IR 05000373/1981006
| ML20003G090 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 03/13/1981 |
| From: | Reimann F, Shepley S, Walker R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20003G085 | List: |
| References | |
| 50-373-81-06-01, 50-373-81-6-1, NUDOCS 8104280233 | |
| Download: ML20003G090 (14) | |
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U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT
REGION III
Report No. 50-373/81-06 Docket No. 50-373 License No. DPPR-99 Licensee: Commonwealth Ediron Company Post Office Box 767 Chicago, IL 60690 Facility Name:
LaSalle County Nuclear Station, Unit 1 Inspection At:
LaSalle Site, Marseilles, Illinois Inspection Conducted: January 5 through February 6, 1981 1~
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Inspectors:
R. D. Walker
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N.ut hN 3J 3 'd I S. E. Shepley
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Approveo By:
F. W. Reimann, Actin Chief Reactor Projects Section 1C Inspection Summary
Inspection on January 5 through February 6,1981 (Report No. 50-373/81-06)
Areas Inspected: Routine, resident inspector, preoperational inspection consisting of a review of licensee action on previous inspection findings, IE Bulletins and Circulars received by the licensee since last inspection, emergency procedures for coping with ATWS events, electrical distribution /
instrumentation power supplies, preoperational test witnessing, followup on significant events which occurred during the inspection period, inspection activities preparatory to licensee issuance, and a plant walkthrough/
operational status review. The inspection involved a total of 249 inspec-tor-hours onsite by two NRC inspectors including 39 inspector-hours onsite during off-shifts.
Results: One item of noncompliance at the Severity Level V (Failure to comply with procedures required by 10 CFR 50, Appendix B, Criterion XIV).
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e DETAILS 1.
Persons Contacted
- B. B. Stephenson, LaSalle County Station Project Manager
- R. H. Holyoak, Station Superintendent R. D. Kyrouc, Quality Assurance Engineer
- G. J. Diederich, Station Operating Assistant Superintendent
- R. D. Bishop, Administrative and Support Services Assistant Superintendent C. W. Schroeder, Technical Staff Supervisor R._Raguse, Senior Operating Engineer J. M. Marshall, Operating Engineer
- W. Huntington, Assistant Technical S aff Supervisor H. J. Hentschel, Assistant Technical Staff Supervisor F. Lawless, Rad-Chem Supervisor
- E. E. Spitzner, Startup Coordinator G. E. Groth, Construction Engineer
- E. Stevak, Quality Assurance
- L. W. Duchek, Project Management Staff
- J. Bowers, Onsite Nuclear Safety Engineering Group (ONSEG)
- T. Borzym, Security Administrator
- J. A. Ahlman, Quality Assurance Engineer
- T. E. Quaka, Site Quality Assurance Superintendent The inspector also interviewed other licensee employees including members of the technical, operating, and construction staff, as well as certain licensee contractor employees.
- Denotes persons present at management interview onsite.
2.
Licensee Action on Previous Inspection Findings (Closed) Unresolved Item (373/79-15-28) Review of response to IE Bulletin 79-08.
The inspector reviewed the licensee response to l
this bulletin and had the following findings.
With respect to items 6 and 8 of " Actions to be Taken by the Licensee's."
The licensee's out of service procedure LAP 900-4, Revision 10, dated January 9,1981 was reviewed to ascertain if the actions required were addressed. The inspector noted that, with respect to returning equip-ment to service, the licensee has not addressed suitable redundancy as required by the bulletin :nd NUREG 0737, item I.C.6.
The inspector
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l was informed that Commonwealth Edison Company (CECO) is currently addressing this issue generically for all CECO nuclear stations.
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inspector will review final resolution of this as open itea (373/81-06-01).
With respect to item 7 of " Actions to be Taken by Licensee's."
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e The inspector noted that several systems do not isolate on high radia-tion signals, but do isolate on other Primary Containment 1 solation Signals (PCIS) as required by the FSAR. The licensee has submitted any logic changes that are required as a result of their review to the Office of Nuclear Reactor Regulation (NRR) for review as a part of their response to NUREG 0578. Any review of logic changes required by the inspector will be accomplished as part of the TMI Task Action Plan.
