IR 05000270/1974004
| ML19308A606 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 07/15/1974 |
| From: | Jape F, Lewis F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19308A603 | List: |
| References | |
| 50-270-74-04, 50-270-74-4, NUDOCS 7911270598 | |
| Download: ML19308A606 (19) | |
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, j 1D.'g UNITED STATES
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DIRECTORATS OF REGUIATCRY OPERATIONS
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RO Inspection Report No. 50-270/74-4 Licensee:
Duke Power Company Power Building 422 South Church Street Charlotte, North Carolina 28242 Facility Name:
Oconee Unit 2 Docket No.:
50-270 License No.:
DPR 47 Category:
B2 Location:
Seneca, South Carolina Type of License:
Type of Inspection:
Routine, Unannounced Dates of Inspection: May 16-17, and June 4-7, 1974
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Dates of Previous Inspection: April 9-12, 1974
Principal Inspector:
F. Jape, Reactor Inspector Facilities Test and Startup Branch Acconpanying Inspectors:
K. W. Whitt, Reactor Inspector Facilities Test and Startup _ Branch G. R. Jenkins, Radiation Specialist Facilities Test and Startup Branch Principal Inspector:
A (48'O-ed SC.
7 ~d" W F. Jape, Reactor Inspector'
Date Facilities Test and Startup Branch
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/d [h Reviewed By:
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~ Da t.e R. C. Lewis, Senior Reactor Inspector Facilities Test and.Startup Branch
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RO Rpt. No. 50-270/74-4-2-
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SUMMARY OF FINDINGS I.
Enforcement Action A.
Violations 1.
The following violations are considered to be Category 11 severity:
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a.
Modifications Implemented Without Review by Nuclent Safety Review Committee r
Technical Specification 6.1.2.2.i(2) requires the
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Nuclear Safety Review Commi*. tee, NSRC, to review and audit proposed changes that may involve an unreviewed
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safety question.
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Contrary to the above, modifications are being implemented without NSRC review and audit. (Details I, paragraph 2)
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b.
Reactor Coolant System Cooldown Limitation The reactor coolant system cooldown limitation, specified in Technical Specification 3.1.3, was exceeded on January 4, 1974.
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Duke Power Company reported the above violation in Abnormal Occurrence Report No. A0 270/74-1, dated January 17, 1974.
(Details III, paragraph 4)
c.
Failure to Verify Containment Integrity Prior to Startup Technical Specification 4.4.1.2.S(b) requires leak detection tests be performed of the personnel hatch every four months, unless che hatch has not been opened.
- Contrary to the above, the containment personnel hatch was not leak. tested during the period of October 29, 1973, and May 6, 1974, as required.
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RO Rpt. No. 50-270/76 -4-3-
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Duke Power Company reported the above violation in Abnormal Occurrence Report No. A0 270/74-3, dated May 15, 1974.
(Details I, paragraph 10)
d.
Overpressurization of Core Flood Tank 2A
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Pressure in core flood tank 2A exceeded Technical Specification 3.3.3 limit by 5 psi on May 23, 1974.
Duke Power Cc=pany reported this violation in Abnormal Occurrence Report No. AO 270/74-5, dated June 3, 1974.
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e.
Underpressurization of Core Flood Tank 2B Pressure in core flood tank 2B dropped 7 psi below Technical Specification limit 3.3.3 on May 25, 1974.
Duke Power Company reported this violation in Abnormal
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Occurre.nce Report No. A0 270/74-6, dated June 4, 1974.
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Removal of Both Reactor Building Spray Loops From i
Service Contrary to Technical Specification 3.3.la, both reactor building spray loops were removed from
service on January 22, 1974.
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i Duke Power Company reported the above violation in Unusual Event Report No. UE 270/74-1, dated February 20, 1974 (Details III, paragraph 5)
2.
The following violations are considered to be Category III severity:
a.
Transfer of Byproduct Material Contrary to 10 CFR 30.41, a metal seal ring containing byproduct material was transferred to a company in the State of Oregon on February 21, 1974, who was not authorized to receive the material.
