service water pump motors which demonstrated that a i 10% power supply voltage variation of the motor nameplate voltage is acceptable. The Instruction Leaflet, IL-5500A, page 23, Paragraph 6, was revised to allow the motor power supply to vary i 10% of the nameplate voltage.
This item is considered closed.
c.
10 CFR 21 Reports (Closed) 456/83002-PP; 457/83002-PP: Qualification of Viton elastomer as the seal material in recombiner application.
The inspector reviewed documentation substantiating the replacement of the Viton seals in the hydrogen recombiners.
The documentation included Field Change Order 10/G-24276, Equipment Installation Records for hydrogen recombiners 00G085A and 00G085B, and Purchase Order 298056.
The Viton seals were replaced with ethylene propylene and graphite seals.
The specific replacement seals used were a Parker Hannifin #848037-2, a Parker Hannifin #848037-20, and a Lamous Grafoil 2" Stainless Seal 1/16" thick.
The inspector reviewed a letter, dated May 28, 1984, from the vendor which identified the replacement seals.
The replacement seals utilized were the same as specified by the vendor.
This item is closed.
(Closed) 456/84001-PP; 457/84001-PP: An NRC inspector identified in Vendor Program Branch Report No. 99900054/83-01 that the APSC Palo Verde Nuclear Generating Station Unit 1 a.ainsteam valves did not meet the design specification 5% blowdown requirement.
Testing done by Dresser Industries (the valve manufacturer) and Wyle Laboratories resulted in Dresser Industries (DI) notifying utilities using DI Consolidated Model 3707RA MAPVPAV Safety Valves with forged bodies that the problem could be solved in one of two ways.
Ceco opted to solve the problem at Braidwood using the same method used at Byron (reference Byron Part 21 454/84-03-PP(DRP)).
More specifically, CECO implemented an adjustment of the upper and lower valve blowdown rings as recommended by valve manufacturer Dresser Industries.
The DI specified ring adjustments were based upon the results of a test program which empirically determined safety valve blowdown over a range of set pressures as a function of blowdown ring adjustment.
The DI specified adjustment provides a blowdown of 6% 1 3% for a given set pressure.
CECO implemented this work with Field Change Orders (FCOs) IMS-34800 thru IMS-34810 and IMS-34791 thru IMS-34799.
The last of these FCOs was signed off by Quality Control as having been acceptably completed on August 25, 1986.
This item is closed.
d.
Safety Evaluation Report Items (Closed) 456/8600009; 457/8600009:
This item pertained to the four Braidwood diesel generator's exhaust systems susceptibility to damage from tornado generated missiles.
There was an SER concern that tornado generated missiles could collapse or crimp and block the diesel engine exhaust stacks to such an extent that the diesel generator would fail to run properly.
-
. _.
-_
_
. _ _ _ _ _
.
.
_.
_
.
-
.
Ceco proposed implementing either one of the following two changes
-
to the NRC to resolve this issue.
(1) exhaust pressure relief via a tornado proof weighted damper system, or (2) strengthening the exhaust stacks to withstand tornado missile impact without unacceptable damage.
The NRC agreed that either proposed solution would resolve the issue.
CECO opted to use item (1) exhaust pressure relief scheme, but made a minor change by using a rupture disk in place of a weighted damper system.
Either method is acceptable. The rupture disks are installed downstream from the mufflers in a horizontal 32" pipe line (lines ID025AB32 and ID025AA32).
This was verified by a system walkdown where the pressure ratigg of the 24" diamete5 rupture disks was observed to be 3.5 psig @ 72 and 2.45 psig @ 650.
A Sargent &
Lundy (S&L) dgsign engineer indicated exhaust gas temperatures are less than 650 at this point in the system and that the actual exhaust system pressure was about 0.75 psig at this point in the exhaust system. This item is closed.
<
(Closed) 456/8600014; 457/8600014:
Ceco committed to install a unit common standby condensate cleanup system (CCS) to clean up
-
the condensate and feedwater systems during startup and for other anticipated operational occurrences.
The NRC reviewed and approved the proposed CCS and CECO has since installed and tested the system.
The CCS preoperational test report (CP-20) was reviewed and a plant walkdown was made to verify the system is in place.
This entire system has been turned over to CECO as fully complete and ready for startup.
This item is closed.
(Closed) 456/8600016; 457/8600016:
Verify that procedures are in place and training is conducted for station blackout.
