IR 05000029/1974002
| ML19339A941 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 04/17/1974 |
| From: | Davis A, Hannon J, Oberg C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19339A939 | List: |
| References | |
| 50-029-74-02, 50-29-74-2, NUDOCS 8011050731 | |
| Download: ML19339A941 (23) | |
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U.S. ATOMIC ENERGY CO:: MISSION
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DIRECTORATE OF REGULATORY OPERATIONS
REGION I
Docket No:
50-29 RO Inspection Report No:
50-29/74-02
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License No: DPR-5
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Licensee:
Yankee Atocic Elactric Company
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Priority:
20 Turnpike Road s
i Westboro, Massachusetts 01581 Category:
C Rowe, Massachusetts
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Location:
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Ty'e of Licensee:
PWR, 600 T.Jt (W)
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Routine, Announ:ed Type of Inspection:
March 11-15. 1974 Dates of Inspection:
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December 12-14., 1973
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Dates of Previous Inspection:
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Reporting Inspector.:
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C.R.Oer,Princi[a1 Inspector
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M4t. 4 L.V f
Accompanying Inspectors:
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//J. N. Hannon, Reactor Inspector
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Other Accompanying Personnel;
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Uate Revicwed By:
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l A. B. Dav'is, ' Senior Reactor Inspector (PWR)
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SU:0!ARY OF FINDISCS
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Enforcement Actiens A.
Primary System Leak Rate Determination
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The procedure for determining primary system leak rates was not followed.
(Paragraph 5.a.)
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t, Radiation Area Boundaries Not Properly Established,
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i Boundaries of a contaminated area in the New Fuel Storage Area were not
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properly identified and marked on all sides.
(Paragraph 16.b.)
Licensee Action on Previously Identified Enforcement Actions A. -Control Rod Receipt Inspection
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The corrective action of the licensed was verified.
(Paragraph 6a)
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B.
Diesel Generator Surveillance Test The corrective action (revising surveillance test) of the licensee was verified.
(Paragraph 14.b)
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Design Changes
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Pres'surizer. Relief Valve Discharge Piping Restraints The licensee is planning to install additional restraints on the two 4 inch discharge lines.
(Paragraph 5.f)
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Unusual Occurrences Uncontrolled Release of Caseous Activity A valve bonnet joint leak in the waste gas system caused an accidental i
release of gaseous radioactivity.
Release limits were not exceeded.
(Paragraph 17)
Other Significant Findings A.
Unresolved Items
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None i
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Current Findings
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1.
Refueling is scheduled to start on May 10, 1974.
(Paragraph A, Managenent Interview)
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2.
The licensee continues to revise plant operating procedures.
(Paragraph 4)
3.
The licensee will up date and correct their refueling procedures by May 1, 1974.
(Paragraph 15)
Management Interview Inspection findings concerning the following subjects were discussed with
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Messrs. Autio, Jones, and St. Laurett at the conclusion of the inspection on March 15, 1974:
A.
Refueling The inspector stated that he understood that the Core X refueling is scheduled to start May 10,' 1974 and requested that Region I be notified
,1f a change in schedule occurs.
The licensee stated that RO:I would be kept informed of any changes to
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the schedule.
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B.
Facility Procedure Review The overall status of the Facility Procedure Program was discussed.
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l Specific concerns addressed included (1) the lack of procedural cov-J erage in certain normal and emergency operations, (2) the omission of references to appropriate Technical Specifications and applicable associated procedures, (3) the absence of precise limits and specific details which could be prescribed for the operator, and (4) the ob-servation that some procedures do not receive a review commensurate with their level of importance.
The licensee stated that completion of the rewrite effort could be expected in 1975.
C.
Refueline Procedures
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The status of the facility refueling procedures was discussed with the licensee. A specific problem was addressed regarding accountability of tools and miscellaneous equipment during the refueling.
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The licensee stated he would review this area and that all refueling procedures would be revised and reviewed by May 1, 1974.
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D.
Primarv System teak Rate Determination The inspector stated that an apparent violation exists wherein the procedure for determining the primary system leak rate has not been followed in several instances. Specific exa=ples were discussed.
