IR 05000409/1980001
| ML19323H847 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 05/30/1980 |
| From: | Baker K, Branch M, Forner W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML19323H844 | List: |
| References | |
| 50-409-80-01, 50-409-80-1, NUDOCS 8006170088 | |
| Download: ML19323H847 (10) | |
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U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT
REGION III
Report No. 50-409/80-01
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Docket No. 50-409 License No. DPR-45 Licensee: Dairyland Power Cooperative 2615 East Avenue - South La Crosse, WI 54601 Facility Name: La Crosse Boiling Water Reactor Inspection At: La Crosse Site, Genoa, WI Inspection Conducted: March 19-20, and May 5-9, 1980 Inspectors:
K. R. Rid ay
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M. W. Branch (May 5-9, 1980)
5/~3c/60 Approved By Sief 5/3o/go Projects Section 3-2 Inspection Summary Inspection on March 19-20 and May 5-9, 1980 (Report No. 50-409/80-01 Areas Inspected: Routine, unannounced inspection of the licensee's quality assurance program; design, design changes, and modifications; plant operations; TMI short-term modifications; and followup actions
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relative to previous items of noncompliance, IE Bulletins, Licensee
Event Reports, and open inspection items. This inspection involved 105 inspector-hours on site by three NRC inspectors.
Results: Of the eight areas inspected, no items of noncompliance or deviations were found in seven areas; one item of noncompliance was found in one area (deficiency - failure to make prompt telephone noti-fication to NRC Operations Center of significtat event).
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DETAILS 1.
Persons Coctacted
- R. Shinshak, Plant Superintendent
- J. Parkyn, Assistant Plant Superintendent
- G. Boyd, Operations Supervisor
- L. Goodman, Operations Engineer
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- L. Krajewski, Health and Safety Supervisor
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- H. Towsley, Quality Assurance Supervisor S. Rafferty, Reactor Engineer W. Angle, Process Engineer
- M. Polsean, Shift Supervisor
- W. Nowicki, Supervisor, Instrument and Electrical R. Wery, QA Specialist G. Joseph, Security and Fire Protection Supervisor In addition, the inspector observed and held discussions with other engineers, plant equipment operators, reactor operators, assistants, and plant attendants.
2.
General The reactor was shut down on April 6, 1980, for equipment installation required by NUREG-0578 which consisted of position indicators for the safety valves and the addition of diverse parameter closure signals and manual resets on several containment isolation valves. Other maintenance work completed during the shutdown included the instal-lation of a double check valve in the Fuel Storage Well drain line, repair of the 1A Forced Circulation Pump seal and repair of one of the generator's hydrogen seals.
On April 26, the reactor was taken critical and training criticals were performed through April 27. At 1921 on April 28, during heat up and at a low power level, the reactor was automatically shut down due to low water level trip while starting up the turbine. Two other automatic shutdowns occurred before the generator was synchro-nized to the grid at 1315 on April 30.
The shutdowns were caused by a failed Control Rod solenoid and to a burned out fuse in the control power circuit caused by a shorted circuit during maintenance activi-ties. The first automatic shut down was not reported within one hour to the NRC Operations Center as required by the new regulation, i
10 CFR 50.72(a). This is considered to be an item of noncompliance of the deficiency level.
l The licensee had issued a new procedure, ACP 2.8, Emergency Phones, dated March 18, 1980, Operating Memo DPC-76, to further clarify what constitutes a "significant event" for reporting within one hour to the NRC Operations Center. Subsequently, on May 1, 1980, the li-censee revised the ACP to include reporting events such as described above. Therefore, the deficiency is considered to be resolved and a response is not required for this ites.
No other. noncompliance items were identified.
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3.
Quality Assurance Program (QA)
The inspectors reviewed changes on revisions and additions made to the QA Program, since the last QA inspection in January 1979, to determine that the changes conform to the QA criteria described in the licensee's QA Program Description submitted in 1976 and approved February 14, 1977. During this time, the licensee made QA Procedure changes that primarily involved format, 9ording and 2 year reviews.
