IR 05000528/1984054
| ML17298B747 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 12/07/1984 |
| From: | Hollenbach D, Miller L, Narbut P, Sorensen R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML17298B746 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-1.A.2.1, TASK-1.B.1.2, TASK-1.C.1, TASK-1.C.5, TASK-2.B.4, TASK-2.D.3, TASK-2.F.2, TASK-TM 50-528-84-54, 50-529-84-37, 50-530-84-26, NUDOCS 8412310281 | |
| Download: ML17298B747 (30) | |
Text
U. S.
NUCIEAR REGULATORY COMMISSION
REGION V
Report Nos.
50-528/84-54, 50-529/84~
and 50-530/84-26 Docket Nos.
50-528, 50-529 and 50-530 License Nos.
CPPR-141, 142 and 143 Licensee:
Arizona Public Service Company P.
O. Box 21666 Phoenix, Arizona 85036 Facility Name:
Palo Verde Nuclear Generating Station Units 1, 2 and
Inspection at:
Palo Verde Construction Site, Mintersburg, Arizona Inspection conducted:
November 12-23, 1984 D. Hollenbach, Reactor Specialist
~l'PQ R.
C. Sorensen, Reactor Inspector P.
Na but, Reactor Inspector ill~/64 Date Signed iz 5/S<
Date Signed (z sI'SV Date Signed Approved By:
L.
ler, Jr.,(C+e Reactor ProjectkAection,2 t7-
/
Date Signed Summary:
Ins ection on November 12-23 1984 (Re ort Nos. 50-528/84-54 50-529/84-and 50-530/84-26 Areas Ins ected:
Routine unannounced inspection by regional based inspectors of licensee followup of construction open items, Bulletins, and violations in
'nits 1, 2 and 3.
In addition, operations activities and procedures involving Unit 1 were examined.
The examined activities involved implementation of Three Mile Island Lessons Learned actions.
The inspection involved 137 inspector-hours onsite by three NRC inspectors.
Results:
Of the areas inspected, no violations or deviations were identified.
84123i028i 84i2i0 PDR ADOCK 05000528
DETAILS 1.
Persons Contacted a.
Arizona Public Service Com an (APS)
"J-H.
"C.
"-R.
-T.
"M AJ
"R
'R W.
W.
J.
K.
F.
D.
K.
C.
Van Brunt, Jr., Vice President, Nuclear Production Ide, Corporate QA/QC Director Bynum, Plant Manager Zerinque, Technical Support Manager Russo, Quality Audits and Monitoring Manager Kimmel, Transition Bloom, Licensing Engineer Van Brunt, Jr., Vice President, Nuclear Production Karbassian, Operations Engineer Ide, Corporate QA/QC Director Bynum, Plant Manager Burgess, Field Engineering Supervisor Jacobss, QA Engineer Craig, Startup Supervisor Fernow, Plant Services Manager Minnicks, ISC Superintendent Cutler, IRC Procedure Supervisor Hicks, Training Manager Mooten, ISEG Engineer Daley, QA Engineer Churchman, NSSS Test Group Supervisor b.
Bechtel Power Cor oration (BPC)
K. Scheter, Civil Structural Group Supervisor
"-R. Vote, QA Manager
"-R. Randels, Engineer-Denotes those persons attending exit meeting on November 21, 1984.
The inspectors also talked with other licensee and contractor personnel during the course of the inspection.
2.
Licensee Action on 10 CFR 50.55(e)
Construction Deficiencies (DERs)
The following 50.55(e)
items were reviewed by the inspector for reportability and to determine the thoroughness of'he licensee's corrective action.
The items marked with an asterisk (=') were judged by the licensee to be reportable under the
CFR 50.55(e) criteria; the others were considered not reportable.
The inspector reviewed nine final reports issued under
CFR 50.55(e)
that were determined by the licensee to be not reportable.
The inspector concluded that the licensee's determinations for reportability versus non-reportability were performed accuratel \\
A t >
U v
I A
(0 en)
"-DER 83-42 Dama ed Thermowells Found in the Reactor Coolant S stem RCS While performing pre-core hot functional testing (HFT) of the Unit
RCS, thermowells in the RCS hot and cold legs were found emitting steam from their temperature elements.