The inspector verified that procedures for loss of all A.C. power were in the control room operation manuals and that operation personnel were aware of the procedure steps for station blackout.
The following procedures were reviewed:
BwCA 0.0, Loss of All A.C. Power; BwCA 0.1, Loss of All A.C. Power Recovery Without SI Required; and BwCA 0.2, Loss of All A.C. Power Recovery With SI Required.
Training on station blackout is performed in both the classroom and simulator phase of licensed operator training.
This item is closed.
e.
10 CFR 50.5E(e) Reportable Items (Closed) 456/82008-EE:
Before June 1980, not all safety related pipe hangers were required to be inspected at the time of installation for proper assembly, configuration, and welding.
In June 1980, construction practices were revised to require these
.
-
_
-
--
. -
.
_ _ _ _.
.
__- -
-
.
inspections on all safety-related hangers. A total of 3069 hangers installed prior to June 1981 had to be inspected in accordance with
'
~ QC Procedure B23 to assure that hanger assembly, configuration,.and welding were in accordance with applicable design drawings.
For Braidwood Unit 1 only, closure of NCRs 776, 6075, and PGCo NCR 2249 indicated satisfactory completion of this inspection effort.
Though Braidwood Unit 2 related PGCo NCR 2248 is satisfactorily closed,;PGCo NCR 6699 remains open precluding closure of Unit 2 related items.
For Braidwood Unit 1,.this item is closed.
!
(Closed) 456/86004-EE; 457/86004-EE:
Two engine failures occurred on the Unit 1 Byron Station 2A diesel generator.
Both failures were identical in nature and had the same root cause; cylinder
liner and head misalignment.
The failures have been satisfactorily
'
evaluated by the licensee and Cooper-Bessemer (engine manufacturer)
personnel.
The maintenance and corrective action procedures developed by the manufacturer have been shown to be effective.
These procedures address the proper installation and alignment of.
the KSV diesel engine cylinder liners, cylinder heads and rocker arms.
The procedures'were followed by Cooper-Bessemer and licensee
,
- personnal in servicing the Byron Station 2A diesel generator engine.
The manufacturer also uses and recommends using an engine analyzer, which can detect abnormal conditions in checking diesel engine rocker arm components.
An ENSPEC-1000; analyzer is available at the
- Byron Station.
The Braidwood maintenance specification requiring l
the.use of the analyzer is incorporated into the diesel generator service procedures (surveillance Procedure BwVS 8.1.1.2E-1 and E-2).
.
!
The corrective actions taken and the maintenance procedures
'
incorporated by the licensee are satisfactory and this item is closed.
,
(0 pen) 456/86005-EE; 457/86005-EE:
During ECCS full flow testing
.for Byron Station Unit'2 in February 1986, a 11/2 inch socket
welded-elbowona(Aloop)highheadcoldleginjectioncrackedand
'
began leaking.
The elbow was replaced and the test continued.
The examination concluded there were micro-cracks which had propagated through the wall.
-
Sufficient supporting information required to resolve this issue is not immediately available.
These items have been requested of the
!
licensee and will be resolved at a later date.
- The A. D. Miosi (Commonwealth Edison) letter to Mr. James G. Keppler (t,iRC), dated June 5, 1986, has been reviewed.
The item was i
discussed with Mr. J. Gavula of the NRC Region III office.
The subject piping was inspected on Braidwood Units 1 and 2.
i f
i
I
, - -
.--v-
,-
--.r.. -,, ~, - - - - -,
,,,.s
-
vg---
m
---wv----,,,..,--,-~--,,,------~n----v,,---,,-,,,,
.-e-.-
.---,-.w-.,-,.-,w-
--w-,-me-r-nn
-
-
.
The inspector has requested supporting analyses and results includina vibration test / analyses; pipina and support stress analyses as required to satisfactorily address and resolve the item.
This item will remain open pending further review.
f.
Generic Letter (Closed) 456/85005-HH; 457/85005-HH:
This generic letter informs PWR owners / operators of the NRC staff position resulting from evaluation of Generic Issue 22, " Inadvertent Boron Dilution Events" and the need for instrumentation to detect boron dilution in operating reactors.
In an April 18, 1985 letter from the Superintendent of the Braidwood Nuclear Station (John F. Gudac) to the Director of Nuclear Licensing (Dennis L. Farrar), Mr. Gudac responded to the generic letter by outlining the four operational situations listed below which could result in an inadvertent boron dilution event:
(1) Primary water could inadvertently enter the CVCS if the CV111A and CB111B valves were to stick open.