With respect to item 6 of " Actions to be Taken by the Licensee's."
Current licensee startup checklist procedures (LGP 1-S1, Rev. 2, 10-17-80 for Master Startup checklist; LGP 1-S2, Rev. 1, 6-23-80 for Minimum Startup; and LGP 1-S3, Rev., 1, 6-23-80 for Pre Startup Line-up) do not identify sufficient requirements to assure that criteria identified in NUREG 0737 Task Action Item II.K.1 (5) are met for ESF valves and equipment. The inspector will review this under previous open item (373/79-23-04). The inspector verified that for all other portions of IE Bulletin 79-08, the licensee's internal response included the information required to be reported by the licensees, that the internal response included adequate corrective action commitments based on information presented in the bulletin, that licensee management forwarded copies of the internal response to the appropriate onsite management representatives, that information discussed in the licensee's internal response was accurate, and that corrective action taken by the licensee was as described in the response.
(Closed) Unresolved Item (373/79-23-05) Operating procedures which are used to return Engineered Safety Feature Systems to service after auto-matic actuation did not have specific precautions against an operator overriding or bypassing Engineered Safety Features, unless specific conditions require such an override or bypass.
The inspector reviewed the following procedures and ascertained that this problem has been properly addressed:
l Vice President's Instruction No. 1-0-17 dated 10-26-79 Procedure LAP 1600-2 Revision 13 Dated 10-07-80 Procedure LOP-HP-04 Revision 1 Dated 05-29-80 Procedure LOP-LP-03 Revision 2 Dated 12-09-80 Procedure LOP-RH-12 Revision 2 Dated 12-17-80 Procedure LOP-RI-03 Revision 2 Dated 12-15-80 (Closed) Unresolved Item (373/79-23-03) " Equipment Out of Service i
Procedure" LAP 900-4 at ' Caution Card Procedure" LAP 900-12 did not l
have proper tagging methods that elimate potential to obscure j
necessary status indication.
The inspector reviewed LAP 900-4, Revision 10, dated January 9,1981 and LAP 900-12, Revision 4 dated August 26, 1980 and found that controls have been implemented which eliminate this potential.
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(Closed) Unresolved Item (373/79-15-22) Minor problems with Operating Prccedure LOP-RI-03, " Reactor Core Cooling System Shutdown." The inspector reviewed Revision 2, dated December 15, 1980 of this pro-cedure and found it satisfactory.
(Closed) Unresolved Item (373/79-15-25) Minor problems with Operating Procedure LOP-LP-03, " Low Pressure Core Spray System after Automatic Initiation." The inspector reviewed Revision 2, dated December 9, 1980 of this procedure and found it satisfactory.
(Closed) Unresolved Item (373/79-15-06) Minor problems with Operating Procedure LOP-SC-01, " Filling, Venting and Draining the Standby Liquid Control System." The inspector reviewed Revision 2, dated January 12, 1981 of this procedure and found it satisfactory.
(Closed) Unresolved Item (373/79-15-11) Minor problems with Operating Procedure LOP-SC-02, " Standby Operation of the Standby Liquid Control System." The inspector reviewed Revision 2, dated January 12, 1981 of this procedure and found it satisfactory.
(Closed) Unresolved Item (373/79-15-12) Minor problems with Operating Procedure LOP-SC-03, " Standby Liquid Control Solution Tank Draining."
The inspector reviewed Revision 2, dated January 12, 1981 of this procedure and found it satisfactory.
(Closed) Unresolved Item (373/79-15-13) Minor problems with Operating Procedure LOP-SC-04, " Standby Liquid Control Solution Tank Filling."
The inspector reviewed Revision 2, dated January 31, 1981 of this procedure and found it satisfactory.
(Closed) Unresolved Item (373/79-38-02) The licensee had no operating or surveillance procedures which identify a valve line up procedure for valves associated with instrumentation inside the root stops.