The thermowell failures were due to flow-induced vibration fatigue caused by vortex shedding.
The vibration caused the thermowells to shear, thus violating the pressure boundary.
Combustion Engineering (CE) redesigned all the thermowells and nozzles in all three units.
The thermowells were tested in Unit 1 under full flow conditions.
These test results were verified by CE as satisfactory.
The inspector examined the DCPs (Design Change Packages)
provided for replacing all the thermowells with the redesigned thermowells.
The DCPs for Units 1 and 2 are complete and signed off by QC.
The nozzles have been installed in Unit 3.'owever, because'f the fragile nature of the thermowells they will not be installed in Unit 3 until hot functional testing.
The technical acceptance of the redesigned thermowells is being examined by NRR.
This item will remain open pending technical review and acceptance by NRR.
(Closed)
DER 83-58 Air Flow Totalizers Can Not Be Calibrated The flow totalizers, made by Measurements Incorporated, are used in the radiation monitoring system supplied by Kaman Sciences Incorporated.
All six flow totalizers used in Unit 1 were unable to be calibrated within the required eight percent.
These flow totalizers have been replaced in Unit 1 with Hastings totalizers.
These have been calibrated and accepted by QC.
Design Change Package (DCP)
2CN-SQ-014 has been issued to change the flow meters in Unit 2.
Kaman will incorporate the updated design in the Unit 3 equipment.
The inspector examined all pertinent NCRs and DCPs.
The NCR for Unit 1 was completed and accepted by QC.
The DCP for Unit 2 has been issued and is being tracked.
This item is considered closed.
(Closed)
DER 83-78 No Documentation of User Tests and No Controllin Procedure for Rock Bolt E ansion Anchors During a QA investigation in response to a
NRC Region IV Violation for PVNGS User Test Program, User Test Documentation for Rock Bolts was found to be missing, incomplete, incorrectly done, and not traceable to Material Receiving Reports (MRR) or any other receiving documentatio I'
CI
The licensee was able to match user tests to rock bolts documented by three MRR numbers using the dates the rock bolts were received and the dates of the user tests.
These user tests did not contain the following required information: yield strength, ultimate strength, and percent elongation.
Two of the three user tests were missing yield strength.
However, the ultimate strength and elongation were well within specifications.
The other user test recorded the elongation based on an incorrectly sized specimen.
It is not possible to determine if the elongation is within specifications based on these specimens.
However, the yield strength and ultimate strength were well within the specified limits.
There were six batches of rock bolts that had no user tests performed on them.
Three of the batches were in the warehouse.
User tests were done and found acceptable for these batches.
Two other batches had been mixed in with five other batches (a total of 230 bolts).
Thirty three bolts were chosen for testing from this batch.
One sample failed the elongation test and reached 97 percent of the required ultimate strength.
Two other samples failed to meet.
elongation specifications.
Based on a conversation with the Bechtel Civil Structural Group Supervisor, the bolts were determined by the inspector to be acceptable on the basis that the bolts had adequate ultimate strength for their application.
As a corrective action to prevent recurrence the licensee has incorporated a procedure change to WPP/gCI 4-0, "Receiving Inspection."
This specification now requires user tests be done on all rock bolts upon receipt.
The specification also requires rock bolts be r'eceived by gCEs ((}uality Control Engineers)
only.
The licensee has also scheduled an audit of all material requiring user tests in December 1984.
The inspector examined WPP/(}CI 4-0 to insure the procedure change was incorporated into the procedure and found it acceptable.
The inspector also examined the MRRs for all rock bolts received and found no additional problems.
The NCRs were reviewed by the inspector to insure all batches of rock bolts had user tests done on them.
All the documentation examined by the inspector was complete and acceptable.
This item is considered closed.
A (Closed)
DER 84-12 Cleanliness Violation of RCP lA and S/G No.
