ML20214A141
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| ML20214A141 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 04/29/1987 |
| From: | Butcher R, Dance H, Will Smith NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20214A112 | List: |
| References | |
| 50-416-87-10, NUDOCS 8705190289 | |
| Download: ML20214A141 (13) | |
See also: IR 05000416/1987010
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UNITED STATES
' NUCLEAR REGULATORY COMMISSION
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[[pn 8ttgo n REGION 11
g j 101 MARIETTA STREET, N.W.
* t ATLANTA, GEORGI A 30323
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Report No.: '50-416/87-10
Licensee: System Energy Resources, Inc.
Jackson, MS 39205
Docket No.: 50-416 License No.: NPF-29
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Facility Name: Grand Gulf Nuclear Station
Inspection Conducted: March 14 through April 17, 1987
Inspect s: /[' 2f V7
w R.C Butt:her, Senior Resident Inspector Date S~1gned
P. Adn ,
W.F. Sm th Resident Inspector
Adn
Date Signe
Approve'd by: A* 27 N
H.C. Dance, Section Chief, Division Date Signed
of Reactor Projects
SUMMARY'
Scope: This routine inspection was conducted by 'the resident inspectors at
the site in the areas. of Licensee Action on Previous Enforcement Matters,
Operational Safety Verification, Maintenance Observation, Surveillance Observa-
tion, Reportable Occurrences, Inspector Followup and Unresolved Items, and
Design Changes and Modifications.
Results: Three violations were identified: Failure to document an identified ~
deficiency, failure to implement the SSW basin acid addition system per the
temporary alteration and failure to install control rod hydraulic control units
per design drawings.
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REPORT DETAILS
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1. Licensee Employees Contacted
J.E. Cross,GGNS Site Director
*C.R. Hutchinson, GGNS General Manager
R.F. Rogers, Manager, Unit 1 Projects
A.S. McCurdy,' Manager, olant Operations
*J.D. Bailey, Compliance Coordinator
M.J. Wright, Manager, Plant Support
L.F. Daughtery, Compliance Superintendent
D.G. Cupstid, Start-up Supervisor
R.H. McAnuity, Electrical Superintendent
*J.P. Dimmette, Manager, Plant Maintenance
W.P. Harris, Compliance Coordinator
*J.L Robertson, Licensing Superintendent
L.G. Temple, I & C Superintendent
*J.H. Mueller, Mechanical Superir,tendent
L.B. Moulder, Operations Superintendent
*S.F. Tanner, Manager, Nuclear Site QA
J.V. Parrish, Chemistry / Radiation Control Superintendent
*J.W. Yelverton, Acting Manager, Plant Operations
*F.W. Titus, Director, Nuclear Plant Engineering
*S.M. Feith,- Director, QA
Other licensee employees contacted included technicians, cperators,
security force members, and office personnel.
* Attended exit interview
2. Exit Interview (30703)
The inspection scope and findings were summarized on April 17, 1987, with
those persons indicated in paragraph-1 above. The licensee did not
identify as proprietary any of the materials provided to or reviewed by
the inspectors during this inspection. The licensee had no comment on che
following inspection findings:
416/87-10-01, Inspector Followup Item, Diesel Generator fuel oil
valve pit cover design change. (paragraph 4)
416/87-10-02, Inspector Followup Item, Determine source of water on
SSW basin floor. (paragraph 4)
416/87-10-03, Inspector Followup Item, Failure of mechanics to
properly assemble a relief valve. (paragraph 5)
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416/87-10-04, Violation, Failure to document an identified
deficiency. (paragraph 5)
416/87-10-05, Inspector Followup Item, Standby Liquid Control
surveillance procedure improvement. (paragraph 6)
416/87-10-06, Inspector Followup Item, Standby Diesel Generator (SDG)
12 start pushbutton troubleshooting. .(paragraph 6)
416/87-10-07, Inspector Followup-Item, Method of venting air pressure
from SDG air start lines. (paragraph 6)
416/87-10-08, Inspector Followup Item, Required SDG fuel oil tank
levels. (paragraph 6)
416/87-10-09, Violation, Failure to implement the SSW basin acid
addition system per the temporary alteration. (paragraph 9)
416/87-10-10, Violation, Failure to install hydraulic control units
per design drawings. (paragraph 8)
3. Licensee Action on Previous Enforcement Matters (92702)
(Closed) Violation 416/84-54-01. The licensee has issued Nuclear Produc-
tion Department Procedure 1.14, Verification and Certification of
Submittals and Information, to establish a method to be used for review
of any information that is to be submitted to the NRC. Attachment II to
Procedure 1.14 provides guidelines for verification and certification of
submittals and information. This action should ensure future submittals
are accurate.