(Details II, paragraph 2;)
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RO Rpt. No. 50-270/74-4.
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Activity in the Component Cooling System Contrary to Technical Specification 6.6.2.1 A an abnormal occurrence report concerning radioactivity in the component cooling system was not submitted,
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as required.
(Details I, paragraph 3)
B.
Safety Items None II.
Licensee Action on Previously Identified Enforcement Matters A.
Violations 1.
RO Inspection Report No. 50-270/74-1 The corrective actions stated in Duke Power Company's response, dated March 21, 1974, to violations identified in R0 Report No. 50-270/74-1 have been inspected and there are no further questians.
(Detcils II, paragraphs
'3 and 4, and Details III, paragraph 2)
2.
RO Inspection Report No. 50-270/74-2 Duke Power Company's response, dated May 24, 1974, has been received and is currently being evaluated.
Followup
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is scheduled for a future inspection.
B.
Safety Items None
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III. New Unresolved Items 74-4/1 Overpressurization of CF Tank 2A Item remains open, pending a followup inspection of the implementation of corrective measures described in DPC's report.A0 270/74-5 dated June 3, 1974.
74-4/2 Underpressurization of CF Tank 2B Item. remains open, pending a follevup inspection of the implementation of corrective measures described in DPC's report A0 270/74-6' dated June 4,11974.
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R0 Rpt..No. 50-270/74-4-5-
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74-4/3 Vital Bus Inverter Shunt Trip
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Removal of the Shunt trip on the vital AC bus static inverters has not been completed.
(Details I, paragraph 7)
IV.
Status of Previously Reported Unresolved Items 74-3/1 Temporary Station Modifications
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The procedure for proce. sing temporary station modifications is currently being prepared.
74-3/2 HP-31 Reset Interlock Licensee is reviewing this item to determine if it should be classed as safety related.
74-3/3 High Energy - Line Break Modifications Work on this modification has been completed.
Item is closed. (Details I, paragraph 5)
74-1/1 Stations Review Committee The licensee's response to the unresolved item concerning the Station Review Committee was reviewed and there are no questions on this item.
Item is closed.
(Details, I,
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paragraph 4)
73-1/2 Communications of Activities Affecting Reactor Operations Measures have been taken to inform shift personnel of activities affecting reactor operations.
Iten is closed.
(Details III, paragraph 3)
74-1/3 Records of Radioanalysis Results The licensee now records the actual value of ninimum detectable activity where analysis results are less than that value, and records statistical uncertainties for low activity and other selected analyses.
Item is closed.
(Details II, paragraph 5)
73-8/2 Valve Wall Thickness of Valve 2-RV-67-
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Information received from DPC, dated ihy 6,1974, vu.
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reviewed and previous questions concerning 2-RV-67 have
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been resolved.
Item is closed. (Details I, paragraph 6)
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R0 Rpt. No. 50-270/74-4-6-
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73-8/l Body 'Jall Thickness of Valves 2-51-244 and 2-51-245 Revision 4 to the Oconee Valve k'all Thickness Verification Report, dated June 3, 1974, has been received and is currently being reviewed by RO:II.
V.
Unusual Occurrences 1.
A0 270/74-1, " Reactor Coolant System Cooldown Limitations"
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The reactor coolant system cooldown limitation, as specified by Technical Specification 3.1.3, was exceeded.
The Abnormal Occurrence Report, dated January 17, 1974, was reviewed during this inspection.
Item is closed.
(Details III, paragraph 4)
2.
A0 270/74-2A, " Seal Leak on RCP 2B2" Addf*ional information concerning the apparent cause and corrective actions taken of reactor coolant pump 2B2 seal failure was reviewed by RO:II.
Item is closed.
(Details I, paragraph 9)
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3.
A0 270/74-3, " Failure to Verify Containment Intecrity Prior Prior to Unit Startup"
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Prior to unit startup on May 5, 1974, the reactor coolant temperature and pressure were raised above the limits specified in Technical specification 3.6.1, before containment integrity of the personnel hatch was verified as required by Technical Specification 4.4.1.2.5(b).