The licensee agreed to recalculate the leak rates beginning approximately February 1, 1974.
In addition the procedure will be updated and training J
of personnel responsible for the calculations will be done in order to make the results censistant.
E.
Review of Primary Systems / Power Conversion Systems
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The inspector stated that one of the objectives of the inspection was to review selected areas of the Primary Coolant and Power Conversion systems.
Specific areas examined are documented in the repoct.
F.
Tour of Facility The inspector stated his observations made during a tour of the Facility.
An apparent violation was observed wherein there was a failure to ad-equately mark boundries of a contaminated area in the new fuel storage
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room.
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Other observations concerning cleanliness of containment areas and lack of protection for electrical and instrumentation terminal connections inside containment were discussed.
The licensee stated that the containment would be examined by the plant superintenden*. He further stated that the doors were left off the
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terminal boxes to permit ready access for repair and to eliminate moist-ure condensation that had been forming inside the boxes.
G.
Posting of Operating Procedures for Personnel Hatch The inspector stated that instructions for the operation of the con-tainment personnel hatch door controls were not posted as stated by the licensee in Yankee letter dated April 11, 1973.
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The licensee stated.that operating instructions would be posted.
H.
Missing Information from Semi-Annual Report The inspector requested that information missing from the Semiannual
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Information missing:
(a) December radioactivity gaseous effluent levels.
(b) Reactor Coolant System leak rate ir. formation on 1st page of Su=ary of Primary & Secondary Chemistry for each conth.
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The licensee stated that the information would be forwarded to Region I.
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DETAILS 1.
Persons Contacted
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j Mr. H. Autio, Plant Superintendent
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Mr. W.. Billings, Chemistry & Health Physics Supervisor Mr. T. Danek, Operations Supervisor Mr. M. Ebert, Reactor Engineer Mr. J. Flanigan, Plant Health Physicist Mr. R. Her:og, Shif t Supervisor Mr. W. Jones, Assistant Plant Superintendent i
Mr. B. Kirk, Shif t Supervisor
Mr. P. Laird, Maintenance Supervisor -
Mr. N. St. Laurent, Tachnical Assistant to the Superintendent Mr. R. Paradis, Control Room Operator Mr. E. Pierce, Control Room Operator
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Mr. D. Vassar, Shift Supervisor
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2.
Operations
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a.
Plant Operations
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At the onset of the inspection Yankee Rowe was operating at approx-imately 75% reactor power in a 'oastdown condition. All rods were c
out at 90-93 inches, Tm - '.80.2 F, 135.1 MWe output, with boron <
15 PPM. All systems aproared to be operating satisfactorily.
b.
Tour of Facility
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Th( inspector made a.cour of the facility, including a visit inside
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containment. The inspector found an apparent need for housekeeping
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improvement inside containment in that cleanliness was poor and tools
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and spare materials were scattered about various areas. ; In addition,
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doors to several large electrical and instrumentation connection
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i terminal boxes were off which lef t the terminals subject to moist-i i
ure, du'st, dirt and physical damage, i
The licensee stated that the doors had been lef t off to eliminate
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j possibility of any moisture condensing inside the boxes. He also
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i stated that this condition would be evaluated in order to determine
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if the need still exists. These items were discussed during the
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Exit Interview and will be' examined during the next inspection.
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. Review of' Semiannual Operations Report and other Plant Records
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The inspector reviewed the Yankee Rowe Semiannual Operations Report
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as part of.the inspection. Missing report information was identified
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to the licensee. Discussions on several of the items took place with the Plant Superintendent and other supervisory personnel.
Results of the discussion are documented in the appropriate section of this report.
l The inspector also held discussions relating to any safety items, violations, excessive personnel exposures, and excessive releases i
of radicactive effluent. Results of the discussions are documented in the appropriate section of this report. No safety items were c
identified by che licensee. The inspector also reviewed the Shift Supervisor's Log for the period August 25, 1973 to November 2, 1973 and December 2, 1973 to March 11, 1974. No abnormalities were ident-ified.
