IR 05000263/1975005
| ML20024G072 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 04/16/1975 |
| From: | Choules N, Jackiw I, Jordan E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20024G069 | List: |
| References | |
| 50-263-75-05, 50-263-75-5, NUDOCS 9102070506 | |
| Download: ML20024G072 (21) | |
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U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCDIENT
REGION III
Report of Operations Inspection
IE Inspection Report No. 050-2'u3/75-05 Licensec:
Northern States Power Corya, <
414 Nicollet Mall Minneapolis, Minnesota 55401 Monticello Nuclear Generating Plant License No. DPR-22 Monticello, Minnesota Category:
C Type of Licensec:
BWR Type of Inspection: Unannounced I. )...
Dates of Inspection: March 11-14, 1975 g
Dates of Previous Inspection: February 18-21, 1975 (REP)
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Principal Inspector:
ou co
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Accompanying Inspector:
ackiw
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Other Accompanying Personnel: None
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Reviewed By:
E. L dordan, Senior Reactor Inspector 8///JT os t-Reactor Operations Branch
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i 910207050675041)63 PDR ADOCK OSOOOm G
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SthiMARY OF FINDINGS.
- Enforcement Action The following noncompliance item was noted during the inspection.
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Violations None Infractions Contrary to 10 CFR Appendix B, Criterion V, a design change modification was made to the HPCI-9 check valve without following the procedures f or controlling design changes as required by the licensee's Power Production Administrative Control Directive No. 3ACD 4.1, Design Change Control.
(Report Details 1. Paragraph 1.c)
This infraction was identified by the licensee and had the potential for causing or contributing to an occurrence with safety significance.
Deficiencies
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None k)
Licencee Action on Previously Identified Enf orcement Matters Maintenance to RHR heat exchanger without detailed procedures.1!
A.
The licensee has prepared RHR maintenance procedures and his corrective action appears appropriate.
(Report Details I, Paragraph 6.a)
Design Change without proper review.2/ The licensee has issued new B.
design change instructions which should prevent reoccurrence of this type noncompliance item.
(Report Details I, Paragraph 6.b)
Failgye to record independent verification signatures in Bypass / Jumper C.
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The licensee has implemented a new Bypass / Jumper form which should prevent reoccurrence of this type noncompliance item.
(Report Details I, Paragraph 6.c)
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IE Inspection Rpt No. 050-263/75-01 2/
RO Inspection Rpt No. 050-263/74-10.
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Unusual Occurrences A.
Six containment isolation check valves failed to meet the minimum"'
leakage requirements of the Technical Specifications.
(Report Details I, Paragraph 1.b)
s B.
A hydraulic snubber located outside the drywell on the shutdown cooling suction line was inoperable due to loss of oil which resulted from a cut seal.
(Report Details 1, Paragraph 1.f)
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Other Significant Findings A.
The plant has been operating at approximately 100% power since the January refueling outage.
The radiation icvel at the of f gas air ejector monitor at the time of the inspection and at 100% power was 8R/hr and decreasing compared to 10 R/hr at 75% power prior to the refueling outage.
B.
The licensee's augmented off gas system is in operation but the licensee does not consider it completely operable due to several problems associated with the system.
(Report Details I, Paragraph 7)
Management Interview The following persons were present at the management interview conducted at the conclusion of the inspection on March 14, 1975.
The inspection was (l also discussed with C. E. Larson, Plant Manager, by telephone on March 19, 1975.
M. H. Clarity, Superintendent, Plant Engineering and Radiation Protection D. D. Antony, Plant Engineer, Operations W. A. Shamla, Plant Engineer, Technical W. A. Sparrow, Operations Supervisor A.
Abnormal Occurrences 1.
A0 263/74-25. The inspector stated that a supplemental report to the original abnormal occurrence report should be made since final corrective action was not indicated on the original report.
The licensee stated that a supplemental report will be issued.
(Report Detailo I, Paragraph 1.n)
2.
A0 263/75-01.
The inspector stated that in discussion with licensee representatives it was indicated that 'ue to the
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continuing leakage problems the licensee is evaluating the replacement of some of the check v associated with the t-3-
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occurrence.