However, there is a positive alarm to alert the operator to a total make-up water flow deviation on the main control board.
There is also a make-up flow totalizer, audible to the SR0, which activates when primary water is making up to the VCT.
(2) An outsurge of the pressurizer when the pressurizer liquid boron concentration is less than the RCS boron concentration could result in a dilution transient.
BwCP 300-1 and BwCP PD-1 will provide for sampling boron concentration in the pressurizer at least daily, and any large difference would be resolved by activating the heaters in order to recirculate with pressurizer spray.
(3) The changing of CVCS or BTRS resin bed is another possibility leading to inadvertent dilution of the RCS.
However, the volume of diluted water added to the RCS would be minimal compared to the total volume of the RCS.
(4) RCS dilution could also result from startup of the RHR system.
The liquid in the RHR piping could have a lower boron concentration than the RCS.
The applicable operating procedure, Bw0P RH-6, calls for sampling the RHR fluids prior to aligning the RHR system to the RCS.
The above four operational situations and related explanations appear to completely address the NRC concerns as outlined in the (Revision 0, approved May 1, 1985) procedures cited above, BwCP PD-1 generic letter.
As a check on the and BwCP-300 (Revision 0, approved July 1, 1986) were reviewed and found to be in agreement with the once per day sampling rate described in the Item 2 response above.
Similarly, review of operating procedure Bw0P RH-6 l
l
l
.
.
(Revision 1, approved January 8, 1986) verified statements made in
.
Item 4 above regarding the operational procedures-used to ensure that the RHR system boron concentration equals or exceeds the boron
"
concentration in the reactor coolant system.
This issue is closed.
(0 pen) 456/85006-HH; 457/85006-HH:
" Quality Assurance Guidance for ATWS Equipment That Is Not Safety-Related." This issue.was discussed via telecon on August 21, 1986, with licensee personnel and the NRR Licensing-Project Manager.
NRR management agreed.that the licensee's previously submitted schedule for implementation of the requirements of 1988 is acceptable.
Previous NRC guidance provided a cut off date of July 1, 1989.
This Generic Letter will remain open until those requirements are completed.
No violations or deviations were identified.
3.
Construction Worker Concern On August 14, 1986, the senior resident inspector (SRI) (operations) was contacted by four construction electricians expressing concerns related to the security force.
The individuals felt that they had been intimidated by a security guards remarks; however, when questioned, all four stated that they had not been threatened directly.
They provided copies of written statements that had been presented to the licensee through their union steward and had not received a response (over two weeks).
After some discussion on the licensee's implementation of the security system, the SRI committed to follow-up on their concern.
On the following day, the SRI contacted the station security administrator and found that the contractors concern had been addressed and the guard involved had been counselled on the matter.
On August 18, 1986, the SRI met with three of the electricians, their union steward, and the L. K. Comstock project superintendent to discuss the resolution.
All expressed satisfaction with the results and the matter is considered closed.
No violations or deviations were identified.
4.
Allegations (Closed) RIII-85-A-0170:
The NRC received a multi part allegation that deficiencies existed in the procedure development, review, and approval process at Braidwood, procedure quality was poor as evidenced by an
'
inadequate procedure for filling the Primary Water Storage Tank, and that plant operators are not following the procedure and the make-up water preheater has been repeatedly thermally shocked. The specifics of the allegation were forwarded to Commonwealth Edison in a letter dated January 16, 1986 from C. E. Norelius to Cordell Reed.
The letter requested that Commonwealth Edison review the allegation, make a determination with regard to its substance, and document their review for inspection by the NRC at a later date.
Commonwealth responded in a letter dated February 25, 1986 from A. D. Miosi to James G. Keppler.
The Braidwood Station Quality First organization conducted the review of the allegation and determined that it was not substantiated.
The
.
._ -
.
inspector reviewed the Quality First. file on this matter and reviewed the station administrative proarams for Permanent Procedure Preparation.
Revision and Approval (BwAP 1300-2, Revision 4) and Onsite Review of Procedures (BwAP 1205-2, Revision 3).
The inspector noted that the responses to all parts of the allegation were thorough; that the procedure development, review, and approval process at Braidwood was adequate; that the specific operating procedure identified by the allegation had been revised prior to Commonwealth Edison being informed of the allegation and was adequate; that the specific component, identified by the allegation as being damaged, had not been damaged because it had not been used; and that the system had been operated under the direction of the startup test engineer.