Corrective measures described in the licensee's report, dated May 15, 1974, were reviewed by tha inspector.
Item is closed.
(Details I, paragraph 10)
4.
A0 270/74-4, " Leakage Fron LPI System Sample Line" on May 11, 1974, a sample line between valves 2LP-14 and 2LP-
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l 39 on the primary coolant side of the 2B LPl cooler was found to be leaking.
Abnormal Occurrence Report A0 270/74-4, dated May 24, 1974, was submitted to the Directorate of Licensing.
Corrective actions were reveiwed during this inspection. Item is closed.
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_RO Rpt[ No. 50-270/74-4-7-
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5.
A0 270/74-5; "Overpressurization of CF Tank 2A" On May 23, 1974, pressure in core flood tank 2A exceeded the Technical Specification limit by 5 psi.
The incident was attributed to difficulty to communicate operator instructions via the public address system, due to high background noise.
This item re=ains open pending completion of corrective measures described in the licensee's report dated June 3, 1974.
6.
A0 270/74-6, "Underpressurization of CF Tank 23"
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Pressure in core flood tank 2B dropped 7 psi below the Technical Specification limit.
Investigation revealed the incident was caused by failure to respond expeditiously to an alarm annunci-ator.
This item rer.ains open pending completion of corrective measures described in the licensee's report, dated June 4, 1974.
7.
UE 270/74-1, " Violation of Cooldown Procedure" Technical Specification 3.3.1.a was violated when both loops of the reactor building spray system were removed from service on January 22, 1974.
The unusual event report, dated February 20, 1974, was reviewed during this inspection.
Item is closed.
(Details III, paragraph 5)
VII. Other Significant Findings
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a.
Plant Startup Operation was resumed on May 23, 1974, thus ending the outage which began Janury 22, 1974, for reactor coolant pump seal repair.
A reescalation testing program was conducted to return plant testing to the orginial program.
Estimated date for completion of test program is August 1974.
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b.
Organization and Personnel Chances J. W. Sigman, Maintenance Supervisor, has transferred to another Duke plant.
D. M. Thompson has been appointed Maintenance Supervisor to replace J. W. Sigman.
This change was effective June'1, 1974.
.VIII..
Management Interview The management interview for this inspection was held in two parts.
The first part was held on May 17, 1974, with J. E. Smith, J. W. Ha=pton,
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O. S. Bradham, L. E. Schmid, R. M. Koehler and C. L. Thames in i
attendance.
Items discussed were the identified violation concerning transfer of byproduct material and resolution of previously identified.
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violations and an unresolved item.
(Details II, paragraphs 2 through 5)
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Part two was held on June.7, 1974, with J. E. Smith, J. W. Hampton, L. E. Schmid, O. S. Bradham, R. M. Koehler and J. Cox in attendance.
Items discussed included the following:
1.
Enforcesent Items
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a.
The failure of the NSRC to review plant modifications as prescribed by Technical Specification 6.1.2.2.i(2).
(Details I, paragraph 2)
b.
The failure to report the radioactivity detected in the component cooling system as required by Technical Specification
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6.6.2.1 A.(1).
(Details I, paragraph 3)
c.
. Abnormal occurrences and an unusual event that involved violations of Technical Specifications requirements.
(Details III, paragraphs 4 and 5)
2.
Previously Identified Enforcement Action The licensee's response and corrective action pertaining to the violation on reactor operations log were reviewed. (Details
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III, paragraph 2)
3.
Previously Reported Unresolved Items The status of previously reported unresolved items was discussed.
(Sum =ary,Section IV)
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RO Rpt. No. 50-270/74-4
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DETAILS I Prepared by:
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F. Jape, Reactor Inspegtor Date Facilities Test and Startup Branch Dates of Inspection: June 4-7, 1974 Reviewed By:
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R. C. Lewis, Senior Reactor Inspector Date Facilities Test and Startup Branch
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1.