The licensee stated that they were evaluating replacement of ECCS (Emergency Core Cooling System) check
--I valves with hydraulically operated valves or adding hydraulically operated valves in addition to the check valves.
(Report Details I, Paragraph 1.b).
3.
A0 263/75-04, 75-05, 75-06.
The inspector stated that corrective actions for the infractions associated with AO's 75-04 and 05 which were identified by the licensee's management control systems appeared to be appropriate.
The inspector stated that failure to perform an engineering analysis and safety review preceeding modification of the HPCI exhaust line check valve, which was identified by the licensee as A0 75-06, was considered to be a citable infraction.
The inspector stated that specific corrective action was considered to be appropriate.
The inspector stated that discrepancies in completing Work Request Authorizations associated with A0 75-05 and 06 may have contributed to the occurrences. Licensee representatives stated that they planned to improve their internal surveillance of work requests to prevent future similar occurrences.
(Report Details I, Paragraphs 1.c, d and c)
4.
A0 263/75-07. The inspector stated that the licensee's corrective actions for this occurrence appeared to be appropriate.
(Report Details 1, Paragraph 1.f)
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B.
Surv'eillance Tisting The inspector stated that surveillance testing was reviewed and no items of noncompliance were identified.
The inspector stated that the licensee's representative indicated that Surveillance test 0120, Reactor Vessel and licad Flange Temperature Limitation, will be revised to indicate where the temperature measurements are to be taken. The licensee acknowledged that the test procedure will be revised.
(Report Details I, Paragraph 2)
C.
Limiting Conditions for Operations (LCO)
The inspector stated a sampling of LCO's, safety limits and limiting safety systems settings had been reviewed and no items of non-compliance were identified.
(Report Details 1, Paragraph 3)
D.
Regulatory Operations Bulletins (ROB's)
The inspector stated that the licensee's action,s.,in regards to ROB's 74-9,74-10B, 74-13, 74-14, and 74-15 were reviewed and appear to be adequate to comply with the bulletins.
(Report Details 1 Paragraph 4)
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E.
Outstanding Items
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Seven outstanding items, including feedwater orifice welds, out of calibration instruments, recirculation pump leakage annunciators, control rod drive valve wedge inspection, Operations Manual Volume A revision, Residual Heat Removal (RHR) Heat Exchange Leak, and control rod drive line liquid penetrant inspection were reviewed with the licensee.
(Report Details, I, Paragraph 5)
F.
Previous Enforcement Action The inspector stated that review of the licensee's actions in regards to noncompliance items related to Bypass / Jumper logs and lack of procedure for RHR heat exchanger repair indicated the actions were adequate.
(Report Details, Paragraph 6)
G.
Augmented Off Gas System Problems with the subject system were discussed.
The licensee indicated that although the system was presently operating, they did not consider it to be fully operational due to a number of problems with the system which remain to be resolved.
(Report Details 1, Paragraph 7)
H.
Condensate Demineralizer Problem i\\l A problem of resin getting into the reactor coolant system from the subject demineralizer was discussed.
The licensee stated they had found some bad element gaskets in the demineralizer and believed this allowed the resin to be transported into the reactor coolant system.
(Report Details 1. Paragraph 8)
I.
Moisture Separator No. 3 Inlet Steam Piping The degradation of the subject piping was reviewed. The licensee stated that they were evaluating the reportability of this item as an abnormal occurrence.
(Report Details 1, Paragraph 9)
J.
Procedures 1.
The inspector stated that Operations Procedure B.4.2, revision dated January 7, 1975, had not been issued. The licensee stated that this procedure was delayed due to the refueling outage and will be issued shortly.
(Report Details II, Paragraph 1.b)
2.
The inspector stated that a number of Operations Procedures were not reviewed in a 2 year period as required by Technical Specifications. The licensee acknowledged this and 8tated-5-i f
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that these procedures received a preliminary safety review in January 1975 and will receive a complete review within 2
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T years of the time the technical specification requirement was issued, which was April 1973.
(Report Details II, Paragraph 1.c)
3.