During a licensee surveillance of the Unit, 1 startup program,
"Zone III" cleanliness violations were discovered at the Reactor Coolant Pump (RCP) lA discharge pipe and the Steam Generator (S/G)
No.
1 manway.
While CE was working on the RCP, Bechtel Construction Personnel replaced the wooden dam in the pump discharge with a metal dam.
During this operation metal chips and cutting fluid were found inside the RCPs.
Subsequently, the RCP internals were returned to the appropriate cleanliness condition.
In an unrelated incident the S/G No.
1 manway was removed without adequate cleanliness
C ll
coordination.
Subsequently, the entire loop was inspected and returned to.the appropriate cleanliness condition.
As a corrective action to prevent recurrence, the licensee has issued a procedure change to WPP/gCI 13.0,
"Housekeeping",
which requires protective covers be placed over NSSS structures whenever left unattended for more than one hour.
The inspector examined the SFRs (Startup Field Reports)
used to return the component to the appropriate cleanliness class.
The SFR's were signed off as completed and accepted by QC.
This item is considered closed.
(Closed)
-DER 84-34'm ro erl Installed Critical Friction e
Hi h Stren th Connectors During a Bechtel inspection walkdown of critical, friction type, high strength connectors (as a result of the CAT inspection),
some connectors on the safety injection (SI) tank keyways were found to be incompletely tightened, improperly installed, and missing washers.
All connectors of this type on all three units were inspected and the deficiencies noted on NCRs.
The inspector reviewed the three DCPs and 6,NCRs documenting the deficiencies.
All incompletely tightened bolts and other nonconforming hardware in Unit 1 were a repaired and accepted by gC.
The repairs needed in Units 2 and 3 are being tracked on NCRs.
To preclude the recurrence of this condition drawing 13-C-ZC5-573 has been revised to clarify the connection hardware requirements.
This item is considered closed.
(Closed)
DER 84-43 HPSI Three Inch Gate Valve Failure to 0 en During Startup Test Procedure 91PE-'1SI08, a three inch 1500 lb Borg-Warner motor operated gate valve failed to open.
The valve operator ceased to operate (per design)
due to excessive torque.
NCR SM-4284 was written by the licensee to document this condition.
The failure of the valve to open was determined to be due to a lack of lubrication of the valve stem.
After cleaning and lubrication the valve opened and closed under test pressure and flow conditions.
To prevent recurrence, the licensee included a step in the Preventive Maintenance Procedure to periodically lubricate the valve stem.
The inspector reviewed the documentation related to this deficiency.
The NCR was completed and accepted by gC.
The inspector verified the preventive maintenance procedure was revised to require lubrication of.the valve stem.
Six similar valves were examined by the inspector to determine if they passed Startup Test 1SI08.
All reviewed valves passed.
This item is considered close t f
V'
i
(Closed)
DER 84-45 Potentiall Bad S ot Welds on the Kaman Skid Mounted Indication and Control SMIC Units During seismic testing at the suppliers facility, Ka'man Instruments identified a mechanical failure on the SMIC units, revealing insufficient welds on some of the retaining studs.
There are six of these units present in both Units 1 and 2.
The SCIM units have not been shipped yet to Unit 3.
Kaman has prepared a field test procedure for testing the stud welds on the SCIM units.
The inspector has verified the welds were tested and accepted in Unit 1.
Unit 2 is being tracked, on Investigation Report IR-2-IR-033.
The SCIM units for Unit 3 will be tested by Kaman before being shipped to the site.
This item is considered closed.
(0 en)
-DER 84-49 Auxilia Feedwater S stem Ex eriences H draulic Resonance During preoperational testing of the auxiliary feedwater pumps, the piping experienced hydraulic resonance when operating in the normal mini-flow configuration.
During additional testing it was observed that this unstable condition disappears when either the mini-flow is increased or the first stage block valve is closed.
CE specifies that auxiliary feedwater be delivered to the steam generators at a
minimum rate of 875 gpm.
This corresponds to a maximum bypass flow of 135 gpm under normal conditions.
CE has proposed increasing bypass flow to 260 gpm which eliminates the hydraulic resonance.