Individuals Contacted Duke Power Company (DPC)
R. M. Koehler - Staff Engineer J. M. Davis - Maintenance Engineer E. E. Hite - Junior Engineer K. Staring - Operator L. E. Schmid - Operating Engineer
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0. J. Bradham - Technical Support Engineer A. Farabee - Junior Engineer Duke Construction
T. D. Mills - Assistant Field Engineer M. Linderman - Assistant Field Engineer
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2.
NSRC Review of Modifications Paragraph 6.1.2.2.i(2) of the Technical Specifications requires the NSRC to review proposed changes in equipment or systems which may j
constitute an unreviewed safety question.
Contrary to this requirement, the minutes dated August 28, 1973, November 19, 1973, January 9, 1974, and March 1, 1974, do not r'eflect that the NSRC is reviewing safety-related modifications.
For example, the following modifications were recently completed and there is no evidence that any were reviewed by the NSRC:
a.
0-145-S: This modification provided an additional pin to lock
the cap screw in the reactor coolant pump impeller hub and to permit both lock pins to be welded in place.
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b.
0-124-S: This modification authorized a change in the seal return valve ti=c delay frca 50 seconds to 5 seconds.' The change is to minimize the flow of hot reactor coolant into the seals after loss of normal seal water, c.
0-125-S: This modification provided an interlock to close the reactor coolant pump seal return isolation valve, d.
0130-S: This modification provides a means to measure seal leakage flow for each reactor-coolant pump.
Station modifications are processed as described in Administrative Procedure No. 10 " Nuclear Station Modification Policy for Safety-Related Modifications." This procedure was recently revised and reissued on May 9, 1974.
Section 4, " Responsibilities," limits the NSRC's responsibility to reviewing only those changes pre-determined to involve an unreviewed safety question.
This is contrary ta the Technical Specification requirement.
This same limitation is also reficcted in Section 6 of the Administrative procedure, " Review and Approvals," where it is stated that the NSRC is only required to review those changes predeternined to involve an unreviewed safety question.
The inspector stated that the Technical Specification requires all proposed safety-related modifications to be reviewed by the NSRC before it is implemented and that the committee should review the
' written safety evaluation in which the determination that the change does not involve an unreviewed safety question.
3.
Radioactivity in the Component Cooling System Evidence of radioactivity in the component cooling system (CCS) was detected in June 1973, and since July 1973 the activity has exceeded twice the background level and remained essentially at that level.
In August 1973, PT 202/4, "HPI and CCS Leakage Test," was conducted to determine the source of the activity.
Results were inconclusive.
Licensee Management stated that efforts to resolve the problem are continuing.
Region II was informed by telephone on April 19, 1974, of the activity in the CCS.
The licensee's representative stated that the water chemistry experience at Oconee Nuclear Station was to be discussed at the American Power Conference on May 1, 1974, and that DPC wanted AEC to know of the situation prior to public release.
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During this inspection, the licensee was informed that this situation should have been reported as an abnormal occurrence when the activity was first detected.
Failure to report the event is considered to be contrary to Technical Specification 6.6.2.1 A.
Licensee management stated that a report would be submitted by June 14, 1974.
4.
Station Review Committee The licensee's reply to the unresolved ites on SRC review of abnormal occurrences, dated March 21, 1974, was reviewed by the inspector. The SRC minutes for the period January 24, 1974, through May 30, 1974, were examined. During this period the SRC met 81 times and reviewed 45 incidents.
Six of these incidents were classed as abnormal occurrences and have been reported to the AEC as required by Technical Specification 6.6.2.1.
The remaining 39 incidents were classed as "other" and were of such a nature that did not require reporting to the AEC.
A log of recommendations made bv the SRC in their review of
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incidents is maintained to ensure followup.
The inspector reviewed this record and found that each SRC recommendation ~was assigned to a member of the station's organization for followup.
The status of all recommendations is reviewed 'cy the Plant Superintendent on a monthly schedule. There were no questions or comments on this
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record and this previously identified unresolved item is considered closed.