Further, it is noted that the Primary Water Storage system is not safety-related.
A detailed review of safety-related operating procedures is documented in Paragraph 6 of this report and in Report 456/86016; 457/86014.
5.
Headquarters Request By telephone request of August 15, 1986, the SRI (0perations) was requested to provide information related to 10 CFR 50.55(e) and NCR reports on the manufacture of the recycle holdup tanks, to the Vendor Inspection Branch (VIB) in NRC Headquarters. The SRI found that the tank material was rolled and some welding (nozzles and manways) was done by Chicago Bridge and Iron (CB&I) in Kankakee, Illinois.
The remainder of the fabrication was performed by CB&I onsite.
This information, including the names of CB&I and CECO contacts, was forwarded to Mr. Jeff Harper of the VIB on August 21 and 22, 1986 via telecons.
Information was requested by the NRR Braidwood Project Manager for NRR Engineering Branch reviews on the seismic monitoring equipment installed at Braidwood.
The resident inspectors walked down all installed units and verified the model, designation, and locations against station plans.
All components were supplied by Terra Technology Corporation, 3860 148th Avenue N.E., Redmond, Washington, 98052.
The models supplied are:
five Fluid Damped Service, Model SSA-302; three Passive, Model PRA-103; and one Geophone, Model SP-115CT.
The passive units were not. installed at the time of the inspection; however, dates and planned locations were verified.
This information was provided to NRR via telecon on August 26, 1986.
No violations or deviations were identified.
6.
Plant Procedures
/
During the inspection period the review of station procedures os continued; refer to Inspection Report 456/86016(DRP); 457/86014(DRP),
Paragraph 9, for general discussions.
Operating Procedures The inspector reviewed the index of plant operating procedures for completeness.
The inspector also reviewed a significant sample of procedures (in part or in their entirety) related to operations from
_
_
_.
_
., _ _
_
.
. -
-
.
administrative procedures, general plant operating procedures, procedures for startup, operation and shutdown of safety-related n
systems and surveillance procedures.
Ine inspector verirled that
'
the sampled procedures were in the format specified in the station administrative procedures and were technically adequate to accomplish their stated purpose.
This included listing proper reference documents, that checklists were provided as necessary, requirements exist for proper removal and returning equipment to service, that Technical Specification LCOs are considered, if temporary procedure changes were in effect, that they were appropriate and properly implemented, and appropriate approvals are required or other necessary prerequisites, precautions, or limitations.
This included verification that the procedures contained applicable operating limits and they were consistent with appropriate regulatory requirements; that startup procedures adequately addressed plant conditions, prerequisites, and limits in effect; that power operating procedures gave adequate surveillance to the process to insure integrity of operation; that shutdown procedures properly guide activities during and after a controlled shutdown or scram; that reference to other procedures is appropriate, and; that system procedures adequately describe steps to insure safe reactor operations.
This review did not constitute a complete step-by-step evaluation or verification, but was a representative sample of the operating procedures in place.
The inspector reviewed procedures under the subjects of administration (AP), general operating (GP), operating abnormal (0A), system operation (0P), and operating surveillance (0S).
The following is a iist of those procedures sampled:
Procedure Title Rev. No.
BwAP 300-1 Conduct of Operations
BwAP 320-1 Shift Manning
BwAP 335-1 Operating Shift Turnover
and Relief BwAP 335-1A1 Turnover Sheets, Shift Engineer
BwAP 335-1A2 SCRE Turnover
BwAP 335-1A3 SF Turnover
BwAP 335-1A4 Aux Services (Radwaste) Foreman
BwAP 335-1A5 Center Desk R0 Turnover
BwAP 335-1A6 Reactor Operator Turnover
BwAP 335-1A7 Electrical Operator Turnover
BwAP 335-1A8 Equipment Attendant
BwAP 335-2 Operator Watchstanding Practices
BwAP 340-1 Use of Procedures for Operating
Department BwAP 340-1T1 BwST Log
BwAP 340-3 Use of Mode Change Checklists
BwAP 360-1 Operating Department Surveillance
Program
-.
. -.
.
.
.. -
_
-
.