However, this decreased the flow to the steam generators to 750 gpm.
The inspector examined NCR's SM-4497 and SM-4500, which set the maximum bypass flow limit at 260 gpm and measured the tested vibration level as acceptable.
Both NCRs are complete and accepted by (}C.
The inspector also examined SAR Change Notice 1239, which when submitted to NRC will reflect (in the FSAR) the change in the flow to the steam generators from 875 to 750 gpm.
The licensee has informed NRR of the proposed SAR change but has not submitted it yet.
This item will remain open pending NRR acceptance of the FSAR change.
(0 en)
DER 84-61 Potential Failure of the Auxiliar Pressurizer S ra S stem APSS The NRC stated in a letter dated April 3, 1984, that the APSS may lose the capability to perform its intended function if the loop charging valve mechanically sticks open, causing insufficient flow through the APSS into the pressurizer.
The licensee had agreed to install a fail-closed pneumatically operated valve in series with the loop charging valve.
The installation of these valves is being
k
performed under Design Change Packages (DCPs)
1SM, 2SM, and 3CM-CH-197 for Units 1, 2 and 3 respectively.
The inspector examined the DCPs.
DCP 1SM-CH-197 was completed and accepted by QC.
The other two DCPs are written but are not complete.
The inspector examined the installed valve in Unit
against its installation drawing and found the installation met drawing requirements'he licensee has informed NRR of the valve installation in a letter preceeding a formal FSAR changes.
This item remains open pending NRR acceptance.
(Closed)
DER 84-70 Diesel Generator En ine Tri ed on Overs eed
The Unit 1 Diesel Generator Engine "A" tripped off during testing due to engine overspeed.
An investigation by the licensee showed the trip was due to galling, contamination, and no lubrication between the local manual control handle and the control shaft for the fuel injection pumps.
This deficiency was documented on NCR SM-4841.
The licensee cleaned and lubricated the shaft.
The fuel control mechanisms on the remaining diesels were inspected for binding and found acceptable.
To prevent recurrence the licensee added Preventive Maintenance Tasks (PMT) 008603 and 008604 to insure the mechanism is serviced quarterly.
The inspector examined the licensee's system for tracking preventive maintenance requirements.
The licensee uses a computer tracking system which contains the maintenance requirements and due dates for each piece of equipment.
The inspector verified tasks 008603 and 008604 have been added to the system and require the shaft be cleaned and lubricated.
This item is considered closed.
(0 en)
DER 84-85 Control Element Assembl E 'ection Accident
~Anal sis The calculated offsite doses, listed in 'ZSAR Table 15.4-1, through the power access purge valve in the event of a control element assembly (CEA) ejection accident did not include the effect of a loss of offsite power on the purge valve closure.time.
During an evaluation of offsite doses, the licensee determined the assumption in FSAR Section 15 '
of the containment purge valves being open was overly conservative.
The NRC Accident Evaluation Branch confirmed the Standard Review plan does"not require the purge 'valve be assumed I
open during a CEA ejection accident.
The licensee has initiated FSAR Change Notice 1272 to revise FSAR Se'ction 15-4 to delete the consideration of an open containment purge valve from the accident, analysis.
This 'item will remain open pending NRR acceptance of this FSAR proposed chang k iI lf
3.
Im lementation of Three Mile Island Lessons I,earned The inspector reviewed the below listed items which represented a portion of a comprehensive and integrated plan to improve safety following the events at Three Mile Island, Unit 2 in March 1979.
(The item numbers are from Enclosure 2 of NUREG-0737)
I.A.2.1 Immediate U
radin of RO and-SRO Trainin and ualifications Closed Open aspects remaining from Inspection Report 84-47 were closed as follows:
1)
An APS checklist, used to track licensed operator control manipulations performed as part of requalification training, failed to include a control manipulation required from Enclosure 4 of the Denton letter of March 28, 1980.
The control manipulation,
"Loss of Protective Channel," is now included in the revised version of the checklist.
This aspect is now closed.
2)
The licensee representative committed to including Mitigating Core Damage as part of operator requalification training beginning in April 1985.