4. Operational Safety Verification (71707)
The inspectors kept themselves informed on a daily basis of the overall
plant status and any significant safety matters related to plant opera -
tions. Daily discussions were held with plant management and various
members of the plant operating staff.
The inspectors made frequent visits to the control room such that it was
visited at least daily when an inspector was on site. Observations
included instrument readings, setpoints and recordings, status of
operating systems, tags and clearances on equipment controls and switches,
annunciator alarms, adherence to limiting conditions for operation,
temporary alterations in effect, daily journals and data sheet entries,
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control room manning, and access controls. This inspection activity
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included numerous informal discussions with operators and their seper-
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Weekly, when the inspectors were onsite, selected Engineered Safety
Feature (ESF) systems were confirmed . operable. The confirmation is made
:by verifying the following: Accessible valve flow path alignment, power
supply breaker and fuse status, major component leakage, lubrication,
cooling and general condition, and instrumentation.
General plant tours were conducted on at least a biweekly basis. Portions
of the control building, turbine building, auxiliar'y building and outside
areas were visited. Observations included safety related tagout verifica-
tions, shift turnover, sampling program, housekeeping and general plant
conditions, fire protection equipment, control of activities in progress,
radiation protection controls, physical security, problem identification
systems, and containment isolation. At least monthly, the licensee's
onsite emergency response facilities were toured to determine facility
readiness.
The folicwing comments were noted:
On April 1, 1987, the. inspectors noted that the valve pits for the
Divisions 1, 2 and 3 diesel generator fuel oil tanks were nearly full of
water. Licensee personnel were in the process of pumping the water out.
This condition was previously identified in NRC Inspection Report 86-20
when an ESF walkdown was conducted. At that time the Division 3 valve pit
was found full of water due to the design of the cover. The licensee
enlarged the covers to overlap the opening of the pit, however the new
design has since proved itself to be inadequate as evidenced by the
presence of water. The licensee is presently working on a solution. This
shall be Inspector Followup Item 416/87-10-01.
While touring the Standby Service Water (SSW) basins the inspectors noted
the following discrepancies:
Basin A SSW pump motor (P41C001A) had a junction box on the motor
housing with the cover off and hanging by a restraining chain.
Several loose wires were hanging out of the junction box and a loose
screw was laying inside the junction box. On the wall behind the
High Pressure Core Spray (HPCS) service water pump motor (P41C002C),
a junction box cover was loose and hanging by only one screw. The
floor in both areas was completely covered in water.
Basin B SSW pump motor (P41C0018) had a similar junction box on the
motor housing as P41C001A. It also had the cover off with several
loose wires hanging down. The floor in the B basin arca was dry.
The licensee was notified of the above discrepancies and took corrective
action. The inspectors questioned the source of tha excessive water on
the SSW basin floor since there was no obvious source. The source of
water inside the SSW pump rooms will be Inspector Followup Iten 416/87-
10-02.
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-The inspectors ' noticed, on repeated occasions, water on the floor in the
RCIC/RWCU mezzanine on auxiliary building level _119. The water appeared
to be leaking from the steam tunnel floor piping penetrations above. The-
. operators explained that due to leakage in the steam tunnel, water must be
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drained from the steam tunnel floor every shift or it overflows through
the penetrati.ons to the area below. One of the areas where water is
collecting _ on the floor is a high contamination area and leakage seems to
persist ' independent of- draining the steam tunnel floor. The inspectors
expressed concern that it is a . poor radiological work practice to allow
this leakage to exist uncollected and uncontrolled. The licensee has
corrected this condition by collecting the water and directing it to a
drain.
No violations or deviations were identified.
5. Maintenance Observation (62703) .
During the report period, the inspectors observed portions of the
maintenance activities listed below. The observations included a review
of the work documents for adequacy, adherence to procedure, proper
tagouts, adherence to technical specifications, radiological controls,
- observation of all or part of the actual work and/or retesting in
progress, specified retest requirements, and adherence to the appropriate
quality controls.