In addition, minutes of the Nuclear Safety Review and Audit Committee (NSRAC) were reviawed covering the two meetings held since the last inspection. Mitutes for the Piant Operations Review Com-mittees (PORC) were also reviewed which covered meeting Nos. 74-1 through 74-11.
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The inspector also held d!scussions with operations personnel at various times durir.g the inspection.
d.
Review of Scram Records
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The inspector examined. plant process records and documentation re-lating to a reactor s' cram which occurred on August 31, 1973. The following charts were examined:
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(1)' Process Records - records of 16 plant parameters including
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temperature, flows and levels.
(2) Incore Thermocouples.
(3) Steam Generator Levels.
(4) Pressurizer Pressure.
D (5) Total Steam and Feed '.fater Flow and Pressure.
(6) Individual Steam and Feed Water Flow.
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(7) Steam Generator Levels (Narrow & wide range indications).
"(8) Flux Channels (No. 1-4)
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No abnormalities were noted in the records examined.
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3.
Administration and Otranization
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a.
Shift Manning The inspector determined that the requirement of Technical Specifi-
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cation,. Change 109, concerning minimum shift manning is being met i
by the licensee.
This item is closed.
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Nuclear Safety Audit and Review Committee (NSARC)'
The licensee reported the following composition of the NSARC, ef-fective March 1, 1974:
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D. B. Pike (Chairnan)
E. G. Wood J. S. Shulman J. W. Stacey (Vice President)
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J. W. Singleton D. W. Riley J. DeVincentis A. E. Ladieu R. H. Graves
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Personnel Chances The following personnel assignment were made since the last in-spection:
Assistant Plant Superintendent - W. G. Jones Reactor Engineer - M. W. Ebert I 5 C Supervisor ' J. H. Shippee
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Chehistry & HP Supervisor - W. D. Billings Technical Assistant to Plant Superintendent - N. St. Laurent 4.
Facility Procedures a.
Procedure Status The overall status of the Facility Procedures Program was reviewed *
against the requirements of the Technical Specifications, Regulatory Guide 1.33 (November 3, 1972) and ANSI-N18.7 (November 12, 1973).
During the review of selected sample procedures, stress was placed on (1). emergency procedures, (2) plant operating procedures, and (3) refueling procedures, with the latter group receiving =ajor emphasis.
The following su=marizes the overall status of the pro-cedure upgrading program as determined during the inspection:
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Procedure
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% In Review
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Not Written Administrative
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Emergency
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Alarm
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ceneral
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12% in old for :
Systems
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47% in old fort:
Radioactivity
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68% in old form:
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Surveillance
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Measuring Equip.
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11% in old for=2 Test Equipment
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Maintenance
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10% in old for=n
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e Radeon (EP)
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Most are covered in the Plant Radeon Manual.
s The licensee has established a target date for the ccmpletion of the program of March 1, 1975. The licensee committed to treat the inspector's co=ments on a generic basis for upgrading the entire program.
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b.
Em?rcenev Procedures The licensee agreed to Laprove the emergency procedures based on the following com=ents by March 1, 1975.
(1) Battery No. 1, 2, or 3 Critical Voltage OP-3754.
(a) According to procedure, higher authority is not required to be notified of this casualty.
(b) Applicable procedures are not referer. cad.
(c) Technical Specification limits are not defined.
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(2) Battery Charger No. 3 1.C. Failure OP-3747. Although there are no 9toblems for the other chargers, reportedly this procedure will be invnked upon the failure of either No. 1 or 2 Battery Chargers.
(3) Total Loss of A. C. - Control Room ' Secondary Plant Operator Cuide OP-3252.
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(a) Symptoms of this casualty are not listed.
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(b) I= mediate actions are not clearly defined.
(c) Manual actions are not specified in the event automatic actions fail to occur.
(d) The reactor shut down is not addressed.
(e) Apo11 cable procedures are not referenced.
(4) Feed Wacer Line Break Emergency Action OP-3203.
(a) Immediate actions and followup actions. both automatic and manual, are not clearly detailed.
(b) Step 7 lists a valve not identified by number.