Mitigating Core Damage had not been included as part of requalification training as required in the Denton letter of March 28, 1980.
This aspect is now closed based on this commitment and will be followed up as part of routine inspection.
(Followup Item 50-528/84-54-01)
I.B.1.2 Inde endent Safet En ineerin Grou (ISEG) (Closed)
The inspector verified the ISEG group to be in place and functioning within the framework of NUREG-0737.
They are located onsite but report offsite to a high level corporate official who is independent of the management chain for operations.
Procedures are in place for the group to perform surveillances of plant operations and special investigations.
All the members have adequate educational and experience backgrounds.
The group is independent of operations and is not responsible for any signoff function involving plant operations.
In summary, the ISEG group in place at Palo Verde satisfactorily meets the requirements and guidelines of NUREG-0737.
Therefore, this TMI Action Plan item is closed.
I.C.1.1 Emer enc Procedure for Small Break LOCA (Closed)
This procedure has been reviewed by the inspector for compliance with the Generic Owners Group Guidelines for Emergency Procedure Preparation (CEN-152, Rev.
01) and found satisfactory.
In addition, all emergency procedures are, in place in the control room of Unit 1, properly approved, and with the latest revisions incorporate While NRR still has comments that are being incorporated into the procedure, it has been found satisfactory from a Region V ISZ standpoint, and no further action is required by'egion V.
I.C.5 Feedback of 0 eratin Ex erience (0 en)
Comments concerning previously identified deficiencies, discussed in Inspection Report 84-47, have been incorporated by the ISEG into their procedure for Operating Experience Review.
However, only a draft version of the procedure (ANPP Procedure 7I405.02.01)
was available for the inspector's review, This item will remain open until the pr'ocedure is approved.
f II.B.4 Trainin for Miti atin Core Dama e (0 en)
The radiation monitoring portion of Mitigating Core Damage was originally missed as part of the initial training for operators and management personnel (See Inspection Report 84-43).
Four STAs and two licensed operators still remain to be trained in this portion that was originally missed.
The licensee training representative committed to completing the training of these individuals on Wednesday, November 28, 1984.
This item will remain open until this training is complete.
II.D.3 Pressurizer Safet Valve Position Indication (0 en)
A work order package has been generated to perform the initial calibration of the acoustic monitors that detect safety valve position.
The calibration data was provided by the vendor from lift tests performed on similar relief valves.
The completed work package will be reviewed by the inspector.
This item will remain open until this review is. complete, with the understanding that the work order package will be completed by Monday, November 26, 1984, if no difficulties are encountered.
New Items II.F.2 Instrumentation for Detection of Inade uate Core Coolin
~Oen NRC Position References:
1)
NUREG-0737 2)
NUREG-0578 1.
Licensees shall develop procedures to be used by the operator to recognize inadequate core cooling with currently available instrumentation.
2.
In addition, each PWR shall install a primary coolant saturation meter to.provide on-line indication of coolant saturation condition.
Operator instruction as to use of this
meter shall include consideration that it is not to be used exclusive of other related plant parameters.
3.
Licensees shall provide a description of any additional instrumentation or controls (primary or backup) proposed for the plant to supplement those devices cited in the preceding section giving an unambiguous, easy-to-interpret indication of inadequate core cooling.
Licensee Commitment Reference:
PVNGS TMI-2 Lessons Learned Implementation Report In summary, the licensee states:
Emergency procedures are being developed which will provide the reactor operators with directions to detect and recover from conditions that could lead to inadequate core cooling and will be in place prior to fuel load.
The NSSS vendor will prepare emergency operating procedure guidelines.
Inadequate core cooling monitoring requirements can be met by appropriately measuring and displaying margin to saturation, reactor vessel water level above the core, and reactor core exit temperatures.
In order to accomplish these functions, the PVNGS Units 1, 2, and 3 design will utilize three instruments:
(1)
The Subcooled Margin Monitor to measure saturation/superheat margin, (2)
The Heated Junction Thermocouple System to monitor level/temperature in the region of the upper portion of the reactor vessel, and (3)
Core Exit Thermocouples to measure the temperature of the reactor coolant as it leaves the reactor core.