The inspector stated that Temporary Memos are not being signed of f to indicate Operations Committee review and approval.
The licensce acknowledged this and stated that steps will be taken to assure that in the future Temporary Memos will be signed in a timely mannet.
(OcsLivu II, Paragraph 1.e)
K.
Plant Operations 1.
The inspector stated that the "SBLC Hi/Lo Temp" alarm in the control room appears to be continuously in the tripped condition.
The licensee acknowledged this and stated that corrective action vill be taken to eliminate this problem.
(Section II, Paragraph 2.a)
2.
The inspector stated that during a plant tour two areas of concern were identified and appear to conflict with good housekeeping practices. The licensee stated that these areas would be cleaned up.
(Section II, Paragraph 2.a)
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l REPORT DETAILS
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Persons Contacted Northern States Power Company C. E. Larson, Plant Manager
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M. H. Clarity, Superintendent, Plant Engineering and Radiation Protection W. E. Anderson, Superintendent, Operation and Maintenance W. A. Sparrow, Operations Sopervisor H. E. Nimo, Maintenance Supervisor P. A. Pochops, Quality Engineer D. D. Antony, Plant Engineer Operations W. A. Shamla, Plant Engineer, Technical L. L. Nolan, Engineer M. F. Hammer, Engineer B. D. Day, Engineer B. L. Jeness, Engineer W. J. Hill, Engineer, Instruments and controls O.
Iverson, Engineer D. E. Nevinski, Engineer Nuclear F. J. Schober, Shift Supervisor (,
Nuclear Services Corporation
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..s,~e 3. Woolf, Quality Assurance Engineer 1.
Abnormal Occurrences
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a.
A0 263/74-25 The licensee informed the inspector on December 31, 1974, by telephone that the RCIC outboard steam isolation valve, MO-2076 failed to close during surveillance tetting.
The details 7 this occurrence are described in the licensee f
report.-
The licensee plans to submit an additional report detailing the final corrective action.
This is the second time this valve f ailed to close.5/
The original failure was attributed to a backscating problem.
Investigation by the licensee indicated the f ailures were caused by the valve operator limit switch bypass span being set too small which did not allow the torque switch to reset when given an "0 PEN" signal. The licensee also replaced the' grease in the valve operator which showed some deterioration, s
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AO Rpt No. 263/74-25, NSP to DL 1/9/73.
jif AP Rpt No. 263/74-23, NSP to DL 9/5/74.
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b.
A0 263/75-01 This abnormal occurrence consisted of a failure of six containment isolation check valves to meet the minimum leakage requirements of Section 4.7.A.2(c) of the technical specifications.
The check valves were found to be leaking excessively when the licensee performed local leak rate tests as required by the thisoccurrencearedescribedinthelicensee'sreport.gyionfor technical specifications.. The details and corrective a The inspector in reviewing this abnormal occurrence noted that all six check valves had previous histories of leaking excessively.
The inspector inquired if the licensee had any plans to impro.a or replace these valves.
The licensee representative indicated that Bechtel had performed a study of the valve problem and recommended the replacement of the check valves with another type valve.
He stated that the licensee was evaluating this recommendation, c.
A0 263/75-04 The licensee informed the inspector by telephone on January 23, 1975, that during refueling operations a rod position interlock was bypassed due to false rod positicn indication which resulted from damaged control rod drive (CRD) connector.
Bypassing the rod position interlock makes (j it possibic to withdraw the control rod while the refueling
bridge is over the core which is an infraction of Technical The details and the corrective actio Specification 2.10.A.for this occurrence are described in the licensee's repor In discussion with the licensee's representative, it was indicated that about two hours after the bypass was installed.
the shif t supervisor determined that a mistake had been made and took proper corrective action.
This was documented in the shift supervisor log.
d.
A0 263/75-05 The licensee informed the inspector by telephone on February 10, 1975, that a relay coil in the control circuit for the reactor water cleanup isolation valve, M0-2399, shorted out resulting in loss of valve control power.
The licensee also determined
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A0 Rpt No. 263/75-01, NSP to DL 2/10/75.
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AG Kpt No. 263/75-04, NSP to DL 1/31/75.