E 1653 M71253, Repair, resetting and retesting of Standby Liquid Control
System Pump A discharge relief valve C41-F029A.
MWO EL3567, General Maintenance Instruction 07-S-12-71, Revision 3,
calibration checks of Time Delay Relays.
MWO E71482, Replace Time Delay Relay Block.
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- MWO E71202, Install chart recorder and monitor points on Division 1
and 2 diesel generators related to previous Division 1 diesel
, generator output breaker trip. . On March 12, 1987, the inspectors witnessed the restoration to service and
retesting of relief valve C41-F029A. When the operator opened the Standby
Liquid Control pump suction valve, water leaked profusely out of the
relief valve gagging screw plug. The operator immediately shut the
suction valve to stop the leakage. The inspectors noted that this was not
an ordinary mechanical joint leak, but rather, the gagging screw plug was
not properly installed when the valve was reassembled in the shop. The
leak was corrected by maintenance tightening the plug, and then the valve
passed the retest. When the NRC inspector questioned a QA inspector who
was present for the retest if a deficiency report or other document was
initiated to document the deficiency, the response was negative, nor did
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he indicate any intention to do so, because he stated he did not consider
the official retest as having started. The operator failed to d]cument
the deficiency also. This is a failure to comply with GGNS Administrative
Procedure (AP) 01-S-03-2, Quality Deficiency Reports (QDRs). Section-
6.1.1 requires all quality deficier.cies, unless identified on another
deficiency document, to be documented in accordance with this procedure.
The inspectors reviewed MWO 71253 to determine if adecuate instructions
existed to ensure the valve was properly reassembled. There were no
detailed instructions nor was the vendor manual-referenced, thus the work
instruction appeared to be inadequate. The inspector expressed concern to
the licensee over the way this job was handled, with particular emphasis
on a QA representative's reluctance to insure that the apparent deficiency
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was documented. Subsequently the licensee issued two QDRs (144-87 and
145-87) which identified a lack of knowledge on the part of the mechanic
reassembling the valve and failure of MWO 71253 to provide adequate
instructions.
Upon investigating, the licensee deternined and r ported to the inspectors
that the applicable vendor manual was used for the work on the relief
valve, although not referenced in the MWO. The licensee stated that
failure of the mechanics to properly assemble the relief valve was
indicative of a training deficiency, and that appropriate corrective
actions will be documented as 00Rs 144-87 and 145-87 are processed.
Followup inspection of actions taken shall be tracked under Inspector
Followup Item 416/87-10-03.
Section 16.0 of the licensee's NRC-approved Operational Quality Assurance
Manual, MPL-TOPICAL-1, Revision 5, requires, in part, that measures shall
be established to ensure that conditions adverse to quality, such as
deficiencies, are promptly identified and corrected. AP 01-S-03-2
established this measure. Failure to pronptly document the deficient
reassembly of C41-F029A as evidenced-by the gagging screw plug being left
loose is a violation of this requirement (416/87-10-04). QDR 168-87 was
initiated by the licensee to identify the failure to document a defi-
ciency, and corrective measures are being taken to ensure that such
deficiencies are recognized and documented in the future.
6. Surveillance Observation (61726)
The inspectors observed the performance of portions of the surveillances
listed below. The observation included a review of the procedure for
technical adequacy, conformance to Technical Specifications (TSs),
verification of test instrument calibration, observation of all or part of
the actual surveillances, removal from service and return to service of
the system or components affected, and review of the data for accept-
ability based upon the acceptance criteria.
06-0P-1C41-M-0001, Revision 25, Standby Liquid Control System
Operability.
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06-0P-1P75-M-0002, Revision 32, Standby Diesel Generator (SDG) 12
Functional Test.
06-IC-SD17-R-1025, Revision 22, Ef fluent System Flow Rate Monitor
Calibration (Turbine Building Exhaust Radiation Monitor).
09-S-06-12, Revision 1,102.5% Core Flow Flux Controller High Limit
Setting Verification.
06-0P-1P75-M-0001, Revision 32, Standby Diesel Generator (SDG) 11
Functional Test.