BwAP 360-1A1 Operating Department Daily
Transmittal Sheet BwAP 360-1A2 Unscheduled Surveillance Form
BwAP 360-1A3 Certified Instrument Use
'
Tracking Form BwAP 360-1A4 Operating Department Surveillances
Part Due BwAP 380-1 Green Board Concept-Control Panels
BwAP 380-2 Handling of Long-Term Annunciator
Alarms BwAP 380-2T1 Valid MCB Alarm Status:
Panel #
BwGP 100-1 Plant Heatup
BwGP 100-2 Plant Startup
BwGP 100-3 Power Ascension 3% to 100%
BwGP 100-Al Modes 5 to 4 Checklist
BwGP 100-A2 Control Board Lineup
BwGP 100-A4 System Lineup Checklist
BwGP 100-T1 Heatup
BwGP 100-T2 Startup
Bw0A ELEC-1 Loss of DC Bus Unit 1
18w0A ELEC-5 Local Emergency Control of Safe
Shutdown Equipment 18w0A ENV-3 Failure of the Cooling Lake Dike
Unit 0, 1 OBw0A REFUEL-l'
Fuel Handling Emergency Unit 0,
1, 2 18w0A SEC-8 Steam Generator Tube Leak - Unit 1
Bw0P AN-1 Annunciator System Startup
Bw0P AN-2 Annunciator System Shutdown
Bw0P AN-5 Ground Isolation For The Plant
Annunciator System Bw0P AN-El Unit 1 Electrical Lineup
Bw0P AN-E2 Unit 2 Electrical Lineup
Bw0P AP-16 Isolating System Auxiliary
Transformer (SAT) 142-1 With Unit 1 UAT De-Energized Bw0P AP-17 Restoring System Auxiliary
Transformer (SAT) 142-1 to Service Bw0P CC-2 Component Cooling System Startup
and Operation Bw0P CC-3 Component Cooling System Shutdown
Bw0P CC-14 Post Loca Alignment of the CC System 0 Bw0P CS-5 Containment Spray System
Recirculation to the RWST Bw0P CV-6 Dilution
Bw0P CV-6-T1 Boration/ Dilution Table
Bw0P CV-6-T2 Dilution Flow
Bw0P CV-7 Boration
Bw0P CV-7-T1 Boration Flow
Bw0P CV-8 Alternate Dilution
,
Bw0P DC-5 125V DC ESF Battery Equalization
'
,
-
.
Bw0P DC-5A1 125V DC ESF Battery Equalization Log 0 Bw0P DC-5T1 125V DC ESF Battery Equalization
Data Sheet Bw0P DG-1 Diesel Generator Alignment to
Standby Condition Bw0P DG-13 Trouble-Shooting Diesel Generators
Bw0P DG-13T1 Diesel Generator Troubleshooting
Checklist Bw0P IP-5
- Removing AC Input From an
Instrument Bus Inverter BwCP MP-5 Isolating Unit 1 Main Generator,
Main Power Transformer 1E and 1W, and Unit Auxiliary Transformers 141-1 and 141-2 Bw0P MP-5T1 Protective Card Placement Worksheet 51 Bw0P RH-6 Operation of the Residual Heat
Removal System for Plant Cooldown Bw0P SI-1 Safety Injection System Startup
Bw0P SI-2 Safety Injection System Shutdown
Bw0P SI-M1 Operating
Bw0P SX-3 Essential Service Water Pump Startup 9 Bw0P SX-4 Essential Service Water Pump
Shutdown Bw0P VC-5 Placing the Control Room HVAC
System Makeup Filter Train and Recirculation Charcoal Absorber in Operation 1Bw0S DC-W3 250V DC Battery Bank 123 - Weekly
Surveillance 18w0S DC-W4 24V DC Auxiliary Feed Pump 1B
Battery Bank A and B Weekly Surveillance 1Bw05 NR-1 Power History Hourly Surveillance
18w0S XLE-R1 Locked Equipment 18 Month
Surveillance 18w0S 1.3.6-1 Control Rod Insertion Limit
Surveillance 18w0S 2.1.1.a-1 Axial Flux Difference Weekly
Surveillance 1Bw0S 1.1.1.1.e-1 Shutdown Margin Surveillance
1Bw05 4.9.3.2-1 Reactor Coolant System Overpressure 51 Protection Systems Operability Monthly Surveillance 1Bw05 4.9.3.3-1 Reactor Coolant System Overpressure 51 Protection Systems Operability Non-Routing Surveillance 1Bw05 7.1.2.1.a-1 1A Train Auxiliary Feedwater Monthly 52 Surveillance 1Bw0S 7.1.2.1.a-2 IB Train Auxiliary Feedwater Monthly 52 Surveillance 18w0S 7.3.2.a-1 Component Cooling Water Pump
Operability Monthly Surveillance
-
.. - -
.