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during management review of the occurrence report that the valve was not closed and was an infraction of Technical Specification 3.7,0.2 which requires the valve to be closed when it is inoperable.
It would have been possible to manually close the valve even though control power was not available.
The details and the
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corrective action pr this occurrence are described in the
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licenste's report.
The inspector reviewed the work request authorization (WRA) by which repairs to the valve control circuit were made.
Two discrepancies were noted on the work request. in that the system was not identified as a critical system and technical specifications considerations were not identified.
Section II of the WRA has appropriate boxes to check to identify these considerations.
The inspector discussed these discrepancies with the licensee's representatives and stated that the infraction of the technical specification might have been avoided if the work request had been filled out properly. The licensee's representative agreed with inspector's statement.
e.
A0 263/75-06 The licensee informed the inspector by telephone on March 14, 1975, that during the review of A0 263/75-01 it was determined that a modification had been made to the liigh Pressure Coolant y;q Injection exhaust line check valve, HPCI-9,,without a proper
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engineering analysis and safety review being performed.
The details and corrective actigys of this occurrence are given in the licensee's report The inspector reviewed this occurrence with licensee's representatives.
The llPCI valve was found to be Icaking dur{g9 containment isolation valve Icakage tests and was reported.- '
A work request was issued to inspect and repair the valve.
When the valve vas disassembled, the hinge arm was found to be bent.
Maintenance personnel then proceeded to straighten the hinge and add a stiffener to the hinge without any approval to do Adding the stiffener is a design change to a safety related so.
system.
The licensee administrative procedures which control design changes is Power Production Administrative Control Procedure No. 3 ACD 4.1, Design Change Control.
This directive requires preparation of a Design Change Package and details the approvals, evaluations and reviews required for proposed design changes.
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A0 Rpt No. 263/75-05, NSP Lp DL 2/19/75 9/
A0 Rpt No. 263/75-06, NSP to DL 2/24/75.
10/ A0 Rpt No. 263/75-01, NSP 2/10/75.
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Procedure No. 3 ACD 4.1 was not followed when the HPCI-9 valve modification was made.
This is an infraction of 10 CFR 50, Appendix B, Criterion V, which requices that activities affecting quality be accomplished in accordance with documented instructions.
When licensee management discovered that modification had been made to the HPCI-9 valve, an evaluation of the codification was made and dye penetrant test of the welds associated with the modification were performed.
The licensee's evaluation indicated that if no cracks were found by dye penetrant test, the modification could remain but the valve hinge should be replaced as soon as possible with a new or modified hinge.
No cracks were found.
Review of the work requests (2) associated with the repair and subsequent dye penetrant test showed the system was not identified as a critical system and technical specification considerations were not identified.
(A similar discrepancy was discussed in paragraph 1.d) The inspector reviewed surveillance test records for February 20 and 21, 1975, when the dye penetrant test was performed and verified that operability checks of systems required by the technical specifications when the HPC1 was inoperable were performed even though the work request did not specify that the operability check be performed. The modification of the valve was performed during tha refueling
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"I outa'g'e and operability checks were not required at that time?*v
The licensce's corrective action to assure that quality assurance
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procedures are adhered to was a presentation on quallt.y
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assurance to the maintenance personnel on March 7, 1975, by a quality assurance engineer.
f.
A0 263/75-07 Th? licensee reported to the inspector on February 18 1975, by telephone, that a hydraulic snubber located outside the drywell on the shutdown cooling suction line was inoperabic due to a loss of oil.
The details and corrective occurrencearegiveninthelicensee'sreport.gypionforthis
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In discussion with the licensee's representative, it was indicated that the probabic cause of the oil leaking out of the snubber was that a manifold connector which passes through an 0-ring had tool marks on it and these tool marks apparently 11/ A0 Rpt No. 263/75-07, NSP to DL 2/27/75.
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cut the 0-ring allowing the oil to leak out.
The licensee
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could not determine if the tool marks resulted from original assembly by she vendor or during seal replacement by the licensee.
The licensee indicated that they had informed their maintenance personnel of the occurrence and cautioned them on use of tools during assembly of snubbers.