Revised procedures were. reviewed by the inspectors and appeared to
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meet the requirements of the approved QA Program Description and to provide proper instruction in document control. QA Procedures re-viewed during the inspection were:
Procedure Issue Date Title l
ACP 03.1
11-5-79 Quality Assurance Department ACP 05.0
2-11-80 Control of Vendor Evaluation
ACP 05.1
2-11-80 Procurement Document Control ACP 06.2
11-5-79 Preparation & Use of Procedures ACP 08.2
7-31-79 Receiving Inspection ACP 08.4
12-13-78 Receiving Inspection-Radioactive
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Material ACP 09.1
2-11-80 Identification & Control of Material, Parts & Components
ACP 10.1
2-11-80 Control of Special Processes
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ACP 10.3
7-31-79 Control of Weld Material ACP 11.1
2-11-80 Inspections
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ACP 11.2
11-5-79 Reactor Vessel Internal Inspection ACP 12.1
11-5-79 Test Control ACP 13.1
8-22-79 Control of Measuring & Test Equipment ACP 16.0
7-31-79 Quality Assurance Deficiency Reports & Disposition of Non-j conformance Material, Parts &
i Components ACP 19.0
1-26-79 Audits Access, Corrective Action
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& Response i
QAI 1
3-22-79 Preparation, Issuance and Dis-J tribution of the Quality Assurance Program Description j
QAI 3
3-22-79 Controlled Distribution of Quality Assurance Department Documents QAI 5
5-23-79 Audit Frequency and Scheduling QAI 7
7-31-79 Requirement for Issuance and
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I Control of Deficiency Reports, and Disposition of Nonconforming Material, Parts & Components QAI 8
3 12-79 QA Personnel Qualification and Training Program
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Procedure Issue Date Title QAI 9
7-31-79 Vendor Evaluation Program QACP 1.4
3-23-79 Liquid Penetrant Examination, Water-Washable, Color Contrast Method QACP 2.1
7-31-79 Receiving Inspection The inspectors determined that the* weekly menos to the Plant Superin-
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tendent do not appear to contain sufficient information to ensure that the Plant Superintendent is kept aware of all deficiences discovered or recommended corrective action. This is considered to be an un-resolved item No. 50-409/80-01-01.
The inspectors also determined that outstanding inspection item 79-07,1/
revision of the QA Program Description, had not been completed. A draft copy was being reviewed. The licensee stated that the QAPD would be completed by September 1980.
No items of noncompliance were identified.
4.
Design, Design Changes, and Modifications Five design changes from five reactor systems were selected and re-viewed tu determine that the changes were made in conformance to regulations, Technical Specifications, and implementing procedures.
The Facility Changes reviewed were-
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FC-78-12, Reactor feed pump oil system.
FC-78-09, Forced circulation pump non-contract vibration pickup.
I FC-79-06, Modification of type "A" integrated leak rate test system.
FC-79-08, Addition of switch in CRD seccadary position indication l
system.
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l FC-64-80-01, Installation of position indicators on steam safety I
valves.
The design, design changes and modifications in the above Facility Change files were reviewed to determine that the following areas had been evaluated and documented:
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a.
Provisions for unreviewed safety questions, 10 CFR 50.59.
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b.
Technical Specification requirements.
c.
Specification of Quality Codes and Standards.
d.
Independent reviews by qualified personnel for design, quality, and health and safety.
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IE Inspection Report No. 50-409/79-02, Paragraph 3.
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e.
Special procedures requirements, test and inspections.
The inspector reviewed procedures and design drawings to determine that the above changes had been incorporated in the applicable pro-cedures and master drawings..
Inspections revealed the following:
f.
For FC-64-80-01 and FC 78-12,' facility drawings have not been
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updated to reflect the installations. This is considered to be an unresolved items (50-409/80-01-02).
g.