During conduct of SLCS surveillance procedure 06-0P-1C41-M-0001 listed
above, the inspectors noted that the precision test gauge connected at
IC43R003 (SLCS pump discharge pressure) was oscillating violently as
pressure was increased. At full discharge pressure the fluctuations from
this positive displacement pump were so wide the end of the gauge pointer
broke off. The operator experienced considerable difficulty in trying to
maintain discharge pressure by throttling C41-F016 and at the same time
throttling the pressure gauge isolation valve to dampen the oscillations.
It was evident that he had marginal control over the system, and the
probability of lifting the relief was high. The inspector expressed
concern to the licensee that failure to obtain more positive control over
this system during testing could lead to damage or unsatisfactory test
results. The licensee indicated that the procedure will be changed to
ensure better control, by the next time this quarterly surveillance is
due, and that hardware ' changes are under consideration. This shall be
tracked by Inspector Followup Item 416/87-10-05.
During conduct of SDG 12 functional test 06-0P-1P75-M-0002 above, the
diesel failed to start when the control room operator depressed the start
pushbutten. i.fter reverifying prerequisites and finding no problems, a
second attempt was made and the SDG started and came up to speed as
required by the Technical Specifications. The Shift Superintendent
informed the inspectors that a Maintenance Work Order (MWO) was initiated
te troubleshoot the pushbutton. The inspectors will follow up to verify
that the failure is identified and corrected as appropriate. This shall
be Inspector Followup Item 416/87-10-06.
Also during the SDG 12 surveillance above, the procedure required that the
drain plugs be removed and reinstalled in the isolated air start strainers
to depressurize the air start lines (steps 5.1.7 and 5.2.7). Since this
section of piping gets pressurized over 200 psig, the mechanic satisfied
this requirement by first loosening an instrument fitting to depressurize
the piping so that the plug could be safely removed. This appeared to be
stretching the intent of the procedure because the procedure does not
specifically authorize the breaking of an additional air pressure
boundary. The inspectors discussed this with the licensee who indicated
that the procedure would be reviewed and revised as appropriate. This
shall be Inspector Followup Item 416/87-10-07.
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Following' the SDG surveillance the inspectors reviewed surveillance
procedure 06-0P-IP75-M-0002, Revision 32. The procedure requires the
operator verify that the fuel oil day tank contains a minimum of 220
gallons of fuel as indicated by at least 12 inches on fuel oil day tank
level indicator LI-R607B on panel 1H13-P864. The inspector reviewed
the control room round sheet and it required a minimum of 22 inches on
LI-R607P. The inspector then reviewed several documents for fuel oil day
tank and storage tank minimum level requirements and found the following
numbers specified:
Division 1 and 2 SDGs Day Tank Storage Tank
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Yellow marking on control room 22 inches 6.2 feet
level indicator
Control room round sheet 22 inches 8.7 feet .
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Alarm Response Instruction 23 inches 8 feet 8
04-1-02-1H22-P400-1A-C1 inches
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DG Functional Test (06-0P-1P75- 12 inches 8 feet 8
M-0001/0002) inches
Division 3 SDG (HPCS)
Yellow marking on control room 10 inches 5.4 feet
level indicator
Control room round sheet 10 inches 6.3 feet
Alarm Response Instruction 9.5 inches 76 inches
04-1-02-1H13-P870-5A-El
DG Functional Test (06-0P-1P81- 12 inches 5 feet 11
M-0002) inches
The Division 1 and 2 SOG day tank fuel oil level meters on the local
panels read in 1/8th tank increments. When observed on April 3,1987,
during a surveillance test, the Division 1 meter read 3/4 full when the
control room indicator read 28 inches (out of 60 inch span). Discussions
with the licensee indicate that the day tank level (for all three
divisions) mest be a minimum of 12 inches to satisfy TS requirements.
This would indicate that the HPCS SDG day tank level could be less than TS
requirements on several indicators noted above and would be considered
acceptable. Since there is an automatic fuel oil transfer pump for
maintaining the day tank level well above TS limits, there is little
possibility of being below TS limits during actual operation. The
licensee was requested to evaluate the numerous values noted, determine
the appropriate value and revise the SDG fuel oil level requirements to be
consistent with TS limits. This will be Inspector Follow-up
Item 416/87-10-08.
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No violations or deviations were identified.
