IR 05000269/1973008
| ML19322A886 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 08/01/1973 |
| From: | Jape F, Murphy C, Whitt K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19322A883 | List: |
| References | |
| 50-269-73-08, 50-269-73-8, NUDOCS 7911270642 | |
| Download: ML19322A886 (17) | |
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UNITED STATES
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ATOMIC ENERGY COMMISSION
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DIRECTORATE OF REGULATORY CPERATICNS g,,8ta re s 0
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RO Inspection Report No. 50-269/73-8
'icensee: Duke Power Company Power Building
422 South Church Street Charlotte, North Carolina 28201 Facility:
Oconee Unit 1 Docket No.: 50-269 License No.: DPR-38 Category:
B1 Location: Seneca, South Carolina Type of License: B&W, PWR, 2568 MW(t)
Type of Inspection: Routine, Unannounced Dates of Inspection: July 17-20,1973 Dates of Previous Inspection: June 20-21, 1973, and July 19, 1973 Principal Inspector: F. Jape, Reactor Inspector Facilities Test and Startup Branch Accompanying Inspector: K. W. Whitt, Reactor Inspector Facilities Test and Startup Branch Other Accompanying Personnel: None Principal Inspector:
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F. Jape, Reactor Inypeltor Date Facili es Test and Startup Branch
!73 Reviewed By:
C. E. Murphy, XhTC/
' Date Facilities Test and Startup Branch s
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E R0 Rpt. No. 50-269/73-8-2-SUMMARY OF FINDINGS
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Enforcement Action A.
Violations None B.
Safety Items None II.
Licensee Action on Previously Identified Enforcement Matters A.
Violations 1.
RO Letter Dated March 30, 1973, Item 4 DPC provided a satisfactory response to item 4 in their letter dated April 20, 1973, and in the unusual event report dated
Nby 4, 1973. Corrective actions described were verified during this inspection and RO:II has no further questions.
(Details II, paragraph 2)
2.
RO Letter Dated March 20, 1973 DPC provided a satisfactory response, dated April 11, 1973, and June 15, 1973, and RO:II has no further questions on this item at this time. The corrective actions described were verified during this inspection.
(Details I, paragraph 6)
B.
Safety Items None
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III. New Unresolved Items 73-8/1 Failure of Reactor Building Spray Valves to Open During ES System Testing All corrective actions described in DPC's unusual event report dated May 4,1973, have not been completed.
(Details II, paragraph 4)
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R0 Rpt. No. 50-269/73-8-3-I I
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Status of Previously Reported Unresolved Items
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73-7/1 Main Steam Relief Valve Popping Open Trip From 15% Reactor
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Power i
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Not inspected.
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73-6/1 Use of Miscellaneous Tests
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Not inspected.
l 73-4/2 Inverse Multiplication Plots
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Not inspected.
73-4/1 Reactor Coolant Pump Flow
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Not inspected.
73-3/1 Nitrogen In-Line Heater Desien Change Documentation
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The documentation for the nitrogen in-line heater design change was reviewed and the inspector had no further
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j questions. This item is resolved.
(Details I, paragraph 7)
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73-1/7 Emergency Operating Procedures
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j DPC has provided emergency procedures to cover the conditions previously identified as being deficient. This item is resolved.
(Details I, paragraph 5)
i 73-1/2 Resolution of Test Deficiencies
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Test procedure 1.
1.10/2, " Penetration Room Flow Pressure Drop and Filter Tests, has been rerun at 1000 cfm. The inspector'
reviewed the test results and had no further questions. This
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item is resolved.
(Details II, paragraph 5)
71-10/1 Flow Meter Error Analysis and Tests Not inspected.
71-7/1 Thin Walled Valves (R0 Inspection Report No. 50-269/71-5, i
Details C.3)
Not inspected.
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RO Rpt No. 50-269/73-8-4-I V.
Unusual Occurrences
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A.
Control Rod Drive Breaker Undervoltage Device (R0 Inspection Report No. 50-269/73-2, Summary of Details,Section IV)
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j Corrective actions described in DPC's report, dated February 23, 1973, were verified during this inspection and RO:II has no further questions.
(Details I, paragraph 2)
B.
Reactor Building Spray Pumos Inoperable During Reactor Operation Corrective actions described in DPC's report dated May 18, 1973,
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were verified during this inspection and RO:II has no further
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questions.