The Subcooled Margin Monitor is an instrument which indicates the degree of subcooling and superheat in the reactor coolant either in terms of temperature or pressure.
The system is designed as a
two-channel system, each channel'measuring four points within the reactor coolant system as follows."
Reactor Coolant Piping Hot Leg Temperature (1)
Reactor Coolant Piping Cold Leg Temperature (2)
Pressurizer Pressure
. (2)
The range of the Subcooled Margin Monitor"is increased by incorporation of signals from'he Heated Junction Thermocouple System and the Core Exit Thermocouples such that the instrument can also calculate and display degrees superheat.
Signal processing and display in the control room of the inadequate core cooling instrumentation is performed by the Qualified Safety Parameter Display System (QSPDS) of which, there are two channels.
The redundant.
QSPDS is powered from redundant Class lE busse I ~
F
~
'p
Ins ector Findin s The inspector reviewed all recovery operation emergency procedures to ensure they contain guidance for operators to recognize inadequate core cooling.
This has been adequately addressed.
Pour indications of adequate core cooling are used in each procedure.
Also, the operators are trained to recognize adequate core cooling.
This aspect is considered closed.
2)
A subcooling margin monitor is installed with two remote recorders to record degrees of subcooling or superheat.
The licensed operators have been trained in the use of the subcooling margin monitor and are also taught not to rely on this as the only means of indicating subcooling.
The inspector verified that the subcooling margin recorders are powered from Class lE electrical power sources.
The applicable prerequisite testing of QSPDS has been completed satisfactorily.
However, preoperational testing of gSPDS is still ongoing.
This aspect will remain open until the inspector has reviewed the completed gSPDS preoperational test.
3)
Additional instrumentation for detection of inadequate core cooling has yet to be reviewed by the inspector for and installation and operational status.
This aspect is open.
II.E.4.2 Containment Isolation De endabilit (0 en)
NRC Position References:
NUREG-0737 (2)
Containment isolation system designs shall comply with the recommendations of Standard Review Plan Section 6.2.4 (i.e.,
that there be diversity in the parameters sensed for the initiation of containment isolation),
'4 All plant personnel shall give careful consideration to the definition of essential and nonessential systems, identify each system determined to be essential, identify each system determined to be. nonessential, describe the basis for selection of each essential system, modify their containment isolation designs accordingly, and report the results of the reevaluation to the NRC.
(3)
All nonessential systems shall be automatically isolated by the containment isolation signal.
(4)
The design of control systems for automatic containment isolation valves shall be such that resetting the isolation signal will not result in the automatic reopening of containment isolation valves.
Reopening of containment isolation valves shall require deliberate operator actio J P
t h
A
(5)
The containment setpoint pressure that initiates containment isolation for nonessential penetrations must be reduced to the minimum compatible with normal operating conditions.
(6)
Containment purge valves that do not satisfy the operability criteria set forth in Branch Technical Position CSB 6-4 or the Staff Interim Position of October 23, 1979 must be sealed closed as defined in SRP 6.2.4, item II.3.f during operational conditions 1, 2, 3, and 4.
Furthermore, these valves must be verified to be closed at least every 31 days'7)
Containment purge and vent isolation valves must close on a
high radiation signal.
Licensee Commitment References:
PVNGS TMI-2 Lessons Iearned Implementation Report In summary, the licensee states:
1)
=A Containment Isolation Actuation Signal (CIAS) is diversely generated by either a high containment pressure or a low pressurizer pressure.
2)
Essential and non-essential systems have been defined.
3)
All non-essential systems are automatically isolated on a CIAS with some exceptions.
4)
Resetting of a CIAS does not automatically reopen all containment isolation valves.
Deliberate operator action is required.
5)
Containment, pressure setpoint of 5 psig was selected.
6)
Containment purge valves that do not satisfy the operability criteria set forth in Branch Technical Position CSB 6-4 or the Staff Interim Position of October 23, 1979 must be sealed closed as defined in SRP 6.2.4, item II.3.f during operational conditions 1, 2, 3, and 4.