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7. Reportable Occurrences (90712 & 92700)
The below listed event reports were reviewed.to determine if the informa-
tion ' provided met the _ NRC . reporting requirements. The determination
included adequacy of event description ' and corrective action taken or
planned, existence of potential generic problems and the relative safety
significance of each event, Additional inplant reviews and discussions
with plant personnel . as appropriate were conducted .for the reports
indicated by an asterisk. The event reports were reviewed using the -
guidance of the general policy and procedure for NRC enforcement actions,
regarding licensee identified violations:
The following license Event Reports (LERs) are closed.
LER No. Event Date Event
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*85-033 August 30, 1985 Unit 2 ultimate
heat sink
components
required for Unit
1 operability.
*86-027 August 15, 1986 Fire watches
exceed TS required
frequency.
The event of LER 86-027 was addressed in Inspection Report 416/86-24.
LER 85-033 reported that portions of the Unit 2 SSW system required to
support the seismic qualification of_the Unit 1 SSW system were not under
the operational -control of the plant. Other Unit 1/ Unit 2 interface
concerns were subsequently identified. In lieu of updating this LER the
licensee submitted a Special . Report, Unit 1/ Unit 2 Interface (AECM-87/
0077) to provide a comprehensive review of actions taken and planned.
This LER is closed and the Unit 1/ Unit 2 interface concern will be tracked
under violation 416/86-17-03 2nd the special report update.
Due to the number of problems associated with the Standby Service Water
(SSW) system and their potential safety impact, an enforcement conference
was held in the NRC Region II office on February 4,1987, as documented in
Inspection Report 416/87-03. Report 416/87-03 and LER 86-029 stated the
licensee would initiate an independent design review of the SSW system.
On March 18, 1987, the licensee reported that during the SSW design review
it was determined that a postulated single failure could cause the loss of
the SSW system / Ultimate Heat Sink (VHS) 30 day water inventory. A loss of
power from Motor Control Center (MCC) 15B61 would leave valves P41F125 and
P41F066A open which could allow the flow of SSW from the A SSW train to
the Plant Service Water (PSW) system through 3 inch cross connect piping.
This would result in the loss of cooling water to the A train control room
air conditioning condenser but the B train would still be available. The
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loss of SSW water through the cross connect piping would eventually result
in the excessive loss of UHS inventory. As interim corrective action the
licensee issued night orders to tag close PSW to SSW cross connect valves
P41F125 and P41F189 with SSW supplying the operable control room air
conditioner (PSW normally supplies control room air conditioner cooling
water).
Tne licensee conducted a safety evaluation to assess the ability of the
SSW/ UHS to meet established design criteria with the design deficiency
noted above but with the implementation of appropriate administrative
controls for operator action. The safety evaluation determined that with
appropriate operator action within 7 hours, no unreviewed safety question
exists. Off-Normal Event Procedure (ONEP) 05-1-02-I-4, Loss of Offsite
Power and ONEP 05-1-02-V-11, Loss of Plant Service Water, were revised to
incorporate the recommended administrative controls. Based on the above
actions, the licensee secured the SSW system and returned the PSW/SSW
isolation valves to their normal lineup.
(Closed) P2184-07, Bonney Forge Non-Conforming Material . Based on a
request from Nuclear Plant Engineering (NPE), Bechtel Power Corporation
conducted a review of materials supplied to the licensee which originated
from Bonney Forge. It was determined that no materials having the
heat numbers identified as potential problem material were used in the
construction of any Unit I nuclear piping systems. Based on the study
results, the licensee determined no action was required.
(Closed) P2185-01, Containment Purge Valve. Based on a Henry Pratt
Information Letter, Nuclear Plant Engineering recommended to Plant Staff
that Locktite be used on all four sides of the shaft key during reassembly
on any Pratt valve and Limitorque operator. The valve driven gear spline
adapter and key are secured to the valve shaft by a force fit together
with the use of Locktite. Based on NPE recommendations the plant staff
revised Maintenance Section Procedure 07-S-08-510, Installation, Reinstal-
lation, Disassembly and Reassembly of valves, to require the use of
Locktite grade 242 or 271 as recommended. No further action was required.
Subsequently, 07-S-08-510 was superseded by Plant Modification and
Construction Section Procedure 15-S-02-502 which also incorporated the
above recommendativn.