(Details I, paragraph 4)
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Failure of the Operating Mechanism to Fully Open the Core Flood
Line Isolation Valve, CF-1
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Corrective actions described in DPC's report, dated May 4, 1973, were verified during this inspection and R0:II has no further (
questions.
(Details II, paragraph 3)
D.
March' 6,1973, lAl Reactor Coolant Pump Oil Fire Corrective actions described in DPC's report dated May 4,1973, were verified during this inspection and R0:II has no further questions.
(Details II, paragraph 2)
E.
Failure of Reactor Building Spray Valves to Open During ES System Testing Corrective actions described in DPC's report dated May 4,1973, were inspected and it was found that actions have not all been completed. The licensee's representative stated that all
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ccrrective actions would be completed by September 1,1973.
This item will be carried as unresolved item No. 73-8/1.
(Details II, paragraph 4)
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Other Significant Findings i
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s VII. Management Interview The inspection results were discussed with J. E. S=ith, Plant Superintendent, and others of his staff on July 50, 1973.
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R0 Rpt. No. 50-269/73-8-5-A.
Control Rod Drive Breaker Undervoltage Device
The inspector stated that the corrective actions described in DPC's
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February 23, 1973, report had been vecified.
(Details I, paragraph 2)
B.
ROB 73-1, " Faulty Overcurrent Trio Delay Devices in Circuit Breaker for Engineered Safety Systecs" Corrective measures described in DPC's response to ROB-73-1, dated April 12, 1973, were verified by the inspector.
(Details I, paragraph 3)
C.
Reportable Events The status of one unusual event and three abnormal occurrences was dis cuss ed.
(Details I, paragraph 4 and Details II, puragraphs 2, 3, 4, and 8)
D.
Emergency Operating Procedures (EP)
The inspector commented that his review of EP's has been completed and there are no further questions on this item.
(Details I, paragraph 5)
E.
Enforcement Action The status of enforcement action, as described in Section II of the Summary of Findings, was discussed.
(Details I, paragraph 6 and Details II, paragraph 2)
F.
Previously Reported Untesolved Items The status of previously reported uaresolved items, as described-in Section IV of the Summary of Findings, was discussed.
(Details I, paragraphs 5 and 7 and Details II, paragraph 5)
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G.
Comments on Previously Reviewed Test Procedures
The inspector stated that a followup review of comments previously discussed has been completed.
(Details I, paragraph 8 and Details II, paragraph 5)
H.
Active EJectrical Jumper and Bypass Log The inspector stated that he had rereviewed this log as a followup from a previous inspection.
(Details I, paragraph 9)
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R0 Rpt. No. 50-269/73-8-6-I.
Followup on Comments on Alarm Procedures, Instrument Procedures, Maintenance Procedures and Ooerating Procedures
The status and resolution of comments discussed during previous inspections were reviewed. All have been resolved and the inspector had no further co= ment.
(Details I, paragraphs 10, 11, 12, 13, and 14)
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Sensitivity of Leak Detection Systems The inspector reviewed this matter with licenaee management.
This item remains active and will be reviewed on subsequent inspections.
(Details II, paragraph 6)
K.
Technical Specification Surveillance Requirements The licensee's representative stated that a revision of several surveillance requiri.ments will be requested.
(Details II, paragraph 7)
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R0 Rpt. No. 50-269/73-8 I-1
DETAILS I Prepared by:
4tM M/4L l'N~73
' F. Jape, Reactdf Irispector Date Facilities Test and Startup
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Branch
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Dates of Inspection: Jul 17-20, 1973 Reviewed by:
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C. E, Murphy / Chiff Date Facilities Test and Startup Branch 1.
Individuals contacted Duke Power Company (DPC)
i J. E. Smith - Plant Superintendent l
J. W. Hampton - Assistant Plant Superintendent R. M. Koehler - Technical Support Engineer O. S. Bradham - Instrument and Control Engineer M. D. McIntosh - Operating Engineer G. W. Cage - Assistant Operating Engineer J. M. Davis - Staff Maintenance Engineer
J. W. Cox - Assistant Pl ut Engineer 2.
Control Rod Drive Breaker Undervoltage Trip Assembly Deficiency The corrective actions described in the licensee's report regarding this design deficiency, dated February 23, 1973, were reviewed by the inspector.
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The undervoltage trip device has been replaced on the control rod
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i drive breakers. The resistance of the coils has been measured
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and found to be,between 24.1 and 25.2 ohms.
Spring tension on
the trip arm has been adjusted to be approximately 500 grams.