(5) Total Loss of Main Coolant 0,,P-3106.
(a) In step III.4, figure 1 is mentioned with no location or reference.
. (b) It is not clear from step' I'I'I.7 how the operator will de-~ ~ ~ ~ ' *
termine minimum injection.
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-10-(6) Total Loss of Main Coolant Flow Op-3103.
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(a) Electrical check-offs are not available for use in ocep III.10.
(b) Applicable procedures are not referencad to restore the plant to normal. OP-2501, Restoration of Normal A.C. Po-wer after a Total Loss of A.C. is in the revision process.
(7) It was noted that no procedure for emergency evacuation of the control roca was available. Discussions with operating personnel indicate that it may be feasible to maintain hot standby remote from the control room, although a re=ote cooldown may not be poscible. The licensee stated that this area would be examined for procedural coverage.
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Operating Procedures
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The licensee agreed to upgrade th'e operating procedures based on the following comments by March 1, 1975.
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(1) Operation of the Control and Service Air Systc=s RP-2600.
(a) Consideration should be given to upgrading this procedure to an OP, based on its relevancy to plant safety.
(b) Prerequisites do not identify plant conditions that must be
met prior to equipment operation.
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(c) There is no reference to applicable Technical Specifications.
(d) Periodicity ci equipment checks is not specified in pre-caution No. 2.
(e) Control roca alarms are not verified. dpan shut down.
(2) Feed Water Line Isolation & Return to Service. RP-2250.
(a) The acceptance criteria for veld inspections are not specified.
(b). Applicable procedure: are not referenced.
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(3) Reactor and Primary Plant Cooldown.
(a) Step 111.1 does not describe the specific Technical Specifi-cation require =ents.
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(b) The Minimum Pressurization Temperature (!CPT) Curve is not referenced.
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-11-(c) Special manning requirements are not addressed.
(4) Primary Plant Startup from Cold Condition, OP-2100.
(a) The electrical pre-startup check-off line is not ref erenced in step II.6.
(b) The !TT curve is not referenced.
(c) Applicable Technical Specifications are not referenced.
(d) The applicable procedure was not referenced in step III.7.
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(e) No provisions are made fo'r heat up if decay heat is neg'ig-
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ible.
(5) Changing Plant Load OP-2107.
(a) Statements such as "all systems" do net provide specific guidance for the operator.
(b) Applicable Technical Specifications are not referenced.
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(c) Check off by initials or signature is not provided for.
(6) It was noted that no specific procedures were in existence for the operation of the Steas Generator Slow Down Systes or the Steam Dump Valve System.
d.
Surveillance and 'Sintenance Procedures.
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The licensee agreed to upgrade the Surveillance and :taintenance Procedures based on the following con =ents by March 1, 1975.
(1) Monthly Test of Safety Injection System OP-4204.
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(a) Technical Specifications are not referenced and limiting conditions for operation are not specified.
(b) Acceptance criteria for satisfactory completion of the test are not specific.
(2) Flew Test of Two ECCS Trains with E=er3ency Power, OP-4209.
No comacnt.
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(3) Emergency Boiler Feed Pu=p Surveillance Test, OP-4211.
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No correlation of previous test data is provided for by this or other similar procedure _
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-12-(4) Maintenance Department Surveillance Schedule, AP-5000.
(a) Technical Specification 4.5 in step 2.e should be 16.5.5 per the proposed Technical Specifications.
(b) AP-0214, Installation and Maintenance of Safety Classified Systems, Components, or structures, is not referenced.
e.
Instrumentation Procedures The licensee agreed to upgrade the procedures applicable to instru-mentation based on the following co=nents by March 1,1975:
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(1) Reactor and Turbine /Ger.erator' Permissive Switch and c.ssociated
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Ti=e Delay Relay Calibration and Functional Test, OP-6103.
(a) The test frequency is not specified by procedure.
(b) Electrical safety rules are not referenced in step 2 under
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Precautions..
l (c) Jumper accountability is presently controlled by the Lif ted
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Lead Log. The licensee stated that a new procedure, AP-0018,
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Lifted Lead, Jumper Control & Accountability, will be writ-
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(2) Inspection and Stroke Calibration of V. C. Trip Valve No.