For FC-64-80-01, Operating and Maintenance Procedures have not been revised to recognize installation. If the PI Detectors are removed for plant testing or testing of the safety valves, i
a procedure or method to ensure proper reinstallation and testing needs to be developed. This is considered to be an unresolved item (50-409/80-01-03).
h.
ACP 07.1, under ORC review, requires ORC meeting minutes to con-tain the statement that the proposed procedure or change does not involve an unreviewed safety question. Meeting minutes of ORC
,do not contain such a statement, but they do contain words to the effect that no problems were found with the procedures under review. This is considered to be an unresolved item (50-409/
80-01-04).
No items of noncompliance were identified.
5.
Review of T erations
.r The inspectors observed plant operations, control room manning, equipment tagging, key-lock control and valve lineups on the shut-down Condenser and Emergency Core Spray systems during a plant tour.
Additionally, the following records were reviewed from April 1, 1980,-to May 5, 1980, to determine compliance with Technical Speci-
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fications and to determine if the Night Order Instructions or Operating Menos conflicted in any way with operational requirements.
a.
S0-18, Daily Log of MW-RMVA-MWH b.
L-84, Control Rod Position Indications c.
L-83, Reactor Plant Logs d.
L-82, Turbine Generator Plant Log e.
L-74, Control Room Panel A f.
_L-42, Radiation Monitor and Dew Cells g.
_ L-119, AGS and Reactor Safety System Log h.
L-72, Turbine Generator Auxiliary Log-1.
L-73, Reactor Plant Auxiliary Log j.
L-68-71, Temperature Recorder Logs k.
Night Order Instruction Book.
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Shift Supervisor Log m.-
Incident Reports-5-
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Jumper and Bypass Key Log o.
Reactor Vessel Heatup and Cooldown Data
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Turbine Operator Shift Log
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Reactor Plant Leakage Log r
No items of noncompliance were identified.
6.
Followup on Licensee Event Reports
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The inspectors reviewed the following LER's to determine if the re-ports were correct and if the evaluations performed and corrective actions taken were appropriate and complete as stated in the LER.
a.
(Closed) LER 50-409/79-17. During the unscheduled shutdo n on November 9, 1979 caused by a sticking turbine control sy-stem, the cooldown rate exceeded the Technical Specification limit of 150*F/hr. The excessive cooldown rate was caused by continuous operation of the shutdown condenser for about ten minutes and above normal vessel water level due to delay in tripping the reactor feedwater pump. The vessel cooldown rate had been exceeded on two previous occasions in 1970 and 1974. Following these events, the vessel stresses were cal-culated by consultants who concluded that the stresses re-mained within the requirements of Section III j
andPressureVesselCodeforNuclearVessels.2/hMEBoiler
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The Vessel I
stresses during this event, which were between the two previous events in severity, were also estimated to be withir e.oder requirements for primary and secondary stresses; ard the inci-dent had negligible effect on vessel or piping usa;,e since th maximum usage factor was greater than 0.0005 (1/2000 cycles).gj b.
(Closed) LER 50-409/80-01. Faiiure! of electrical penetrations No.s 3, 4, and 5 to meet the Type B containment lesk rate test.
Seven of the eight individual penetration leaks we re found to be caused by cracked brass glands. Five of these vere re-placed completely, including new MI cable. The ofher two
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were repaired by soldering. One small thread lead was cor-rected using teflon tape. Only the two small leahs on penetrations 3 and 5 were found to compromise containment integrity. After repairs were made, the penetrations were satisfactorily retested. No leakage was detected on those penetrations sealed with a new sealant used for the first time in early 1979.
c.
(Closed) LER 50-409/80-02. Failure of the insid,, containment building exhaust damper to meet the Type C leak eate test. The outer damper maintained containment integrity. The leakage 2/
Letter UNC to US AEC Division of Reactor Development and Technology dated 6/9/70.
3f Nuclear Engineering Services Report No. NES 81A0014, 9/13/74.
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Nuclear Engineering Services letter to DPC dated 11/28/79.