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l The field change to install a 5 amp fuse in each of the four reactor i
protection system channels has been completed. The fuse was in-l stalled to limit the current through the reactor trip module to
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the CRD system trip devices in case of a component failure.
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RO Rpt. No. 50-269/73-8 I-2 (b
The inspector had no further questions regarding the corrective measures taken to resolve this deficiency.
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3.
ROB 73,
" Faulty Overcurrent Trip Delay Device in Circuit Breakers
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for Engi_eered Safety Systems" The corrective measures for the DB-50 relays discussed in DPC's response, dated April 12, 1973, to Regulatory Operations Bulletin No. 73-1, " Faulty Overcurrent Trip Delay Device in Circuit Brakers for Engineered Safety Systems," were verified during this inspection.
The overcurrent trip devices on the six DB-50 relays in service at the Keowee Hydro Station have been replaced. Changeout was completed on July 17, 1973. Plans to annually inspect these breakers have been completed.
- The inspector had no further questions on this item.
4.
Abnormal Occurrence Regarding Reactor Building Spray Pump
The corrective actions as described in the licensee's report, dated May 18,197 3, regarding this abnormal occurrence were reviewed by the inspector. The following actions were noted by the inspector:
a.
OP 1102/01, " Controlling Procedure for Unit Startup" This procedure, dated May 18, 1973, has been revised to re-quire the shift supervisor to review the preheatup and pre-critical checklists before giving permission to begin reactor startup.
b.
Standing Order No.11 has been issued by the Operating Engineer on May 24, 1973. This instruction requires tagging (as per administrative procedure No. 2) any power supply which is racked out for any reason.
c.
Review of Operating Procedures
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All operating procedures have been reviewed to determine if any other safety related equipment should be tagged when disabled. This review was discussed by the Station Review Committee. It was concluded that the reactor building spray pumps were unique and no other safety related equipment would require tagging.
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RO Rpt. No. 50-269/73-8 I-3 (.
The inspector determined that DPC has implemented the corrective action as stated in the incident report, and has no further questions on this matter. This item is closed.
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Emergency Operating Procedures
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During a previous inspection, / a lack of eight emergency pro-
cedures was identified. These were:
a.
Loss of containment integrity.
b.
Loss of flux indication.
Inoperable control rods or inability to drive rods.
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High activity in reactor coolant system.
e.
Acts of nature (other than earthquake).
f.
Loss of feedwater.
g.
Abnormal releases of radioactivity.
h.
Irradiated fuel da= age.
These procedures have been prepared and have been reviewed by the inspector. With the addition of these procedures, the requirements of AEC Safety Guide 33, " Quality Assurance Program Requirements (Operations)," and ANS 3.2, " Standard for Administrative Controls for Nuclear Power Plants," are fulfilled. The inspector had no questions or comments on these procedures. This previously identified unresolved item, No. 73-1/7, is resolved.
6.
Administrative Procedure 10, " Station Modifications" The inspector reviewed Administrative Procedure 10, " Station Modifi-
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cations," and the revisions made to that procedure as of June 14, 1973, and did not observe any deficiencies in the procedure. This review was made to determine if the requirements of Criterion V of Appendix B to 10 CFR 50 were being complied with.
1/ RO Inspection Report No. 50-269/73-1, Details III, paragraph 2.
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RO Rpt. No. 50-269/73-8 I-4
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This matter was the subject of a violation which is discussed in the enforcement letter issued by R0:II on March 20, 1973, and responses from DPC on April 13, 1973, and June 15, 1973. The cor-
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rective actions described in DPC's correspondence were verified i
during this inspection and RO:II has no further questions.
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Nitrogen Inline Heater Desi;n Change Documentation During a previous inspection, / the QA documentation for the in-
line nitrogen heaters design change was requested by the inspector.
This information was provided by the licensee and reviewed by the inspector. This change originated as a Design Engineering Change Order No. 293, approved on October 16, 1972.
The modification was implemented in accordance with DPC QA procedures, which were in effect at the time the change was installed, and is in agreement with Criterion III of Appendix B to 10 CFR 50. This previously identified unresolved item, No. 73-3/1, is resolved.
8.
TP 600/21, " Hydrogen Addition Test"
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A question was raised i
setpoint for the regulated hydrogen pressure on the letdown storage tank. This item has been resolved. A wider range regulator has been provided and has been set at 35 psig. RO:II has no further question or comment.
9.