The inspector
reviewed a procedure which the licensee uses to replace snubbers seals and it should be adequato to assure proper assembly of snubbers.
2.
Surveillance Testing The inspector selected several technical specification surveillance test;ng requirements and verified that the licensee has surveillance test p
.edures which accomplished the required surveillance testing.
Surveillance tests which were reviewed are as follows:
Number Title 0007 Condenser to Vacuum Scram 0012 APRM/ Flow Reference Scram 0022 Condenser to Lo Vacuum Scram Calibration 0023A Turbine Control Valve Fast Closure Scram Calibration 0070 Radiation Monitoring Off-Gas Isolation Test 0073 Radiation Monitoring Off-Cas Isolation Sensor Check 0087 Standby Liquid Control System Explosive Charge Check f- )
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0100B%**
LPCI Pump Operability
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0102 LPCI Flow Rate Test 0112 Auto Pressure Relief Valve Operability 0113 Auto Pressure Relief Simulated AUTO Actuation 0116 RClc Flow Rate Test
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0120 Reactor 11 cad Flange Temp Limitations 0133 Reactor Jet Pump Operability 0150 Standby Gas Treatment Auto Initiation 0152 Primary Containment Isolation Valves Group 1 0187 Standby Diesel Generators Load Test 0189 Standby Diesel Generators Simulate Auto Start 0195 250 V Battery Quarterly Check 0198 125 V Battery Discharge Test Review of these surveillance test procedures showed that prerequisites and preparation of the tests are specified, acceptance criteria are specified, and operational checks prior to returning equipment to service are specified when required.
Test procedures 0120 did not specify general locations on the head and vessel where temperature measurements are to be made. The licensee's representative stated
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that the test procedure would be revised to indicate the areas where
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measurements are to be made. Review of a sampling of completed
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test procedures for the above tests performed during Harch 1974 to March 1975 indicated that they are being performed at the frequency required by the technical specifications, the test results were in
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conformance with the technical specifications and the completed procedures were reviewed by somebne other than the tester or individual
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directing the test.
3.
Limiting Conditions for Operations During the review of the completed surveillance test procedures described in paragraph 2 the inspector verified that the limiting conditions for operations, safety setting, and limiting safety settings for the systems associated with these surveillance tests were met as required by'the technical specifications. The inspector verified by visual observation of control room panels that the HPCI, RCIC, LPCI, Auto Depressurization. Emergency
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Diesel Generator offsite power and associated buses. 125 y and 250 V.
batteries systems were lined up for operation as required by the technical apecification and no components were indicated to be out of service.
A review of the local. power range monitor (LPRM) system with the licensee's representative indicated that the number of operable.
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Local Pgwep* Range Monitors (LPRM) was greater than the minimum requite 5 by the tichnihil' specification."Revie8*9f he:at balances v~
and resetting of Average Power Range Monitors performed per surveillance test 0017, APRM Scram Instrument Calibration from April 1974 to March 1975 indicated the test was performed every three
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days as required by the technical specifications. The heat balances indicated that 100% power was not exceeded.
Surveillance test results for test 0124, Reactor Coolant Chloride / Conductivity)
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indicated samples were taken as required by the technical specifications with the reactor shutdown on January 10, 1975, and during startup on February 6 and 7, 1975, and the conductivity and chlorides were less than the maximum allowed by the technical specifications.
Surveillance test results for Test 0002, high i
pressure reactor scram, indicated that the pressure settings were all set atf1 1068 psig on February 12, 1975, which is less than the maximum of 1075 psig allowed by the technical specification.
Surveillance test results for test 0128 reactor safety / relief valve check indicated that the seven reactor safety / relief valves were all' tested to relieve at f;1080 psig on November 21, 1974, as required by the technical specifications.
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4.
Regulatory Operations Bulletina (ROB's)
RO Bulletin 74-9, Deficiency in General Electric Model a.
_4 KV Magne-Blast Breakers The inspector reviewed the work request by which the licensee inspected the subject breakers for trip roller bar clearances and de'termined there wa g possible interference as described in the licensee's reply-to the bulletin.
b.