Furthermore, these valves must be verified to be closed at least every 31 days.
7)
Containment purge and vent isolation valves must close on a high radiation signal.
Ins ector Findin s References:
1)
2)
3)
4)
5)
6)
Procedure 93PE-1SAOl-Integrated Test of ESF Procedure 92GU-OZZ61 Procedure 92PE-1Sg04 Procedure 41ST-1CP02 Procedure 41ST-1CPOl Procedures 92SB-lSB10, ll, 12, 13, 14-
~r P
Preoperational Tests of ESFAS.
7)
Procedure 91HF-1ZZ11 - Safety Valve Respon'se Time Testing The inspector verified that a CIAS is generated from either a
high containment pressure or a low pressurizer pressure signal.
The licensee's definition of essential and non-essential systems has been found acceptable.
However, in the Palo Verde SER, NRC has required that Class lE electrical power be provided to the isolation valves for seal injection and for the.
normal charging path to the plant, since these non-essential systems do not isolate on a CIAS.
The inspector reviewed the applicable electrical schematics and verified that Class 1E power has been supplied to valves CH-HV255 and CH-HV524 (RCP seal injection and charging paths).
The inspector verified that all systems identified as non-essential by the licensee and as isolating on a CIAS do, in fact, isolate on a CIAS, based on evidence in Procedure 93PE-1SAOl.
However, the inspector identified two Post-Accident Sampling System Containment Isolation Valves that were not stroke time tested in accordance with Procedure 91HF-1ZZll.
In addition, the acceptance criteria for the valves'troke time is not contained in the Technical Specifications.
The licensee indicated that these valves would be stroke time tested prior to licensing.
This aspect will remain open until resolution of these concerns.
The inspector verified that resetting the CIAS does not result in the automatic reopening of containment isolation valves and that they must be opened by deliberate operator action.
The inspector verified that the containment pressure setpoint for a CIAS has been reduced to 3 psig, which is consistent with the final version of the Technical Specifications'he eight-inch Power Access Purge Valves have been demonstrated operable per BTP CSB 6-4 as documented in Supplement 5 to the Palo Verde SER.
This allows limited open time for the valves in Modes 1-4 in the Technical Specifications.
However, the 42-inch Refueling Mode Purge Valves have not been demonstrated operable and therefore must be sealed shut in Modes 1-4.
The inspector verified that a surveillance procedure is in place to check shut the Refueling Purge Valves every 31 days and to record the accumulated open time of the Power Access Purge Valves once weekly.
The inspector verified that Containment Power Access Purge Valves and Refueling Purge Valves close on a High Radiation signal as well as a CIAS.
Radiation monitors RU-37 and RU-38 monitor both streams and generate a closing signal from a high
~ J l?
I lt
'I
)
radiation setpoint.
All applicable calibration, prerequisite testing and preoperational testing has been completed satisfactorily.
4.
(Closed) IE Bulletin 83-07 (Includin Su lements 1 and 2) Fraudulent Products b
Ra Hiller Incor orated This bulletin required licensees to determine if they had received products from Ray Miller Inc., to determine where they were installed, to evaluate the safety significance and to dispose any unused materials.
The licensee initially responded to this bulletin on March 23, 1984, but did not specifically address Supplements 1 and 2 of the bulletin, which identified additional vendors which had received Ray Miller Inc. products and may have used them in components supplied to nuclear sites.
The NRC requested a clarifying response which was provided on November 7, 1984 (ANPP-31077 WF(}/TJB).
The inspector reviewed the licensee's responses and backup information files.
The licensee's approach to resolve this Bulletin was to require the contractors (that supplied safety-related material to Palo Verde) to review if they or their subvendors had utilized Ray Miller products.
The results of that review were that none of the contractors or subtier vendors have supplied safety-related material to Palo Verde which contains products provided by Ray Miller Inc.