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was attributed to an indentation in the damper seat ring caused by the valve dise when in the open position. The depressions have been noted before and when the test pressure is applied in the normal direction from inside containment, no leaks have been observed from these indentations. The seat ring was replaced, both exhaust valves repacked and retested satisfactorily.
d.
(Closed) LER 50-409/80-03. Failure of the containment building condensate return isolation valve to meet the Type C leak rate
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test. This was the first failure of this valve and the failure was attributed to normal wear. The valve seals were replaced and the ratest was satisfactory.
No items of noncompliance were identified.
7.
Followup on IE Circulars (IEC).
The inspectors reviewed the licensee's evaluation of the following IECs to determine that any necessary actions had been taken.
a.
IEC 79-13, Replacement of Diesel Fire Pump Starting Contactors.
b.
IEC 79-15, Bursting of High Pressure Hose and Malfunction of Relief Valves on SCUBA.
c.
IEC 80-04, Securing of the Threaded Locking Devices on Safety Related Equipment.
No items of noncompliance were identified.
8.
Followup on IE Bulletins a.
(Closed) IEB 79-27, Loss of Non-Class-I-E Instrumentation and Control Power System Bus During Operation. The licensee re-viewed the design of and emergency procedures for the five
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Class I-E and Non-Class I-E buses supplying power to control systems. During the review, it was noted that the IC inverter did not have a " loss of power" annunciator in the control room.
This has been installed. Reviews requested by IE Circular 79-02 concerning failures of 120-volt AC vital power supplies
were again made and no changes were deemed necessary.
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i b.
(Closed) IEB 80-01, Operability of ADS Valve Pneumatic Supply.
Since LACBWR does not have an Automatic Depressurization Sy-stem (ADS), a review of the Manual Depressurization System (MDS) conducted by the licensee reviewd no mechanism which
l could prevent operation of the MDS when required.
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c.
(Closed) IEB 80-02, Inadequate Quality Assurance for Nuclear Supplied Equipment. None of the equipment referred to in the
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Bulletin is used at LACBWR. The manufacturer had not supplied any other equipment to LACBWR.
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d.
(Closed) IEB 80-03, Loss of Charcoal From Standard Type II, 2-inch, Tray Absorber Cells. LACBWR does not use any Flanders filters and a review of their specially designed filter did not reveal any of the problems associated with them.
e.
'(Closed) IEB 80-07, BWR Jet Pump Assembly Failure. LACBWR does not use jet pumps.
f.
(Closed) IEB 80-09, Hydramotor Actuator Deficiascies. None of the ITT General Controls hydramotor actuato-are in use at l
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No items of noncompliance were identified.
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9.
Followup on Open Inspection Items (OII)
(Closed) (OII 79-06).5/ The licensee had reviewed and approved the I
QA Program of Nuclear Engineering Services, their engineering con-
sultant, and the review had been properly documented on a QA Check-i i
list Qualifications of Suppliers and Contractors dated August 21, 1979.
No items of noncompliance were identified.
10.
Special Inspection Regarding TMI Short Tern Items (NUREG-0568)
On March 19-20, 1980, the inspector attended a meeting /
k between the licensee and members of the Office of Nuclear Reactor Regulation re-l garding short-term commitments made by the licensee in replys to NUREG-0578, TMI Lessons Learned.
During this inspection, the inspector verified that the following
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commitments had been completed by the licensee during the shutdown in April 1980:
a.
NUREG-0578, Item 2.1.3.a, Installation of position indication of
the three safety valves (also see Paragraph 4). The inspection
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facility change procedures and the environmental qualifications of components used were documented and appeared to be satisfac-tory. The backup safety valve position indication has been in use previously and procedures were in place. As noted in para-graph 4, operating and maintenance procedures have not been
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completed. This is considered to be an unresolved item.
b.
NUREG-0578, Item 2.1.4, Installation of manual reset switches on containment isolatten signals to prevent automatic valve opening upon clearance of the isolation signal. The installa-tion was made under Facility Change 79-25 and Maintenance 5/
IE Inspection Report No. 50-409/79-02.