RO Bulletin 74-10B, Failures in 4-Inch Bypass Piping at Dresden 2 The inspector reviewed the licensee operating instructions which will assure compgnce with the bulletin as described in the licensee's reply-to this bulletin.
R0 Bulletin 74-13, Improper Factory Wiring on General Electric c.
Motor Control Centers at Fort Calhoun The inspector reviewed preventctive maintenance procedures PH 4831, PM 4832 PM 4842, and Work Request Authorization El-153whichindicatedthatthelicenseehgjerformedthe p
inspection as stated in the licensee reply-- to the bulletin.
d.
RO Bulletin 74-14, BWR Relief Valve Discharte to Suppression Pool The' inspector reviewed the licensee's operating instructions
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whichwillassurecompliancewigjthe subject bulletin as stated in the licensee's reply.-
e.
RO Bulletin 74-15, Misapplication of Cutler Hammer Three Position Maintained Switch Model No. 10250T The inspector reviewed a memo concerning the potential for mispositioning the subject switches which was issued gj operating personnel as stated in the licensce's reply-to the subject bullecin.
5.
Outstanding items The following items were reviewed.
12/ NSP to Director, Region III 2/3/75.
_13/ NSP to Director, Region III 1/3/75 M / NSP to Director, Region III 12/19/74.
15/ NSP to Director, Region III 11/21/74.
16_/ NSF to Director, Region III 1/3/75.
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Crac'ted Feedwater Orifice Welds a.
Thelicenseepreviouslyreportedthatoneoffourwe}jy for feedwater orifices had cracked and was repaired.--
During the January outage the remaining three welds were ground out and rewelded. The licensee indicated there were no cracks in the three welds.
b.
Inoperative and/or Out of Calibration Instrument The inspector inquired if anything had been done to identify -
instruments which were inoperative and/or out of calibration.18/
The licensee representative stated they were updating their computer input so that the output would identify out of calibration instruments. The inspector inquired how inoperable instruments would be identified.
The licensee representative stated that a " BAD" sticker and/or work request stickers have been used, but that formal instructions did not exist to accomplish this.
The licensee's representative stated they would review this matter further.
Recirculation Pump Seal 1.eakage Annunciators c.
On a previous inspection, it was noted that the subject annuncirtors were lit all the time.
The licensee's representative stated that a design change proposal was in progress to remove the annunciators.
lie stated that pump seal leakage would be*Ik l
detected by monitoring seal pressure and drywell sump pumping rate.
Final resolution of this item has not yet been completed, d.
Control Rod Drive Valve Wedge Inspection
The licensee previously indicated that three control rod drive vggye wedges would be inspected during the January ou t a ge.--
Tne licensee inspected three wedges during the
,e January outage and no cracks were detected, llowever, there were oxidation products on the rough surf aces of the wedges.
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The licensee contacted the vendor regarding the oxidation and they indicated that they did not consider this a problem.
The wedges are made from 410 stainless steel which is less resistant to oxidation than austenitic steel. As a precaution, the licensee is polishing the rouga surfaces of any replacement wedges which are installed.
j 17/ A0 Rpt No. 263/74-19, NSP to DL, 6/24/74.
18/ R0 Inspection Rpt No. 050-263/74-10.
19/ Ibid.
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The inspector verified that the licensee has the Power Production Directive 3ACD 4.1, Design Change Control, in his
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stated in the licensee reply--
Proper use of the directive should help prevent reoccurrence of this noncompliance item.
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Augmented Off-Gas System The inspector discussed the subject system with the licensee representatives.
The system was in full operation during the inspection, but several problems still exist with the system and the licensee does not consider it to be fully operabic.
Problems which remain to be resolved are as follows:
a.
Moisture accumulates in the sample line to the hydrogen analyzers and cause false hydrogen indication.
There have been spurious trips of the recombiner trains due to this problem.
b.
Airborne activity problems due to leaks.
c.
Eductor water problem.
d.
Inlet flow meter to recombiner range and moisture problems.
e.
Noise interference from switchyard on the recombiner (,
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instrumentation.
8.