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5.
High Energy - Line Break Modifications l
I The modifications installed to protect against postulated piping breaks were inspected to determine construction completness. The inspection consisted of a visual check of the penetration room blowout panels; reinforcement of the penetration room walls; cable tray impingement deflector; emergency feedwater bypass piping, valves, hanges and restraints; main feedwater restraints; and main steam line restraints.
The inspector found that all of the modifications have been completed.
The licensee has revised OP 1102/01, " Controlling Procedure for Unit Start-Up," to require a check that all reinforced doors are closed.
The licensee has alse provided PT 600/1 which requires that all reinforced doors be checked once per shift.
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This previously identified unresolved item is considered closed.
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Power Operated Relief valve, RC-66
Region II has completed a review of reports submitted by DPC concerning the valve wall thickness and integrity of the power operated pressurizer relief valves installed on Units 1, 2 and 3.
The results of this review satisfactorily answered the inspector's questions.
There are no further questions or co=ments concerning the use of RC-o6 as a reactor coolant boundary valve.
Resolution of this previously identified unresolved item relieves DPC of their commitment to operate with the block valve (RC-41) closed.1/
7.
_V1tal Bus static Inverters On February 10, 1974, while Unit 2 was in a cold shutdown condition, three of the four XC vital bus inverters tripped concurrently while performing PT 610/53, "Electromechanical Relay Breaker Trip Test,"
at Unit 1.
The results of DPC's investigation of the cause of this incident are summarized in a letter report to N. C. Moseley, Region II, Director, from A. C. Thies, DPC, dated April 17, 1974.
The letter report revealed that three of the four static inverters were tripped by a shunt trip device when AC power to the battery chargers was lost.
The shunt trip was installed as a convenience item for construction and maintenance.
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To prevent recurrence of this incident, the undervoltage shunt trip is to be disconnnected and this function will be deleted from the
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inverters.
Station modifications request 0-223-D was issued April 23, 1974, to authorize removal of the shunt trip devices from each of the four inverters for Units 1, 2 and 3.
The request is currently undergoing review and hence the work has not been co=pleted.
Item will re=ain open.
8.
Regulatory Operations Bulletin 74-1, " Valve Deficiencies" In response to R0B 74-1, the licensee has examined all Walworth valves and found no deficiencies.
The examination was conducted as prescribed by PT 0/270/36, " Inspection of Motor Operated Walworth
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Valves." The data sheets indicate that there are 39 Walworth
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valves in service at Oconee 2.
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1/ Letter to A. Giambusso, DL, from A. C. Thies, DPC, dated October 5, 1973.
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RO Rpt. No.'50-270/74-4 I-5
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The licensee's response to this ROB is dated January 31, 19,74.
The inspector had no questions or coc=ents on this matter.
9.
Seal Leak on Reactor Coolant Pu=p 2B2 Corrective actions described in the licensee's report, A0-270/74-2 dated February 1, 1974, and A0-270/74-2A, dated April 30, 1974, were reviewed b the inspector.
The corrective actions consisted primarily of mcdifications of the reactor coolant pump seal injections system and an.mproved method for ensuring retention of the Spirol pins on the era of.the pump shaft.1/
To demonstrece proper operation of the reactor coolant pumps following eacpletion of the corrective actions, the licensee has conducted TP 200/53, " Retest of Reactor Coolant Pumps After Repairs and Modifications," starting April 25, 1974, and ending May 22, 1974. The inspector reviewed this procedure and had no comments or questions.
10.
Failure to Verify Containment Integrity Prior to Startup Prior to unit startup on May 5, 1974, reactor coolant temperature and pressure were raised above the limit specified in Technical Specification 3.6.1, before. containment integrity of the personnel hatch was verified as specified by Technical Specification 4.4.1.2.5(b).
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- Reactor. coolant pressure and temperature were reduced and on May 6, 1974, at 0100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, PT 150/8, " Personnel Hatch Leak Rate Test,"
was satisfactorily conducted. Measured leakage rate was within the acceptance criteria of 0.5 pounds / hour.