The inspector reviewed Procedure LIP-GM-09, Revision 0 which is a generic procedure to be used in conjunction with individual check lists within the surveillance procedures to perform this alignment.
The inspector found this satisfactory.
(Closed) Unresolved Item (373/79-44-09) Confusing nomenclature on alarms for monitoring ECCS system line integrity on control room panel 1H-13-P601. The licensee has re-engraved these alarms to eliminate this confusion.
(Closed) Unresolved Items (373/80-56-03 and 373/79-38-41) Review of IE Bulletins 80-24 and 75-09 respectively. These bulletins were sent to the licensee for information only and do not appear to have applica-bility to the licensee either now or when an operating license is issued.
Based on the above evaluation by the inspector, the licensee was not asked to submit an answer for review.
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(Closed) Unresolved Items (373/80-24-06, 373/80-56-06, 373/79-44-02, 373/79-33-08, 373/79-33-16, 373/79-38-07, and 373/79-38-06) Review of Circulars 80-08, 80-23, 79-24, 79-11, 78-09, 77-03 and 77-02 respectively.
The inspector verified that for the above IE Circulars, the circulars were received by the licensee management, a review for applicability was performed, and for circulars applicable to the facility further action taken or planned is appropriate.
(Closed) Unresolved Item (373/80-56-07) This item is upgraded to a
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noncompliance at the Severity Level of V.
Description of the circum-stances are as follows. On December 23, 1980, the licensee was con-ducting a portion of preoperational test PT-DG-202A, which called for simultaneous initiation of diesel generators 0, IA, and IB.
Diesel generator 0 was forced to be shutdown before diesel generators IA and IB when a 1.5" line was cut into, which poured water onto eleva-ation 673' in the north end of the Reactor Building. The line is the Core Standby Cooling Equipment Cooling System (CSCS) water line, from the diesel generator 0 cooling water pump discharge, which supplies the LPCS motor cooler.
The inspectors' subsequent investigations revealed that the cut was made to reroute the line (IDG22A), but the part of the line cut was not properly tagged out of service, isolated, or safe for work.
This resulted from the construction workers doing something different than what operations expected the construction workers to do.
The problem appears to have resulted when the supervisor in charge of the work (1) did not review the proposed outage with the con-tractor before the outage, (2) did not check the outage after it was in place to assure the line was safe for work, and (3) did not assure that the supervisor directly responsible for the work began work on the part of the line that was protected. As a result the outage was incorrect for the work.
The licensee states that the person in charge of the work is required to do all these, but the inspectors can find only the second require-ment in written procedures.
This is considered a further example of problems identified in a non-compliance documented in Inspection Report 50-373/80-05 and answered in the D. L. Peoples letter to James G. Keppler, dated March 24, 1980.
The inspectors are concerned that this lack of control over the activities of construction personnel will reduce the safety of opera-tions once Unit I becomes operational. This is considered an item of noncompliance (373/81-06-02).
No other items of noncompliance were identified.
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3.
Review of IE Bulletins Received Since Last Inspection Report The licensee has received IE Bulletin 81-01 since the last inspec-tion report. This bulletin was sent to the licensee with instruc-tions that would indicate that it was for information only. The inspector and his section chief have had discussions with IE Head-quarters and the Licensee's Nuclear Licensing Administrator, Mr. L. DelGeorge to clarify the status of this bulletin with respect to LaSalle County Nuclear Station, a Near Term Operating License (NTOL) construction permit holder. All parties reached agreement that the bulletin response under " Actions to be Taken by the Following Licensees Holding Construction Permits" must be performed by the licensee (LaSalle County Nuclear Station) because they are applicable. The inspector will review the licensee's response under open item number (373/81-06-03).
No items of noncompliance were identified in this area.
4.
Review of IE Circulars Received Since Last Inspection Report The licensee has received IE Circular 81-01 since the last inspection report. The licensee is still reviewing this circular at this time, and the inspector will review the licensee's response during a subsequent inspection under open item number (373/81-06-04).