Condensate Demineralizer Problem The licensee has experienced problems of demineralizer resin getting into the reactor coolant system from the D condensate demineralizer.
This occurred on February 27, 1975, and again on March 12, 1975.
In both cases, the pH dropped from about 7.5 to about 6.3, conductivity increased and there was a drop of 8 to 15 megawatt in thermal power.
According to the licensee, a change in thermal power with a change in pH has previously been observed in experiments at Shippingport and Yankee Rowe.
For the February 27 occur-cace, the problem was identified as one of the 302 demineralizer elements becoming unseated. Rescating the element appeared to correct the problem although several small conductivity spikes were observed and were attributed to resin trapped in the basket strainer downstream of the demineralizer.
26/ Ibid.
jl/ RO Inspection Rpt No. 050-263/74-01.
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After the March 12 occurrence, the licensee removed all 302
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elements from the demineralizer vessel and discovered that 4 of the seating gaskets for the elements were gouged and 7 cups which support the gaskets were out of round.
The gouged gaskets apparently allowed the demineralizer resin to escape.
The licensee replaced the gouged gaskets and out of round cups and cleaned the basket
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strainer. The licensee indicated in a telephone conversation on March 27, 1975, that no further evidence of resin getting into the reactor coolant has been observed.
9.
Turbine Moisture Separator No. 3 Inlet Steam piping The licensee has made wall thickness measurements of the subject piping during the past three refueling outages.
The average thickness of the active errosion areas of the subject pipe are as follows:
April 1973 - 0.507 inches April 1974 - 0.474 inches January 1975 - 0.405 inches The errosion rate since April 1974 has increased. The licensee is evaluating when corrective action will have to be taken.
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REPORT DETAILS S 'tio I
Prepared by: C [ > LM 7I
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d a 1. N.
hckiw (Date)
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H. C. Dance
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~+ersons Contacted
.. Antony - Plant Engineer, Operations
. Clarity - Superintendent, Plant Engineering and Radiation Protection
.. Wolfe - Quality Engineer
~__ _. Day - Engineer
.. Sparrow - Operations Supervisor
._.. Rodgers - Plant Equipment and Reactor Operator C. Schober - Shift Supervisor Procedures
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a.
The inspector confirmed that the following Operations Procedures were reviewed and approved by the licensee in accordance with the Technical Specifications.
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(1)
B.2.2 Reactor Clean Up System (2)
B.2.3 Reactor Core Isolation Cooling System (3)
B.4.1 Primary Containment Cooling & Ventilation System (4)
B.2.4 Main Steam System (5)
B.6.5 Feed Water System (6)
B.6.4 Circulating Water System (7)
B.9 Electrical Systems (8)
B.1.1 Reactor llead Seal Detection System (9)
B.5.1.1 Process Radiation Monitoring The inspector noted that each of the above procedures contains a listing of current pages.
This listing also indicates the date of the latest revision.
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b.
The inspector reviewed the Operations Manual Section B.4 and noted that procedure B.4.2, Standby Gas Treatment System, which was reviewed and approved by the Operations Committee on January 7,1975, had not been issued as of this inspection.
DurinA the review of procedures in the Operations Manual c.
Section B, the inspector noted that procedures B.5.1.1, B.S.4, B. 5. 6, B. S. 9, B. 7. 3, D.4, and E.1.4 had not been reviewed in the 2 year interval as required by the Technical Specifications Section 6.2.B.4.h.
The inspector noted that even though these procedures had received a preliminary safety review by the Operations Committee in January 1975, five of the seven procedures were last reviewed 2 or more years ago.
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The inspector also confirmed that on March 12, 1975, an Operations Committee meeting was held during which procedures B.S.4 and B.S.6 were reviewed.
d.
The inspector reviewed the following Abnormal Procedures in Volume C of the Operations Manual and noted that they had been reviewed and approved in accordance with the technical specifications.
(1) C4-34 Service Water System Failure (2) C4-68c Loss of Neutron Monitoring (3) C4-19 Electrical System Failure
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c.
The inspector reviewed Temporary Memos 415, 419, 420, 421, 423, and 425 covering the period from August 1974 to February 1975, and noted they had not been signed and dated by the chairman of the Operations Committee as required by Operations Manual Volume A, Section A.6.