OP-6450.
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(a) Statements such as " observe general safety precautions" are not sufficiently specific.
(b) Frequency of the test is not specified.
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. (c) Remova' of the valve from service is not addressed, includ-l
. ing r.ty limits on plant operations that may be imposed.
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Health Physics Procedures.
The licensee agreed to upgrade the Health Physics procedures based on the following comments by March 1, 1975.
(1) Establishing.and Posting Control]-d Areas, OP-8100.
(a) This procedure (D7-8100)-was not available (except fo r the --.
first page) at the HP control point.
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(b) HP personnel were not familiar with specific limits regard-i ing requirer. cats for shoc cover..
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_13 (c) One procedure with-this title was numbered OP-3105, while the master was numbered OP-8100.
(2) Use of Protective Clothing RP-8400.
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Condition for the use of protective clothing are not detailed as to the requirements for each specific article.
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Ad-inistratire Procedures.
The licensee agreed to improve the administrative procedures based
.on the following comments by March 1, 1975 unless othervise indicated.
(1) Plant Procedures, AP-0001.
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(a) Operator adherence to written procedures is not specified.
(b) Provisions for temporary changes are not specific regarding impact upon procedural intent.
(c) Low Power Physics Tests and Power Ascension Tests reportedly will be prepared prior to refueling.
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~(d) Safety related operating memos are not addressed and do not receive proper managerial review.
5.
Primary Systems a.
Determination of Primary System Leak Rate Yankee Operating Procedure No. OP-4220, dated January 31, 1974 re-quires a water inventory. determination be made three times a day.
Once a day an overall water inventory change is made for the pre-vious 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The purpose is to monitor the primary system leak-age.
The inspector exaeined calculations for several days and found in-consist 'ncies in the way various shif t supervisors determined Leak Rates.
- rocedure was not followed in that a change in the Primary Drain 2g tank was not accuractely reflected in the calcula-tions on Maren 2'& 3, 1974.(0900). Other examples were brought to the attention of the licensee. This is an apparent violation of OP-4220.
The licensee stated that water balances would be recalculated, the procedure would be revised-to clean ambiguities, and training would__..
be conducted for those who are responsible for-the calculations.
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This ites was discussed in the manage =ent interview and is considered
open.
b.
Primarv Svsten Check Valves The inspector held discussions with the licensee and exanined re-cords pertaining to the primary system check valves. A history of j
worn bushings and pins was evident. The valves become noisy under constant flow conditions and are =onitored by listening to the valve.
Repairs have been made in the past:
November,1972 - Replaced disc, ar=, pin and blocks of loop No.2.
October, 1972 - Replaced internal assembly, disc, arm, blocks, pins,
space washers. The RH (facing closed disc) block was almost completely worn through so that cost of
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the pin was exposed. The disc was-unable to close closer than one inch and hanging crooked. The ar:
was worn sufficiently to allow the discs to drop.
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l The licensee has a design change approved and is in the process of obtaining proper materials to modify the valves during the Core XI
refueling. At the present ti=e all check valves are operating satisfactorily.
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c.
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The inspector held discussions with the licensee and axamined approp-riate records relating to the performances of the steam generators.
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The steam generators have performed satisf actorily since the codifica-tion to the feedwater lines in April,1963. Prior to that ti=e severe feedwater hammer was experienced during startup operations.
The modification involved the installation of a loop seal just be-fore the pipe enters the steam generator.
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Some steam generator tubes have been plugged. Records of the licensee indicate the following:
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Steam Generator Number of Tubes Pluczed No'. 1
No. 2
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No. 4
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Tiere are 1620 "U" type stainless steel tubes in each steam genera-
.There is some evidence that there may be steam generator tube 2.a r.
fouling (Paragraph 19). The licensee is also monitoring a possible leak in No. 4 steam generator..At.presen4 calculations indicate~ ~ ~ ' *
that the leak is less than 2 gal / day.