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Letter, Dennis L. Zieman, BRB #2 to Frank Linder, DPC, dated April 25,
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1980.
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Request 3038, which the inspector reviewed for completion and testing. The inspector reviewed procedures to determine.that valve switches associated with the isolation systems affected would be p12ced in the closed position before the circuits were reset.
c.
NUREG-0578, Item 2.1.6.a. The inspector verified that a periodic leak test procedure had been established for the component cooling water system and the off gas ' system. Procedures had been written and the tests conducted prior to plant startup on April 28, 1980.7j
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The inspector reviewed the test results and found them to be satisfactory. The ir.spector noted that the tests had yet to be placed on the refueling surveillance schedule. This is considered to be an unresolved item (50-409/80-01-05).
d.
NUREG-0578, Item 2.2.1.
(1) (a)
The inspector reviewed Operations Memo, DPC-84 dated December 31, 1979, Shift Supervisor's Responsibilities and found it to be adequate to meet NUREG-0578 commit-ments.
(2) (b)
The inspector verified that the licensee had established a Shift Technical Advisor (STA) function by reviewing Administrative Control Procedures. A Shift Technical Advisor Committee meeting had been held to further define the STA Program and operating experience as-sessment.
(3) (c)
The inspector reviewed ACP-2.3, Shift Transition (Turnover), Issue 3, April 3, 1980, to determine that a uniform shift turnover procedure had been established and the inspectors observed that the procedures were being carried out.
e.
NUREG-0578, Item 2.2.2.
(1) a.
The inspector reviewed Operations Memo DPC-83, dated December 31, 1979 to determine that authority to limit access to the Control Room had been established and assigned to the Shif t Supervisor.
(2) b. and c.
The inspector reviewed the following procedures to determine that an adequate plan had been established to activate the Technical Sv7 port Center and Opera-tions Support Center:
ACP 2.7, Issue 0, Technical Support Center, January 11, 1980.
ACP 2.8, Issue 0, Emergency Phones, March 18, 1980.
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Letter, Frank Linder, DPC to Harold Denton, NRR dated May 6, 1980.
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ACP 2.1, Issue 3, Authorities and Responsibilities for LACBWR Operation and Shutdown, November 5, 1979.
EPP 5, Issue 2, Emergency Evacuation of Onsite Per-sonnel, January 6,1980 EPP 6, Issue 2, Emergency Evacuation Point Operation, January 6,1980.
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NUREG-0578, Item, High Point Vents f.
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The inspector reviewed operating and emergency procedures for the use of the Manual'Depressurizing System for venting the reactor and found them to be adequate.
g.
NUREG-0578, Item 2.1.8, Items a, b, and c.
I The inspector verified that the stack noble gas monitor and Control Room and Tecanical Support Center radioiodine monitors were in place. Procedures for calibration and operation were
not reviewed. This is considered to be an unresolved ites
- 50-409/79-01-06).
The above unresolved items in the short term TMI lessons learned program will be reviewed and closed in a future inspection.
No items of noncompliance were identified.
11.
Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whetaer they are acceptable items, items of non-compliance, or deviations. Unresolved items disclosed during the inspection are discussed in Paragraphs 3, 4,10 c. and 10 g.
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12. Exit Interview The inspectors met with the licensee representatives (denoted in Para-graph 1) at the conclusion of the inspection. The inspectors summarized the scope and result of the inspection. In response to several items discussed by the inspectors, the licensee agreed to:
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a.
Revise the approved QA Program Description to bring it into agreement with changes in organization and QA procedure systems, estimated to be complete by' September 1980.
b.
Review ACP 03.1 requirements for weekly memo to Plant Superin-tendent and ensure menos contain sufficient information for the Plant Superintendent to maintain both administrative and tech-nical control'over the QA Department.
I c.
Review ACP 07.1 requirements for ORC meeting minute information and revise requirements or ensure ORC meeting minutes contain the required statements.
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