The licensee stated that action will be taken to assure compliance with Section A.6 of the Operations Mant al.
f.
The inspector reviewed the system used by the licensee for changing procedures to reflect any Technical Specification revisions and noted that members of the staff review Technical Specification revisions and initiate procedural changes as applicable.
2.
Plant Operations The inspector performed a review of general plant operations including a tour of the plant, reviews of log books, daily orders and significant operating events.
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a.
Plant Tour During a physical inspection of the overall plant with a member of the staff, the inspector noted that generally acceptable housekeeping practices exist.
Iloweve r ; two areas appeared to require additional house cleaning.
ii.e Torus area was observed to have a storage of scaffolding material and some laxity'in house cleaning was noted in the Control Rod Drive Room. The licensee stated that these areas would be cleaned up.
Review of Control room recording instruments and annunciator alarms indicated that the "SBLC Hi/Lo Temp" alarm was continuously illeminated on the annunciator panel and appeared to be the only alarm in the off normal condition.
Discussions with the control room personnel revealed that this i
alarm is tripped in the high temperature condition and hr.a been this way for some time.
The licensee stated that the apparent cause for the problem is an instrumentation problem in the temperature sensing circuit.
The licensee also stated that immeliate action will be taken to investigate and correct this problem.
b.
Log Books
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This inspector reviewed the Reactor and Control Room log and the Shift Supervisor's log for the period of December 10, 1974,
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to March 11, 1975, and confirmed that all entries are being filled out and initiated as required by Administrative Control Directive 4.7 and also that the operations Supervisor is reviewing i
and initialing the log books.
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The inspector also reviewed a selected number of entries in the Shift Supervisor's log and the Reactor and Control Room log and noted that an acceptable level of darail is being furnished.
c.
Significant Operating Events (SOE)
The inspector reviewed SOE reports 75-01 and 75-02, which
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occurred during the first 3 months of 1975.
SOE 75-01 - During refueling activities at the plant, fuel rod spacers for fuel bundle MTB-099 were slightly damaged.
The licensee reviewed this event and determined that no significant safety question was involved.
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_SOE 75-,
While conducting a reedwater Sparger inspection
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a r, pool piece from a portable light cable reel came loose and dropped into the reactor. The spool piece was retrieved and the inspection continued.
No damage to any equipment resulted f rom this event.
d.
fmily Orders The inspector reviewed the Shif t Supervf 3or's Hight Order Book for the period from January 14 1975, to March 13, 1975, and confirmed that instructions written into the book are within requirements of the Administrative Control Directives and the Technical Specificattor.s.
It was also noted that the Shift Supervisor closes out each item as it is co:npleted, J.,
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UN TED STQTES
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'l NUCLEAR REGULATORY COMfAl8SiON
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I 0FFICE OF INSPECTION AND ENFORCEMENT REGION til
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799 ROOSEVELT moAo SLEN ELLYN. 0LLit90ll 4W137
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IE Inspection Report No.
050-263/75-0s
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Transmittal Date
Apr11 21. 1975
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I Distribution
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Distribution
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IE Chief. FS&EB IE Chief FSLEB I
IE HQ (5)
IE:HQ (4)
L:D/D for Fuels & Materials DR Central Files
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DR Central Files j
, Regulatory Standards (3)
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Licensing (13)
IE Files
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L D/D for Rear" Project
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IE Inquiry Report No.
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Transmittal Date
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Distribution:
Distribution:
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IE Chief, FS&EB '
IE Chief, FS&EB
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i IE:HQ (5)
IE HQ DR Central Files l
DR Central Files i
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Regulatory Standards (3)
IE Files
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Licensing (13)
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IE Files
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C.
?ncident Notification From:
(Licensee 6 Docket No. (or License No.)
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Transmittal Date
i Distribution:
Distribution:
f IE Chief, FS&EB 18 Chief, FS6EB
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IE:HQ (4)
IE HQ (4)
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L D/D for Fucis & Materials l
Licensing (4)
DR Central Files DR Central Files
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$
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