-
18w0S 8.1.1.2.a-1 1A Diesel Generator Operability
Monthly Surveillance 18w0S 8.1.1.2.a-2 IB Diese1' Generator Operability
Monthly Surveillance 18w0S 1.1.1-la BorationContgolShutdownMargin
T
< 200 F LC0AR Action Chart 18w0S 1.1.2-la LC0gGReactivityControlSysgems
> 200 F 18w05 4.4-la LC0ARPressurizerRhYkeTValves
1Bw0S 4.9.2-la LC0AR Pressurizer Temperature Limits 51 The inspector found that the plant operating procedures were prepared to adequately control safety-related operations within applicable regulatory requirements.
No violations or deviations were identified.
7.
Status of Licensee Programs for Compliance With Specific Sections of NUREG-0737, " Clarification of TMI Action Plan Requirements Subsequent to the Three Mile Island Unit 2 accident certain NRC requirements were developed, from review of the incident, to reduce the possibility of any reoccurrence of such an accident.
These requirements are delineated in NUREG-0737, " Clarification of TMI Action Plan Requirements." The inspector's review of activities relative to compliance with NUREG-0737 are as follows:
I. A.2.1 Immediate Upgradinc of Reactor Operator (RO) and Senior Operator (SRO) Training anc Qualification This item was initially reviewed during Inspection Report 456/86031; 457/86024.
Further review of shift manning requirements and licensee's operator training status shows the licensee has sufficient R0 and SR0 personnel to meet the requirements of NUREG-0737 and Technt
Specifications Sections 6.3 and 6.4 tor Unit 1 startup.
em is considered closed.
I.C.4 Control Room Access Through discussions with licensee personnel and further review of applicable procedures and NUREG-0737 the inspector has resolved concerns relative to personnel qualification for control room access control.
Pending implementation of applicable BwAP's, this item remains open.
I.C.5 Feedback of Operating Experience The inspector has reviewed approved Procedure BwAP 1260-1, " Operating Experience Review Program." The licensee has submitted a September 2, 1986 date for formation of the onsite Nuclear Safety Group (0NSG).
This item remains open pending formation of ONSG and revision of the Organization and Administration Manual for the licensee's Nuclear Safety Department.
- _
_
_
__.
_
.
_ _. - -
.
..
o
.
L I.C.1.1 Short-Term Accident Procedures Review for Small Break LOCA The review of accident procedures for a small break LOCA accident
.was reviewed and found to be adequate.
Inspection Report 456/86016; 457/86014 documents the review of BwEP-1, " Loss of Reactor or Secondary Coolant." This item is considered closed.
II.E.4.1 Dedicated Hydrogen Penetrations The review of this item included review of off gas system procedures, discussions with the system test engineer, and review of NUREG-0876 and NUREG-0737.
The licensee's system configuration and operating procedures meet the requirements of NUREG-0737.
This item is considered closed.
II.F.1.2 D-E Accident Monitoring for Containment Pressure, Water level, and Hydrogen The inspector verified by visual inspection that control room indication exist for containment pressure, water level and hydrogea.
These installations complete the NUREG-0737 requirements for these indicators.
This item is considered closed.
II.F.2-1,2 Instrumentation for Detection of Inadequate Core Cooling The instrumentation for detection of Inadequate Core cooling (ICC)
is used to calculate the subcooling margin which is an iconic display on the SPDS system.
The inspector also reviewed a number of Braidwood Emergency Procedures (BwEP).
Procedures reviewed were BwEP-ES 0.1, 0.2, and 0.3. " Reactor Trip Response, Natural Circulation Cooldown, and Natural Circulation with Steam Void in Vessel.
All the reviewed procedures addressed the subcooling margin.
The inspector determined that the licensee's system and procedures for ICC monitoring are adequate.
This item is considered closed.
II.G.1 Power Supplies for Pressurizer Relief Valves, Block Valves, and Level Indicators Through discussions with the pressurizer system test engineer and review of applicable electrical drawings, the inspector has determined that the licensee meets the requirements of NUREG-0737.
The pressurizer relief valves, block valves, and level indicators are powered from emergency sources.
This item is considered closed.
No violations or deviations were identified.
8.