The results are documented in responses form the A/E (Bechtel)
Backup information in the form of subtier vendor responses to Bechtel and CE were available for review and, because of the size of the files, were sampled by the inspector to verify that the subtier vendor responses did not indicate Ray Miller products had been provided to the Palo Verde site.
The only Ray Hiller products identified in the licensee's program were products provided through the NSSS supplier (CE) and involved the non-safet related Boric Acid Batching Tank (CE Purchase Orders 9770392 and 9770393
.
No safety-related products were identified.
This Bulletin is considered closed.
5.
Follow-u of Licensee Actions a.
(Closed) Violation 50-530/84-03-01 Motor Control Center Was Stored Outdoors This violation had been issued because a motor control center (HCC)
had been stored outdoors wrapped in plastic whereas ANSI N45.2.2 requires storage indoor r
The licensee
'response stated that, since the motor control cabinet was designed for outdoor service (a NEMA 3 cabinet),
the construction organization improperly construed that this met ANSI Level "B" indoor storage requirements.
The licensee conducted training to prevent recurrence and performed a review for similar improper storage conditions.
The review identified six additional safety-related liCC's...,A nonconformance report was generated and dispositioned "use-as-is."
The inspector reviewed the vendor technical manual for this motor control center.
The vendor technical manual does not supersede the ANSI requirements but does provide guidance as to what the vendor considers necessary for satisfactory service of a component.
The inspector noted that the vendor permits Level C (outdoor) storage for this item for a period of two to five years.
The control center in question was stored outdoors for four months.
Additionally, the vendor manual states the subject control centers are designed for outdoor service.
The inspector noted the same model motor control centers in service at the cooling towers.
This item is considered closed based on the licensee's actions and based on the vendor technical manual information which indicates the storage conditions were technically satisfactory.
(Closed) Violation (50-528/83-34-20)
Lumber Was Found Stored in Electrical Racewa s durin the CAT Ins ection The inspector examined the licensee's corrective action which was recorded in Corrective Action Report (CAR) C83-1469.
NCR's EX 3668 and EX 3492 had been written and dispositioned to clean the specific raceways.
Procedural controls are included in construction procedures WPP/(}CI 13.0, WPP/gCI 251.1 and WPP/gCI 31.0, which require periodic surveillance, in-process controls, and final acceptance walkdowns for cleanliness.
The licensee presented inspection records (CIPs from WPP/gCI 13.0)
which demonstrated current housekeeping surveillances performed in Units 1, 2, and 3.
The inspector performed an inspection for cable tray cleanliness in Unit 1 Containment and Control Buildings and found cable tray cleanliness generally satisfactory.
Isolated cases of debris were found in the cable spreading room, however, scaffolding was in evidence in many areas and work (fire protection related)
was in progress.
This item is considered closed based on the general cable tray cleanliness, the ongoing licensee inspections for cleanliness and examinations for cleanliness to be conducted in the normal course of future inspections as work is complete ~g Pl
c.
(Closed) Unresolved Item (50-528/84-15-10)
Debris Was Found in the Area of the Unit 1 Reactor Vessel Su ort Columns The licensee presented inspection surveillance and monitoring reports dated October 21, October 20, October 16, September 5,
July 30 and June ll, 1984, which demonstrated that debris had been routinely reported, cleaned and recurred in the normal course of work in that area.
This condition is understandable given the fact that installation of permanent reactor vessel shims and replacement of insulation was ongoing at the time of the NRC observations.
The inspector examined the Unit 1, Reactor Vessel Support Column Pedestal area for debris and damage to the in core instrumentation tubing.
No heavy debris was observed.
Minor items such as insulation wire and tape were observed.
These items should be removed when the temporary staging and lights currently installed are removed.
No damage to the in core instrumentation tubes was observed by the inspector.
The Maintenance Manager stated that he plans to perform a final compartment closeout inspection after fuel load prior to sealing the containment.
Procedure 41ST12209 requires an inspection for loose debris.
This item is considered closed.
)'.
Exit Interview An exit interview was conducted with the licensee 'personnel indicated in paragraph 1 on November 21, 1984.
The scope and findings of the inspection as described in this report were discussed.
jh
)4
h ll