The SRC reviewed this incident on May 14, 1974, and recommended that Adminstrative Procedure No. 11 be clarified to prevent postponement of tests beyond that required by Technical Specifications.
Administrative Procedure 11 was revised, reissued on June 4, 1974, and now clearly states that Technical Specifications are not to be contradicted or violated.
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1/ See.RO Inspection Report No. 50-270/74-3, paragraphs 4 and 5
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for descriptions of the modificationc.
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'/h DETAILS'II Prepared by:. ' '
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I enkins, Radiation Specialist
.Date G.R.J,dicalandEnvironmental
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Radioly
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Protection Branch
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Dates of Inspection: May 14-17,1974 I
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Reviewed by :
z J/ T. SutMerland, Chief Radiological and Envirencental Protection Branch
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1.
Individuals Contacted J. E. Smith - Plant Superintendent L. E. Schmid - Operating Engineer R. M. Koehler - Staff Engineer C. L. Thames - Health Physics Supervisor M.. G. Kriss - Assistant Health Physics Supervisor D. L. Davidsen - Assistant Health Physics Supervisor F. Stansell - Junior Engineer 2.
Transfer of Bvproduct Material a.
On February 21, 1974 the licensee transferred a metal "0" rir.g, that had been removed f rom Unit 2 reactor ecolant pump 2 A-1,
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to Bingham-Willamette Company, Portland, Oregon.
Review of the Radioactive Shipcent Record and discussion with licensee represen-tatives revealed that the iten contained byproduct caterial as j
fixed contamination when transferred to Bindaan-Willamette Company j
under License No. ORE-0027-1.
On questioning, a licensee repre-sentative stated that the basis for the transfer was a letter to the licensee f rom the Site Operations Manager, Babcock and Wilcox,
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which stated that Bingham-Willamette Company is licensed to handle radioactive material by the State of Oregon under Oregon license number 0027-1.
(The inspector was provided a copy of this latter.) The licensee representative acknowledged that he did not otherwise attempt to verify that the recipient was authorized to receive the =aterial being transferred.
The in-spector stated that this was an apparent violation of 10 CFR 30.41 Transfer of Byproduct Material which (paragr. ph (c)) states in part that, before transferring byproduct =aterial to a licensee,
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the licensee transferring the material shall verify that the
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transferree's license authorizes the receipt of the type, form,
and quantity of byproduct material-to be transferred.
The in-
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spector also informed the licensee that License No. ORE-0027-1 A
only authorizes sealed sources used in an industrial radiography program and possession of 36,000 lbs. of depleted uranium.
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b.
Af ter reviewing the Radioactive Shipment Record of this transfer, the inspector pointed out several itens on the form representative of carelessness in adhering to estab11.shed procedure.
A licensee representative stated that greater care and attention would be
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applied to these records in the future.
i 3.
Failure to Perform Tritium Monitoring The licensee response to this violation was reviewed during this inspection, including a review of the calculation referenced in the response. The health physics superviaor agreed that in future
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situations involving a spill or release of radioactivity, prompt
surveys will be made, where feasible, to document levels of
radioactivity.
The inspector had no further questions concerning this item.
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4.
Improper Alarm Setpoint
i The licensee response to this violation was reviewed during this j
inspection.
The alarm setpoints of both 2-R1A-46 and 1-R1A-46 were verified by the inspector to be set properly.
Records of weekly checks
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' of the setpoints on radiation monitors, which were started on April 3, 1974, were reviewed.
The inspector had _no further questions concerning this item.
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5.
Records of Radioanalysis Results (74-1/3)
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Previous inspections had revealed that the licensee frequently entered "<MDA" on radioanalysis records when no radioactivity was detected, and that statistical uncertainties were not specified in
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the records.