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An insurance company inspection report of February, 1972 stated that
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No. 2 & No. 4 Steam generator did not have any abnor=alities.
Radiation levels were 200-400 mr/hr in No. 2 and 50-150 mr/hr in No. 4.
d.
Main Coolant Stop Valves The inspector examined records and held discussions with the licensee concerning the Main Coolant Stop Valves.
These valves are =otor
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operated and are required to open or close in 120 sec. + 10%. The latest exercise of the valves indicated the following:
MOV NO Close (min-see)
Open (min-see)
301 1-58.5 1-58.1 302 2-3.2 2-1. 9 309 1-59.8 1-59.8 310 1-58.4 1-57.9
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318 1-59.1 1-58.7
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319 1-59.4 1-59.5 325 2-3.1 2-3.2
326 1-59.0 1-58.5
The operators of the valves were last inspected in February of 1972.
MOV 325 was found to contain hardened grease. This was replaced.
All operations will be inspected again during the 1974 refueling.
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e.
Other Primary System Valves The inspector examined reports and held discussions with the licensee concerning primary system valves.
Records indicate that 13 valves have various amounts of leakage.
The majority of these valves have been capped.
The licensee conducts visual inspections of the primary system (in-
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si'de containment) whenever the reactor is shutdown and hot. Each valve is inspected and results are documented.
f.
Pressurizer Safety and Relief Discharge Piping The inspector reviewed a report concerning the Stress Analysis for Pressurizer Saf ety and Relief Valve Discharge Piping.
The pressurizer is provided with two safety valves and one power-operated relief valve. The safety valves discharge into four inch lines and the relief valve into a three inch line. The three dis-charge lines merge into a common leader which connects to a relief tank.
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-16-Two methods of analysis were used and both revealed that a stress problem exists in the discharge piping, and in particular the two four-inch lines. Additional restraints will be required.
The licensee plans to install the additional restraints during the refueling shutdown starting in lSy, 1974.
t This item remains open.
6.
Reactivity and Power Control
a.
Control Rod Receipt Inspection
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References:
RO Inspection Report 50-029/73-04
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Yankee letter to 20:1 dated November 6,1973
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Ihe ccrrective action s tated in the letter of Nove=ber 6,1973
was vet!fied by the inspector.
i b.
Malfunction af Safety Related Westinghouse W-2 Switches The inspector determined that the Yankee Rowe facility does not utilize W-2 switches. They had not received the NSD technical
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Bulletin (NSD-T3-73-26, dated December 12, 1973).
7.
Core and Internals
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Not inspected.
8.
Power Conversion Svstem a.
Steam Turbine Evpass System
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The inspector examined records and held discussion with appropriate licensee personnel concerning the Steam Turbine Bypass System.
The system provides capability to remove steam during startup and shutdown by using a flew contro' valve in either automatic or re-mote manual operation. The 6 inch line passes sufficient steam for approximately 5% reactor load, exhausting into the =ain condenser.
No problets have been experienced with the operation of this sys-tem.
b.
Steam Generator Blowdown Svs tem
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The inspector reviewed available records and held discussions with appropriate licensee personnel cotterning the Steam Cencrator Blowdown Sys, tem.
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Only one problem exists in this system. A pin hole leak in the blowdown line for No. 1 Stems Generator will be repaired during the 1974 refueling period. The leak has a te=porary patch on at the present time.
9.
Auxiliarv Svster.s
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Not Inspected.
10. E1cetrical Systems DB-50 Reactor Trip Breakers
'The inspector determined that the Yankee Rowe Facility does not utilize DB-50 reactor trip breakers, and the require:ent of NSD-TB-74-1 does
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not apply.
11. Containment
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Yankee letter to Licensing, dated December 12, 1973 References:
Licensing letter to Yankee, dated January 14, 1974
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The inspector examined the records relating to monitoring containment
, integrity by continuo,us leak monitoring. The examination revealed that the licensee is utilizing the system and that the integrity of the con-tainment is intact.
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This item is considered closed.
12. Emeraenev Core Cooling System
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Not inspected.