Implementation of Strike Plans During the inspection, on August 14, 1986, the senior resident inspector (operations) was informed of the potential for a strike by the security guards against Burns Security (Ceco contractor).
Initial indications were that the strike was to occur on August 16, 1986, if the contract
-
_
..
,
-
- - - - -.
_-
--.
- - - -. - - _
.,. _
_.
.-
-
..
.
.
vote was rejected by the guards.
Through interviews with licensee personnel, the inspector found that the licensee had a contingency plan and was ready to place it in effect if the strike had taken place.
Although the station is not a licensed facility, the security system was-
'being implemented in preparation for licensing.
The result of the vote-
.
was a rejection of the contract offer; however, the strike did not take place.
.
At the writing of this report, the guards had reentered nagotiations with Burns Security, but with a pocket strike option. The resident inspectors have followed the activities and passed any information on to Region III.
No violations or deviations were identified.
9.
Events Occurring During the Inspection During the inspection period, the licensee reported two occurrences where
the fuel transfer tube valve was not fully shut (auxiliary building spent i
fuel pool to Unit 1 containment reactor cavity area) without the full benefit of security guards being present.
The events were immediately i
investigated, corrective measures were taken, and were written up as a
" logged 'ccurrence." The resident inspectors were informed of the events and relayed the information to Region III security specialists who will evaluate the events as part of an upcoming security audit.
No violations or deviations were identified.
10.
Technical Specifications The final draft of the Braidwood Technical Specifications was issued by NRR under cover letter, dated August 22, 1986.
The SRI (0perations)
reviewed the changes as compared to the Proof and Review version, e.g.i, (1) added RHR. pump curve on pages 3/5 5-5, and 5-6a; (2) corrected Specification c.1 on page 3/4 6-3; (3) dropped Specification 3.8.2.lc on
,
page 3/4 8-10; and (4) corrected addressee in Specification 6.9.1.9 on page 6-21.
The. inspector verified the changes and found them to be acceptable.
No violations or deviations were identified.
11.
Plant Tours and Independent Assessments
,
The inspectors conducted routine plant tours during the inspection period to make an independent assessment of equipment conditions, plant
conditions, construction activities, security, fire protection, general personnel safety, housekeeping, and adherence to applicable regulatory
'
requirements.
During the tours, the inspector reviewed various logs, daily orders, interviewed personnel, attended shift briefings and plan of
the day meetings, witnessed various construction work activities, and independently determined equipment status.
During the shift changes, the 4,
inspector observed operator and shift engineer turnovers and panel walkdowns.
,
i i
'
-.
-
- -.
- -.. -..
.. _, -.- -.- -
. -..
- -
.., -.. -..
.. ~.... -.
,
,
No violations or deviations were identified.
12.
System Release to Operations - Release to Operations Control (RTO-RT0C)
During the inspection period, the inspectors reviewed portions of several system packages that were Released to Operations Control (ROTC).
Although the systems were not at the full Release to Operations (RTO)
status, some general observations were that the information was complex and difficult to follow; some deficiencies apparently lacked formal closure; and some poor practices, such as pencil entries, use of white-out, and entries lined out without initials and dates.
These comments were passed on to the licensee.
No violations or deviations were identified.
13.
Meetings, Training and Other Activities Regional Administrator Onsite On August 7, 1986, the Region III Administrator, Braidwood Project Director and Section Chief were onsite for an informational tour of the plant with the resident inspectors and to meet with the licensee for a plant status meeting.
The Regional Administrator expressed some concern with the licensee's ability to complete the demanding work load prior to licensee's scheduled fuel load date.
However, he complimented the licensee's pride in the plant, ability to meet a schedule, and their operator licensing exam results.
He also complimented the cleanliness of the plant, e.g., the Unit 1 containment, turbine deck, and model spaces were impressive; however, the auxiliary building was in fair condition due to the construction activity.
He completed the meeting with discussions on forthcoming activities related to licensing and operation.
News Media Contact On August 29, 1986, a reporter for the Reddick-Essex Courier contacted the Senior Resident Inspector (Operations) for information for a series of upcoming articles.
The discussion covered the SRI's experience, the NRC enforcement policy and some present enforcement activities in progress.
14.
Exit Interview The inspector met with licensee and contractor representatives denoted in Paragraph 1 during and at the conclusion of the inspection on October 8, 1986.
The inspectors 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.
.
-
_.
-_
-.
--
.
....
.
.