No items of noncompliance were identified in this area.
5.
Inspection of Emergency Procedures for Coping with ATWS Events at Operating Power Reactors a.
Inspection Purpose The purpose of this inspection was to verify that the licensee
has emergency procedures adequate to respond to Anticipated l
Transient Without Scram (ATWS) events.
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Inspection Requirements i
i (1) Review licensee procedures that address any or all of the following plant conditions:
(a) Failure to scram when required.
l (b) Move or drive control rods.
(c) Failure to automatically scram when a parameter exceeds its trip value.
(d) Failure to complete scram when initiated automatically or manually.
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l (e) Criteria for use of the Standby Liquid Control System l
(SLCS).
l (f) Reactor trip or scram.
i (g) Anticipated transient without scram.
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(2) Review the authorities and responsibilities of operators governing the use of SLCS.
c.
Acceptance Criteria (1) IE Bulletin 80-17, Action No. 4 of " Actions to be Taken by Licensee" which states the licensee's emergency operating procedure to scram should have operator actions which include:
(a) Place the reactor mode switch in a position other than RUN.
(b) Determine whether either of the two conditions below exist:
1.
Five (5) or more adjacent rods not inserted below the 06 position.
2.
Thirty (30) or more rods not inserted below the 06 position.
(c)
If either condition b.1) or b.2) exists:
Trip the recirculation pumps.
2.
Insert rods manually.
If rods cannot be inserted manually, alternately reset the RPS and scram the reactor until all rods are fully inserted.
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Manually open or bypass the scram instrument volume drain and vent valves, if possible.
(d)
If, at any time, either condition b.1) or b.2) exists and either RPV water level cannot be maintained or suppression pool water temperature cannot be maintained below the suppression pool water temperature scram limit, initiate the SLCS.
(e) Review the Browns Ferry occurrence with all licensed operators and train them in the procedures to recog-nize and mitigate the event. Verify that preliminary training of operators is completed within 10 days of the date of this Bulletin and that full training is completed aithin 30 days of the date of this Bulletin.
(2) Guidance contained in Inspection and Enforcement TI 2515/46 issued December 5, 1980 which states that "The operator should have complete authority to activate the Standby Liquid Control System, and he should be responsible for doing this when the situation requires it.
If the Standby Liquid Control System (SLCS) is key operated, the key must be readily available to the operator. Criteria for the use of SKOS relative to in-ability to insert nagative reactivity by other means should be included in emergency procedures."
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O d.
Inspection Results The inspector reviewed the following items and had the^following findings with respect to each of the' inspection requirements identified:
(1) Procedure LAP 1600-2 " Conduct of Operations," Revision 13, dated October 7, 1980, step F.2.c contains instructions which pertain to the operator's complete authority to initiate ti.e SLCS during an ATWS event. The instructions do not adequately define the operators authority and should be changed. (373/81-06-05)
(2) Procedure LOA SC-02 " Initiation of Standby Liquid Control,"
Revision 0, dated July 10, 1979 step C.1 and C.2 contain instructions for the operator to obtain and use the key to initiate the SLC..The instructions do not meet current criteria and should be changed open item.
(373/81-06-06)
(3) Procedure LGP 3-2, " Reactor Scram," Revision 1, dated June 23, 1980 contains no reference to the ATWS procedure LOA-NB-09 nor does it contain any instructions to the operator at step F.3 (check to see that all control rods have inserted to full in) to determine conditions as identified in part 1 of the acceptance criteria referenced above. This problem can be taken care of by referring the operator to procedure LOA-NB-09 if significant control rods fail to scram.
(373/81-06-07)
(4) " Draft" Procedure LOA-NB-09 " Transient with Failure to Scram" should be modified to incorporate the conditions identified in part 1 of the acceptance criteria referenced above within the immediate and subsequent operator actions of the procedure as discussed with the inspector rather than in the discussion section of the procedure.