(Closed) P2186-02. The Electro-Motive Division of General Motors reported
that upper connecting rod bearings had been found with mislocated dowel
holes. Morrison-Knudsen Company, manufacturers of the High Pressure Core
Spray (HPCS) diesel generator system notified General Electric (GE) which
supplied the HPCS diesel to SERI. GE informed SERI that only connecting
rod bearings purchased between November 1985 and April 1986 were suspect
and require inspection. GE records indicate they neither purchased nor
supplied any connecting rod bearings to MP&L (now SERI) for the Grand Gulf
Plant during this time frame and no further action is required.
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(Closed) P2186-04, BBC Brown Boveri, Inc. K600/K800 Circuit Breakers Wire
Harness. On June 30, 1986, the NRC was notified of a 10 CFR Part 21
report identifying a condition in the type K600S circuit breakers stpplied
by BBC Brown Boveri, wherein a control wire harness came into direct
contact with a racking gear inside the breaker. The gear teeth had worn
through the protective cover and severed a wire causing a grounded
condition. The licensee was also notified in June 1986. During the first
refueling outage (October 1986 thru January 1987) the plant staff
inspected 26 out of 26 E5F breakers scheduled and 40 out of 60 balance of
plant breakers scheduled for inspection and found no damage to the wiring
harness. The remaining breakers will be inspected during the normal
inspection program. General Maintenance Instruction 07-S-12-50, Inspec-
tion and Calibration of.480V ITE K600S-K1600S Breakers, has been updated
to inspect breakers for this problem.
No violations or deviations were identified.
8. Inspector Followup and Unresolved Items (92701)
(Closed) Inspector Followup Item 416/85-05-01. This item was addressed in
Inspection Report 416/87-06 and was inadvertently listed as an open item
when it should have been closed. This is to document the closure of
416/85-05-01.
(Closed) Inspector Followup Item 416/86-39-03. Correction of records on
G.N. Bettis valve actuator seal replacements. The licensee issued a
Quality Deficiency Report (QOR) to formally identify the problem and
implement appropriate corrective actions. The inspectors reviewed the
actions taken which included correction of the records by issuance of a
record supplement and issuance of a memorandum to mechanical personnel
frcm the Mechanical Maintenance Superintendent emphasizing their
responsibilities as they relate to work documentation.
The records, however, were not adequately corrected. The inspectors noted
that in an effort to correct the record for valve E61-F009 (Task Card No.
ME0027), the licensee added another copy of the erroneous material ticket
instead of deleting or annotating that it was not applicable to E61-F009.
The inspectors have since verified that this has been corrected. QA issued
another QDR (87-1073) to document the incorrect records correction and
actions taken to ensure that record supplements are more carefully
screened in the future.
(Closed) Unresolved Item 416/87-01-04. Resolution of missing fasteners
on control rod drive Hydraulic Control Units (HCU's). This item was
previously addressed in NRC Inspection Reports 416/87-01 and 416/87-05.
On February 6, 1987, during a tour of the containment the inspectors
found that one fastener required for securing HCU 36-13 to the foundation
was missing. It was apparent to the inspectors that the fastener had
never been installed, and thus a concern was raised as to the as-built
condition, and thus the seismic qualification of the other 192 HCUs. The
licensee was requested to provide documentation that supports this and any
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other deficient HCU installations. There was no ' documentation on file
which would allow the fastener to be missing from HCU 36-13. The fastener
was replaced, a 100% inspection was conducted on the other HCUs and other
deficient fastener installations were corrected, confirmed acceptable by
review of previous documentation, or accepted in their as-found condition
by additional engineering analysis. From the results of these actions, it
appeared that about 36 HCUs were operated in an unanalyzed, deficient
condition with regard to seismic qualification from plant licensing until
February 1987. A major effort was implemented by the licensee to review
HCU installation records, to accurately document the as-found condition of
the 193 HCUs and to perform proper engineering analyses of the condition
as-found and after repairs (as-left). As was reported in NRC Inspection
Report 416/87-05, the inspectors expressed concern that the licensee
had not probed this problem to the depth necessary when the concern was
initially raised. As a result, the licensee also implemented a review of
program requirements for handling such problems.
On April 10, 1987 the licer.see presented a documentation package
containing the investigation results and disposition of deficiencies found
on HCU mounting fasteners. The inspectors met with the 1.censee to
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discuss the package and then reviewed it for accuracy and completeness.