Discussions with licensee representatives and review of records during this inspection showed that:
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actual value of minimum detectable activity (MDA) is now recorded where analysis results are less than that value;
b.
standard deviations are now specified for results of samples where low activities are measured, i.e., environmental samples, j
as well. as all samples ' counted with the _ Ge(Li) detector;
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I R0 Rpt. No. 50-270/74-4 II-3
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a written procedure has been developed as guidance for computing c.
MDA and standard deviation.
A licensee representative agreed to continue to evaluate the use of statistical uncertainties for other sample analyses, where appropriate.
The inspector had no further questicns on this item.
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I RO Rpt. No. 50-270/74-4 III-1
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DETAILS III Prepared By :'-
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K. W. Whitt, Reactor Inspector,
Date Facilities Test and Startup Branch Dates of Inspection: June 5-7, 1974 Reviewed By: [.d.
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R. C. Lewis, Senior Reactor Inspector Date Facilities Test and Startup Branch
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1.
Personnel Contacted Duke Power Cocpany (DPC)
J. E. Smith - Plant Superintendent J. W. Hampton - Assistant Plant Superintendent
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R. M. Koehler - Staff Engineer S. A. Holland - Assistant Operating Engineer L. E. Schmid - Operating Engineer H. R. Lowery - Shif t Supervisor
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M. Kimray - Shift Supervisor
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T. Crawford - Assistart Shif t Supervisor 2.
Review of Loqs The inspector reviewed the shift cupervisor and console logs for the time period f rom May 23, 1974, through June 6,1974.
The logs appeared
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to have been prepared in accordance with the instructions provided by Standing Order No. 9, " Reactor Operators Log."
The DPC response to the violation of this subject which was reported in R0 Report No. 50-270/74-1 was reviewed.
An interoffice letter from the operations engineer to the shift supervisors dated January 29, 1974, informed the operations personnel that better log records were necessary.
Specifically, the letter stated that log entries were to be made at the ti=e of the incidents.
Facts, as kncvn, must be recorded at this time, and further descriptiens recorded af ter appropriate investigations are conducted.
This item is considered closed.
3.
Ccamunication of Activities Af fecting Operations This unresolved item was initially discussed in RO Report No. 50-270/74-1.
Each incident report is new being routed to the control rooms for review by the operations personnel.
The reports are filed in an " Incident In-vestigation Report ihnual" in the control room library which is located in the Unit 2 supervisor's office.
When incidents are determined to be abnormal eccurrences or unusual events, reports are transmitted to the
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RO Rpt. No. 50-270/74-4 III-2
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Copies of these reports are also routed to the control room and filed in the " Incident Investigation Report Manual." A licensee representative stated that RO reports and DPC replies to questions raised by the reports are being routed to the control room and a report file is being established similar to that used for the incident reports.
This item is considered closed.
4.
Reactor Coolant Sys tem Cooldown Limitation On January 4, 1974, Unit 2 tripped from 75* power as a result of a switchyard isolation. The reactor coolant pumps and the main feedwater pumps were rendered inoperable. The unit was cooled down by natural circulation. The cooldown rate exceeded that specified by Technical Specification 3.1.3.
RO:II was notified on January 4, 1974, and Abnormal Occurrence Report A0-270/74-1 was transmitted to the Directorate of Licensing on January 17, 1974. The abnormal occurrence report and corrective actions were reviewed during this inspection.
This item is
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considered closed.
5.
Removal of Both Loops'of Reactor Building Spray System From Service While cooling down Unit 2 on January 22, 1974, both trains of the reactor building spray system were removed from service, while the primary system pressure was greater than 600 psig and the temperature was greater than 300*F.
This is a violation of Technical Specification 3.3.la which requires that one reactor building spray pump and its associated spray nozzle be operable when fuel is in the core and the i
reactor coolant system pressure is greater than 350 psig and the.
- temperature is greater than 250*F.
The event was reported to RO:II on January 23, 1974, and Unusual Event Report UE-270/74-1 was transmitted to the Directorate of Licensing on February 21, 1974, i
The inspector reviewed the unusual event report and corrective action i
during this inspection.
This item is considered closed.
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