(373/81-06-08)
(5) Training on the Browns Ferry Event is being provided to all licensed operator candidates in the licensees current license exam refresher course. The licensee will provide training on procedures to recognize and mitigate the event as part of their routine training when the procedures are modified.
No items of noncompliance were identified in this area.
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6.
Review of Electrical Distribution / Instrumentation Power Supplies The inspector conducted an independent review of instrumentation power supplies in response to concerns addressed in IE Bulletin 79-27. The inspector had the following concerns with respect to this review which he will refer to the Office of NRR for resolution:
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The power supply for certain Emergency Core Cooling System (ECCS)
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control room indication is from the 6.9 KVac distribution network.
This network is lost on loss of offsite power and cannot be re-stored via any onsite power source. The inspector feels this does not meet requirements defined in Regulatory Guide 1.97, Revision 0 and amplified in Revisions 1 and 2 to this Regulatory Guide. The systems effected are as follows:
(1) Low Pressure Core Spray System (LPCS) system pressure and flow indication (2) Residual Heat Removal System "A" (RHRA) system pressure and flow indication (3) Residual Heat Removal Service Water System (RERSV) system flow indication The licensee's operating staff became aware of this problem and requested corrective action via an Operating Design Change Report (ODCR) #PT-RH-101-4 which was addressed to the Station Nuclear Engineering Department (SNED). The ODCR was subsequently not approved by SNED on the basis that the instruments effected are passive and therefore, do not require essential power. The inspector does not agree with SNED's disposition of the ODRC because loss of this instrumentation violates current regulatory requirements defined in the referenced Regulatory Guide. The regulatory guide specifically addresses instrumentation needed to properly determine whether ECCS systems are properly functioning or to allow the operator to make manual control functions of the ECCS required by accident condition. The Regulatory Guide also addresses single component failure and electrical power supplies with respect to this instrumentation. The inspector has learned that SNED is rereviewing this item and may reconsider the earlier diaposition of the ODCR. The inspector will review resolution sf this matter under open item number (373/81-06-09).
b.
The power supply for all Source Range Monitor (SRM) and Inter-mediate Range Monitor (IRM) detector drive motors is from 480 VAC MCC 136Y-2. These detectors are used to determine that the reactor shutdown must be capable of performing tFis function independent of single failure of a component.
In the case in question, loss of a single power supply would indeed, result in all detectors losing the capability to be driven into the reactor core. The inspector has not been able to obtain from the licensee any verification that these detectors can determine suberitically in the reactor core if they are in their fully withdrawn position.
This question is complicated by an assumption of a scram of only portions of the control rods such as that described in IE Bulletin 80-17.
The inspector will review resolution of this matter under open item number (373/81-06-10).
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Both trains of the steam condensing mode of the RHR system will c.
be lost on the occurrence of loss of offsite power concurrent with the failure of the most limiting single component failure.
This matter is of concern because the licensee takes credit in the FSAR at 3.1.2.4.5 for the steam condensing mode of RHR to meet 10 CFR 50, Appendix A, Criterion 34 requirements.
The licensee also takes credit in the FSAR at 6.2.1.1.3.1.4 for the steam condensing mode of RHR for recovery operations after a small size break. The credit in this case is taken for limiting the blowdown flow to the drywell to a period of six hours after the accident and thus minimizing the duration of the super-heat ecndition in the drywell after the accident. The use of the steam condensing mode of RHR in this manner is governed by 10 CFR 50, Appendix A, Criterion 38.
The inspector will review the resolution of this matter under open item number (373/81-06-11).
d.
The FSAR Section 5.4.7.1 discusses the Safety Design Bases of the RHR system and states that the only safety related function of the RHR system is the low pressure coolant injection (LPCI)
mode. The inspector finds this statement to be in error as the shutdown cooling mode (per 10 CFR 50, Appendix A, Criterion 44),
containment cooling mode (Per 10 CFR 50), Appendix A, Criterion 38),
and steam condensing mode (Per 10 CFR 50, Appendix A, Criterions 34 and 28) as well as the LPCI mode of the RHR system have safety-related functions. The inspector feels that the current statement of the Safety Design Bases is misleading and that the licensee should be required to change the statement via an amendment to the FSAR. The inspector will review the resolution of this matter under open item number (373/81-06-12).
e.