Based on the information provided in the package and independent observa-
tions made on the equipment by the inspectors, it appears that prior to
discovery of the missing fastener on HCU 36-13 on February 6, 1987, 150 of
the 193 HCUs had an average of about three fasteners (of eleven) with less
than the recommended torque shown on drawing 767E800, Hydraulic Control
Unit, five HCUs had a mounting bolt missing without having in its place
the equivalent weld required by Bechtel Supplier Deviation Disposition
Request (SDDR) M-316.0-016 and General Electric Field Deviaticn Disposi-
. tion Request JB1-471. There were also a few isolated instances of missing
washers, broken bolts, missing nuts, and welds which were the incorrect
size. It was erroneously presumed by the licensee, however, that as of
May 30, 1982, all HCOs were mounted properly by the subcontractor, Reactor
Controls, Incorporated (RCI), in accordance with the applicable drawings
and specifications, and that any missing bolts had an equivalent weld, as
approved by the above SDDR. Therefore the system was operated in an
unanalyzed condition until after February 6, 1987, when the deficient,
as-found condition was corrected as much as practicable and as- corrected
analysis was performed by Nuclear Plant Engineering. 10 CFR Part 50,
Appendix B, Criterion V states, in part, that activities affecting quality
shall be accomplished in accordance with documented instructions and
drawings of a type appropriate to circumstances. Contrary to this, 156 of
193 HCUs were not mounted in accordance with the applicable drawings and
specifications during plant operation. This is a violation. (416/87-
10-10).
The licensee performed a safety analysis of the HCUs in the as-found
condition. The analysis assumed that no fasteners were torqued to provide
the preload recommended, the various configurations of missing fasteners
were considered, welding provided under the above SDDR was conservatively
neglected in the calculation, and fasteners that were not fully seated
were assumed to be missing. The results of that analysis concluded that
the as-founJ conditions did not result in safety concerns which would
have jeopardized the integrity of the reactor coolant pressure boundary,
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12
reduced capability to achieve and maintain reactor shutdown, or reduced
capability to prevent or mitigate the consequences of an accident. The
licensee has indicated that corrective actions for this violation will
include a review of other safety related work performed by RCI and will
provide documentation as to results and additional actions if appropriate.
9. Design, Design Changes and Modifications (37700)
The inspector reviewed the licensee's program for temporary alterations.
Administrative Procedure (AP) 01-S-06-3, Control of Temporary Alterations,
Revision 20, was reviewed to verify controls were in accordance with
TSs,10 CFR 50.59 and the approved QA program, MPL-TOPICAL-1. Adequate
controls are specified in tne procedure to ensure independent verifica-
tion, modification status, proper restoration, and records retention and
periodic verification of outstanding alterations. The Temporary Altera-
tion log was reviewed and several Temporary Alterations were verified.
The inspector had the following comments:
Temporary Alteration 87-0005. The Temporary Alteration Request Form
(Attachment I to 01-S-06-3) did not have the tag numbers entered
and the installed tags did not have individual numbers. The Shift
Supervisor initiated a QDR to correct the tag numbers.
Temporary Alteration 86-0034. The installation of an acid storage
tank and related piping to add chemicals to the Standby Servi e Water
(SSW) basins was accomplished by this alteration. A Temparary
Alteration Request Addendum, Attachment III to 01-S-06-3, shows
the installation details for this temporary alteration. A visual
inspection of the actual installation showed that the installed
piping and valves did not agree with the temporary alteration. At
each basin the installed piping divided into two lines with an
installed valve in each line. Additionally the inspector observed
the following discrepancies.
Valve FTA6, fill line to the acid tank, was broken off and laying in
the tank pit area.
Valve FTA4B had no handle and no identification (temporary altera-
tion) tag attached.
The 4 inch drain line from the acid tank berm to the SSW B basin was
disconnected between the berm and the first isolation valve (FTA7)
which provided a leak path directly to the ground.
TS 6.8.1 requires written procedures to be established, implemented and
maintained covering the procedures recommended in Appendix A of Regulatory
Guide 1.33, Revision 2, February 1978. Appendix A of Regulatory Guide
1.33 recommends procedures covering the bypass of safety functions and
jumper control. Administrative Procedure 01-S-06-3, paragraph 6.1,
requires temporary alterations to be documented and controlled using
temporary alteration forms (Attachments I and III). The failure to
implement the installation of the SSW basin acid storage tank and related
piping in accordance with the approved temporary alteration is a
violation. This is violation 416/87-10-09.
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}}