Active Electrical Jemper and Bypass Log During a previous inspection,3/ the inspector noticed that this
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record contained a nu=ber of jumper and bypass records that appeared to be no longer required. During this inspection, a revicw of this record revealed that jumpers and bypasses in effect appeared to be applicable for the current phase of operation. The inspector had no questions or comments on the
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active jumpers and bypasses.
1/ RO Inspection Report Nos. 50-269/73-1, Summary-of Details,Section V.b, and 50-269/73-3, Details I, paragraph 6.
2/ RO Inspectibn Report No. 50-269/73-2, Details I, paragraph 6.
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3/ RO Inspection Report No. 50-269/73-2, Details II, paragraph 4.
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RO Rpt. No. 50-269/73-8 I-5 10. Alarm Procedures During a previous inspection,1/ comments on alarm procedures were discussed with the licensee representatives. Each of
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these comments were discussed and resolved with the assistant
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plant superintendent and the operating engineer. The inspector had no further question on this item.
j 11. Periodic Test and Instrument Procedures i
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Comments on the following periodic test and instrument pro-f cedures were discussed yith the licensee's representative during a previous inspection.-.
Followup on these comments was com-pleted and the comments have been resolved as follows:
IP 201/1A, " Core Floed__ Tank Level Instrument" a.
The comment involved a check of the status of test equipment.
Signof f space for this item has been added to the procedure.
b.
IP 202/1D, " Emergency HPI Flow Instrument Calibration"
l The previous comment involved test sequence. This item has
been resolved. The individual modules are checked followed
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by an integrated test of the whole logic string.
IP 340/4, " Absolute Position Indicator Calibration"
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The reference to Figure 8.1 has been corrected. The figure t
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is now included in the procedure and the instructions referring to repeat steps have been cleared up, d.
PT 620/15, "Keowee Hydro Operational Test" Previously, the rated speed and voltage of the Keowee Hydro i
Plant were not specified in the test procedure. These values
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have been added.
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-1/ RO Inspection Report No. 50-269/73-1, Details III, paragraph 3.
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i 2/ RO Inspection Report No. 50-269/73-1, Details III, paragraph 4.
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R0 Rpt. No. 50-269/73-8 I-6 12. Maintenance Procedures A followup review on the maintenance procedure pggram was completed
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by the inspector. During a previous inspection,- comments were discussed and the licensee proposed a resolution. These items were
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reviewed and found satisfactory. The maintenance procedures were found to compare favorably with Safety Guide 33, " Quality Assurance Program Requirement (Operation)," and ANS 3.2, " Standard for Admin-istrative Controls for Nuclear Power Plants."
13. Test Equipment Control A followup review of the licensee's program for control of measuring a test equipment was completed by the inspector.
Previously, comments were discussed and the licensee proposed a resolution for these comments.
The inspector found the program to be in agreement with the requirements of Criterion XII, " Control of Measuring and Test Equipment," of Appendix 3 to 10 CFR 50, and with DPC's Admin-istrative Policy Manual for Operational Quality Assurance.
Section 2.3.
The inspector's coc:ments have been resolved.
14. Operating Procedures Several comments were discussep with the licensee's representative during a previous inspection.1 The operating procedures were rereviewed and the co= ment has been satisfactorily resolved. The inspector had no further com=ents on the operating procedures.
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1/ RO Inspectioh Report Nos. 50-269/73-1, Details I, parag:aph 7.g, and j
50-269/72-11, Details II, paragraph 3.
2/ RO Inspection Report Nos. 50-269/73-1, Details I, paragraph 7.g, and 50-269/72-11, paragraph 3.b.
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R0 Inspection Report Nos. 50-26ir/72-10, Details II, paragraph 9, and 50-269/72-11, Details II, paragraph 4.
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RO Report No. 50-269/73-8 II-1 ().
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N M '/-3/ 9.3 DETAILS II Prepared By:
e K. W. Whitt V
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Date Reactor Inspector
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Facilitnes Test and Startup Branch
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Dates of Inspection: July 17-20, 1973
- / 23 Reviewed By:f
C. E.. Murphy /Ch(fi'
'Dste Facilities Test and Startup Branch 1.
Individuals Contacted Duke Power Company (DPC)
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J. E. Smith - Plant Superintendent
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J. W. Hampton - Assistant Plant Superintendent M. D. McIntosh - Operating Engineer R. M. Koehler - Technical Support Engineer G. W. Cage - Assistant Operating Engineer C. L. Thames - Health Physics Supervisor O. S. Bradham - Instrument and Control Engineer D. Lanning - Maintenance Supervisor J. W. Cox - Assistant Plant Engineer L. E. Summerlin - Staff Engineer 2.