The current design of the load shed features associated with ESF electrical buses powered by the diesel generators is such that if the diesel generator is synchronized to its bus for testing and is loaded during the occurrence of a loss of offsite power, the
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loadshed features will not activate and the diesels generators will trip on over current. The inspector discussed this with the licensee and determined that the test engineer is not certain i
if this problem will be corrected by a modification to the diesel l
generator in the near future.
It is the inspector's belief that l
the current design should be modified to correct this problem and that subsequent to the modification this feature should be tested.
The inspector will review the resolution of this matter under open item number (373/81-06-13).
No items of noncompliance were identified in this area.
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l 7.
Preoperational Test Witnessing The inspector witnessed portions of the preoperational tests of the Control Rod Drive System. The particular portion of this preopera-tional test that was witnessed by the inspector verified the ability l
of the system to complete a full core scram. Documentation of the I
details of this inspection effort will occur in IE Inspection Report i
No. 50-373/81-07.
8.
Review of Proposed Facility Technical Specifications The licensee has issued a complete new revision of the proposed Facility Technical Specifications and the inspector is currently
reviewing that document as well as recent changes to the Technical i
Specifications issued by the Office of NRR.
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9.
Followup on Significant Event that Occurs While Inspector is on Site a.
On January 22, 1981, at approximately 2:45 p.m. in the afternoon, a fire occurred in the Unit I drywell. This fire was not reported to operations until about 10:00 a.m., January 23.
The fire seems
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to have been the result of welding slag or a hot I beam falling against cables in a vertical cable pan. The fire was apparently extinguished by the construction subcontractor organization. The inspector observed what seemed to be the residue of dry chemical
extinguishing agents in the tray and on the cables and found that
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l of the 15 safety-related cables in the tray, seven had obvious heat damage.
(See Table 1.)
The fire part of this incident has been referred to the NRC construction project inspector for
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followup. The delay in reporting part of the incident was i
investigated further. The inspector could not find a formal
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procedure that requires construction personnel to promptly report a fire to operations. The licensee has cosamitted to require construction personnel to immediately notify operations
of all fires. The licensee has evaluated the cable damage and intends to repair / replace the cables as denoted in Table 1.
No items of noncompliance were identified in this area.
b.
On February 2, 1981, at approximate 1v 7:35 a.m.,
a fire occurred in a wood construction shack locn ed in the Auxiliary Building.
The fire was immediately extinguished by the station fire bri-gade. The automatic sprinkler system failed to function because l
of a closed water supply valve. The fire appears to have been
caused by a worn power cord on a radio. No apparent damage
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occurred to safety related equipment.
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The licensee has instituted a program to:
check sprinkler l
system root valves; survey all temporary enclosures to deter-mine if each is needed; and conduct regular patrols throughout L
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all operating and construction spaces. The licensee sent samples of the fire debris to the Illinois Bureau of Scientific Services, Joliet Laboratory, for analysis.
In their response of February 16, 1981, the Bureau documented that no evidence of accelerants was found in the debris.
No items of noncompliance were identified in this area.
10.
Inspection Activities Preparatory to License Issuance (Status of Licensee Procedures and Preoperational Testing Program a.
Operating, Maintenance, Surveillance, Abnormal and Emergency Procedure Status The licensee projects 4981 procedures to be required in these areas. Currently the licensee has approved 4483 procedures, 451 procedures have been draf ted but not reviewed, and 47 procedures remain to be drafted.
b.
The licensee projects a total of 125 Preoperational Tests / System Demonstrations required for Unit 1 operation of which 115 of these are specific Unit I and the remaining 10 are specific to Unit 2.