13. Other Eneineered Safety Features Not inspected.
14. Emergenev Pcwer DC Control Circuits Blown on No. 3 Emercency Diesel Generator a.
( AO-7 3-05)
Reference:
RO Inspection Report No.'50-29/73-04
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Modification to the control circuit has been completed.
This item is considered closed.
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b.
Diesel Generator Surveillance Test References: R0 Inspection Report No. 50-029/73-04 RO:I letter to Yankee, dated October 19, 1973 Yankee letter to RO:I, dated November 6, 1973 The insnector confirmed that the licensee action was as stated in
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their letter of Nove=ber 6,1973.
This. item is considered closed.
15. Fuel Storace and Handling Refueline crocedures The licensee agreed to upgrade the refueling procedures based on the
following coenents by !by 1,1974, except where otherwise indicated:
(1) Inspection of Fuel Handling Equipment (OP-4505)
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(a) Sources of control power are not specified.
(b) Jumper installati6n in step 40 lacks appropriate controls and accountability.
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(2) Testing of Fuel Handling Equipment (OP-4226)
Interlocks are not clearly defined or referenced.
(3)~ Refueling Accidents (OP-3117).
(a) Inspection and evaluation requirements for new fuel damaged during refueling are not addressed.
(b) The chemical shut down procedure, OP-3107, was not referenced in step I.2.
(c) Appropriate Health Physics procedures are not referenced in
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step III.3.
(d) Step 11.4 does not make provisions for clearing the alars con-dition.
(4) Reactor Fuel Loading, Component Replacenent, OP-1200.
(a) This procedure was due for annual review on March 10, 1974.
The licensee statec that it would be reviewed and revised where necessary at a Plant OFerating Review Committee (PORC) =ecting ~~ *
during the month of March.
If there are no revisions to a pro-cedure, the PORC meeting minutes reportedly will provide doc-
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_ 19 umentation of conpliance with the periodic inspection require-ments.
(b) Calibration and response checks of incore and excore detectors nd a prescribed 9eriodicity for same has not been addressed.
It was noted that RP-1602, Refueling Nuclear Channel Operation, was in the review process.
(c) Incore flux monitor requirements have not been addressed. The licensee stated that the vendor for the movable incore system (to be installed durir.g the refueling outage) would provide the technical interface for the sys tem, including how it relates to re.aeling.
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(d) Prequisite number 1 does not'specify exactly what is to be tagged I
out of service. The licensee stated that 0?-1202, Locked Valve Checklist for Refueling, would cover this area.
i (e) Verification of containment integrity including closing and tagging valves and confirming operability of valves and hatches was questioned..The licensee stated that his interest was only to control the large openings in the lower portion of contain-
ment and not to provide an air tight boundary, and that the pro-
, cedure would be revised accordingly.
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(f) The spent fuel cooling system was not addressed. It was noted that RP-2164, Placing the. rent Fuel Pit Cooling and Purifica-tion System in Service, was being prepared.
(g) Installation of neutron startup sources was not addressed. The licensee stated that OP-1000.5, Master Refueling Procedure Core X-XI, and 0?-1209, Operation of the VC Manipulator, will cover verification and installation of the sources.
(h) Radiation protection require =ents are not addressed. CP-4812, Calibration Check of the Ca==a Guards, is in preparation.
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(i) The status of all systems required for fuel loading, including operability and lineup, was not specified.
(j) The operability of the VC purge system is not confirmed by pro-cedure prior to refueling operations.
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(k) Inspection of fuel within a specified time prior to fuel load-ing is apparently not prescribed, although OP-7200, New Fuel Inspection, is in the review'prdcess.
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(1) The operability of fuel handling crancs, equipment, and tools does not appear to be verified by procedure within a specified
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tine prior to fuel loading.
It was noted that DP-5951, Vapor Container Crane-Inspection and Maintenance, was in preparation.
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(m) The status of protective syste=s, including special trip func-tions, does not appear to be verified by procedure. RP-1600, Refueling Level & Pressure Detector Protection Alarm, covers this area, although it may not be all inclussive.