Oil Fire of March 6, 1973 The March 6, 1973, oil fire was initially reported in R0 Report No. 50-269/73-3, Details I, paragraph 5.
A report transmitted from DPC to the Directorate of Licensing entitled " March 6, 1973, 1Al Reactor Coolant Pump 011 Fire Incident Report," and dated May 4, 1973, was reviewed by the inspector. This report described the events leading to the fire, actions taken to control and extinguish the fire, and the cleanup and repair effort. Action
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taken to prevent recurrence of si=ilar incidents was reviewed on site by the inspector.
This action consisted of (1) preparation and approval of a procedure for filling and draining the reactor coolant pump (RPC) motor upper and lower bearing reservoirs, (2) an intrastation letter to all operations supervisors which provides guidelines and requirements for use of procedures in the performance of operations, and (3) a standing order which provides instructions on marking and s6tting recorder charts.
The procedure for filling and (
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RG Rpt. No. 50-269/73-8 II-2
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I (raining the RCP motor upper and lower bearing reservoirs was added to r
I the RPC operation proceduta (OP/1/A/1103/06) as section 6.0. subsection 6.1,
" Initial Conditions," which requires, among other things, that portable CO
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l fire extinguishers be available at the RCP to be filled or drained.
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This item is considered closed.
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Failure of CF-1 to Fully 0 _en
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I This item was initially reported in RO Report No. 50-269/73-4,
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Details I, paragraph 13.
The licensee has submitted a written
report to the Directorate of Licensing concerning this occurrence.
The report, dated May 4,1973 describes the occurrence, the immediate corrective action, and certain investigative action. The internal limit switches remain unchanged. Ihese switches illuminate lights in the control room to indicate the position of the valve.
In addition to and separate from these switches is
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the external limit switch, which has been added to each core flood
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i motor operated valve. This switch is tripped by an extension of
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the valve stem and activates an alarm in the control room when the valve is fully open. The circuits of the internal and external l
switches are completely separate.
The locknut on this valve and all
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other limitorgue operated valves in the engineered safeguards system have been inspected and all loose nuts retightened and properly secured.
This item is considered closed.
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Failure of Reactor Building Spray Valves, BS-1 and BS-2, to Open
During Engineered Safeguards System Testing
This item was initially reported in RO Report No. 50-269/73-4, j
Details I, paragraph 14.
A written report outlining the
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circumstances of the event and the corrective action was transmitted
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to the Directorate of Licensing by a licensee letter dated i
May 4, 1973. The inspector has reviewed the report, and during this inspection, he inspected the action taken to prevent recurrence.
Signs have been affixed to the manual hand wheels of BS-1 and BS-2 which state, "Do not jam valve closed by hand."
However, no
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revisions had been made to any operating procedures to require j
testing of any engineered safeguards valve electrically after it
has been manually operated.
This was part of the action specified
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in the licensee report to prevent recurrence of a similar problem.
This item will remain open until the specified action is completed.
A licensee representative estimates the necessary procedure changes i
will be completed by September 1, 1973.
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do Rpt. No. 50-269/73-8 II-3
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Comments on Previousiv Reviewed Test Procedures A check was made by the inspector to verify that comments for which
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commitments had been made were being incorporated into the procedures,
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and otherwise performed as agreed.
The findings were as follows:
a.
TP-110/2, " Penetration Room Flow Pressure Drop and Filter Tests" In RO Report No. 50-269/73-5, Details I, paragraph 7, it was reported that the only item remaining from the preoperational testing program was the repeat of TP-110/2, which would be rerun at 1000 cfm.
During this inspection, the inspector verified that the test had been rerun at the specified flow rate on March 8, 1973. This item is considered closed.
b.
TP-200/22, " Reactor Vessel Internals Inspection" Three comments were reported in R0 Report No. 50-269/73-11, Details I, paragraph 7.
The resolution of these comments was as follows:
(1) Comment a - A definition of " class A cleanliness" was provided in Section 3.5.5.2.1 of the Administrative Policy Manual for Operational Quality Assurance.
Therefore, it was not necessary to delete or change the referenced precaution.
(2) Comment b - All discrepant items were noted and recorded on enclosures to the procedure and evaluated by Babcock and Wilcox Company (B&W) and DPC personnel.