The licensee reported that 122 systems have been turned over for preoperational testing, that 113 Preoperational Tests and System Demonstrations have been started, that 68 Preoperational Tests and System Demonstrations have been completed, and the pre-operational testing program is approximately 80% complete at this time. The licensee stated that final Preoperational Test or System Demonstration results for 17 tests are ready for NRC review, i.e.,
the entire test is complete and the results have been reviewed and accepted by the licensee. These seventeen tests are JT-DO-101
" Diesel Oil," PT-VY-102 "LSCS Equipment Ventilation,"
PT-PV-101 " Reactor Vessel Internals Vibration," PT-DO-201
" Unit 2 Diesel Oil," PT-FC-101 " Fuel Pool Cooling and Cleanup,"
PT-NR-101A " Source Range Monitors," PT-SC-101 " Standby Liquid Control," PT-VJ-101 " Machine Shop Ventilation," PT-V0-101 "Off Gas Ventilation," SD-WR-101 " Reactor Building Closed Cooling Water System," SD-CW-01 " Circulating Water System," SD-CX-102
" Rod Worth Minimizer," PT-AP-202 " Unit No. 2 DC Distribution,"
SD-CS-101 " Condensate and Condensate Booster System," SD-HD-101A
"Feedwater Heaters and Drains," SD-CY-101 " Cycled Condensate System," and SD-WE-101A " Equipment Drain and Waste Collector Process System."
c.
Deficiency Status The licensee is currently listing 1690 Station Operations defi-ciencies and 3944 Station Construction deficiencies as outstand-ing items. The licensee is still attempting to segregate these
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deficiencies into those that will impace fuel load and those that will not.
The licensee has reviewed approximately 2425 of these deficiencies for segregation and preliminary assessment is that 857 of those reviewed would need to be cleared prior to fuel load. The inspector expressed concern that the_ segregation of.these deficiencies into those that will effect fuel load versus those that will not is progressing slowly. The inspector will continue to review this matter.
No items of noncompliance were identified in this area.
11.
Plant Walkthrough/ Operational Status Review The inspector conducted walkthroughs and reviewed the plant operations status including examinations of control room log books, routine patrol sheets, shift engineers log books, routine patrol sheets, shift engineers log books, equipment outage logs, special operating orders, and jumper tagout logs for the period of January 5,1981 through February 6, 1981. The inspector observed the operations status during three off shifts during the same period as above. The inspector also made visual observations of the routine surveillance, functional, and preoperational tests in progress during the same period as above.
The inspector also made visual observations of the routine surveillance, functional, and preoperational tests in progress during this period.
This review was conducted to verify that facility cperations were'in conform nce i th the requirements established under 10 CFR and adminis-trative proce.ces.
The inspector conducted tours of Units 1 and 2 reactor, auxiliary, and turbine buildings throughout the period and noted the status of construction and plant housekeeping / cleanliness.
With respect to housekeeping / cleanliness, the inspector expressed concern for deteriorating conditions of general cleanliness throughout the facility. The inspector expressed particular concern for reactor vessel cleanliness. The inspector stated that something must be done to improve this area of licensee performance. The Station Project Manager acknowledged the inspector's concern and stated that something would be done. The inspector observed shift turnovers to verify that plant component status and problem areas were being turned over to a relieving shift.
No items of noncompliance were identified in this area.
12.
Management Interview The inspector met with licensee representatives (denoted in Paragraph 1) at the conclusion of the inspection period. The inspector summarized the scope and findings of the inspection activities.
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o Table 1 Obvious Licensee Cable No.
System Heat Damage Evaluation IRH250 RHR Yes Replace IRH646 RHR Yes Replace INB103 ADS Yes Repair in place IRH466 RHR Yes Repair in place IRH467 RHR Yes Repair in place ISC017 SLCS Yes Repair in place IRH463 RHR Yes Scortched casing no repair needed 1HP259 HPCS No No repair needed INB107 ADS No No repair needed 1RH247 RHR No No repair needed 1RH464 RHR No No repair needed 1RH644 RHR No No repair needed IWR057 RBCCW No No repair needed l
IWR059 RBCCW No No repair needed 1WRO77 RBCCW No No repair needed l
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