(n) Minimum requirements for maintaining coolant circulation in the core are not specified.
RP-2162, shutdown Cooling Syste= Start-up, and RP-2163, Shutdown Cooling System Removal from Service, reportedly will include these requirements.
(o) Limits on the water level in the fuel pool are not prescribed.
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OP-1203, Filling of Shield T'ank Cavity, and OP-1214, Trans fer of New Fuel from the New Fuel Vault to the Spent Fuel Pit, re-portedly will provide limits on the fuel pool water level.
(p) The Operating License lists 1400F as the reactor coolant sys-tem te=perature li=itation. Procedure 504MA lists 130-1500F as the limit.
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(q) Control and accountability of tools, eyeglasses, flashlights,
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, rags, paper, tape, supplies and other like items during refuel-ing is not defined.
The licensee stated that he would take this item under consideration ahd =ake a decision prior to refueling.
(r) Limitations cn fuel loading in the event of a com=unications
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failure are not addressed.
(s) The analysis, frequency, and acceptance criteria for sampling borated water during refueling are presently not covered by pro-cedure.
(t) OP-3105, Emergency Boron Injection, is not referenced.
(u) Dual independent verifications of each fuel assembly serial nu=-
be'r and core position are not required by procedure prior to insertion.
(v) The procedure does not presently require two persons to be pre-sent at any location where fuel handling is taking place.
(w) Low count rates are not addressed. The licensee stated that a procedure would be prepared for use during this event if it occurs.
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d (x) Appropriate Health Physics procedures which control shift ex -
-posures are not referenced.
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-21-16. Radiation Protection a.
Loop Seal Monitors i
The inspector questioned the nine failures of low pressure loop
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seal monitors reported in the facilities semi-annual report. The licensee reported that the water condenses on the inside of the GM tube window shorting out tube. The licensee is obtaining a different type to replace one presently used.
b.
Radiation Area Boundaries Not Properly Established.
During the tour of the facility, the inspector noted that radiation control procedures were not being carried out correctly. An access torad{ationarea(conta=inated)wasnotproperlycarked(470dpa/
100 cm ).
This'is an apparent violation of procedure OPS 100, "Es-tablishing and Posting Controlled Areas," dated June 13, 1973. This procedure states, in part, "All c'entrolled areas shall te adequately surveyed to insure that the boundaries established conpletely define the controlled area and that no loose surface contamination, airborne activity, or excessive radiation exists i==ediately outside the local boundary."
(Underlining provided)
The rope boundary did.not conpletely enclose the contaminated area.
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This item was discussed in the canagement interview and is consider-cd epen.
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c.
10 CFR 19 Inspection The inspector verified posting by the licensee of notices required by 10 CFR 19.
Th'is item is closed.
17. Radioactive Waste Systems Unplanned Release of Radioactive Material (A074-1)
References; Yankee letter to Licensing dated March 8, 1974 The inspector verified the action taken to preclude a repetition of the valve leak.
This item is considered closed.
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18. Experiments and Tests Core Flow Test
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The inspector reviewed documents and discussed with appropriate licensee personnel the results of a test conducted to determine if any significant changes had taken place in reactor core flow. The Semi-annual report indicated "---a noticeable decrease in core flow from Core VIII to Core X.
This result was found to be due to Core X fuel."
Subsequent analysis of the data by Westboro Engineering Staff indicated
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that (1) Core X fuel (Zircoloy Asse=blies) have caused a reduction in flow at less than 1%, (2) Precise cause of the increase in core and loop LT's has not been determined (3) An increase in average main cool-ant temperature is cost likely caused by crud deposits in the steam gen-erators.
The licensee indicated that this catter would be followed and that, for
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safety and transient anaylsis cf. cord XI, the loop AT would be in-creased from 42 F to 44 F.
This item remains open.
18. Ifiscellaneous
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, Facility tour.
During a tour of the " Waste Dispdsal Building, it was noticed that the hydrogen =eter was incorrectly marked "0xygen."
The licensee stated that the nameplate would be corrected.
It was noted that the system contained approximately 45% hydrogen which is normal' for the facility.
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