This evaluation was then used as acceptance criteria.
(3) Comment c - There is a row of bolts that connects the thermal shield to the lower grid and another row that connects the flow distributor to the lower grid. Both rows of bolts were inspected. The thermal shield upper restraint block attachments were also ' inspected.
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Ihis item is considered closed.
c.
TP 330/3A, " Control Rod Drive Test To Determine Rod Drop Times" In R0 Report No. 50-269/72-11, Details II, paragraph 11, the inspector commented that'DPC managenent had agreed to drop and time test all control rods after initial core loading and to drop and time test the fastest and slowest rods an additional 25 times. This' testing was to be in addition to the rad testing
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, 3 R0 Rpt. No. 50-269/73-8 II-4 done during hot functional testing per TP-330/3A.
The inspector verified during this inspection that all rods had been dropped and time tested following initial core leading and that the
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fastest and slowest rods had been teste'd at least 25 additional times per 1P/0/A/330/3A, " Control Rod Drive Test to Determine Rod Drop Times." This item is considered closed.
d.
TP-800/18, " Induced Power Oscillation Test" The inspector noted that enclosure 13.9, " Power Imbalance Plot," had been added to the procedure to provide instructions on how to determine that oscillations were within the required !3%.
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addition was made in response to a comment recorded in R0 Report No. 50-269/73-5, Details I, paragraph 6.
This item is considered closed.
TP-800/22, " Thermal Measurements For IB OT SG Fluids" e.
R0 Report No. 50-269/72-11, Details II, paragraph 5, states that this procedure did not contain acceptance criteria but that the licensee had indicated that an appropriate statement of acceptance criteria would be added.
During this inspection, the inspector noted that a statement of acceptance criteria had been added to the procedure. This item is considered closed.
f.
TP-800/25, "NSS Heat Balance" Acceptance criteria has been 'added to this procedure as a result of the comment recorded in R0 Report No. 50-269/73-5, Details I, paragraph 6.
This item is considered closed.
6.
Sensitivity of Leak Detection Systems The inspector discussed the sensitivity _of the leak detection systems with licensee management. Data from the reactor building air particulate monitor and the total reactor coolant inventory was
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reviewed. The iodine monitor and the gaseous munitor were not considered because of their recognized low sensitivity. No data was available for review from the reactor building normal sump. The health physics supervisor stated that he felt that sufficient steady state reactor operating time had not been accomplished to sufficiently evaluate the reactor building air part,1culate mgnitor operation in detection of leaks. The inspector was unable to make a meaningful correlation between the radioactivity reading and the reactor coolant leakage.
The reactor coolant is I
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RO Rpt. No. 50-269/73-8 II-5 inventoried once per day. A computer program is used by which the volumes of coolant in the pressurizer, the quench tank, and let down storage tank are determined. These volumes are printed out
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at a specified time and again at the end of one hour. The difference in total volume at the start and finish of the hour is calculated
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as the total reactor coolant system leakage.
The computer prints out the amount of coolant I each of the inventoried tanks to the thousandth of a pound and then converts this to gallons. This system appears to provide a high degree of sensitivity.
If the computer is inoperable, the coolant leakage is determined manually in i
accordance with an operating procedure.
It appears that this manual calculation will detect a leak of 0.5 gpm conservatively and probably as little as 0.25 gpm.
During the management interview, the inspector asked why readings of the reactor building normal sump were not being recorded. A management representative stated I
that the inventory method being performed by the computer was l
believed to be satisfactory, but that this along with other aspects of the leak detection system would be further evaluated.
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Technical Specifications surveillance Requirements i
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The inspector discussed this matter with licensee management. He explained that there might be some surveillance requirements in
the Technical Specifications for which relief from the stated i
frequency would be desired by the licensee during reactor shutdown.
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l It was emphasized that the reactor being in the shutdown condition i
does not relieve the licensee of the responsibility of conducting
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surveillance tests in accordance with the schedule specified by the technical specifications unless specific relief is authorized
'
by the technical specifications. A licensee representative stated
-
that he thought a number of technical specification changes would be desirable. He said that a review was being made to determine
,j what changes would be requested.
He stated that any relief found
)
to be necessary would be incorporated into the technical specifications
',
for Unit 2 prior to receivic ; a license for that unit, and that these changes would be reflected in the Unit 1 technical specifications.
'
at the same time. No further inspection effort is planned concerning i
this item.
l.
8.
Staking of Limitorque Valve Locknuts A licensee management representative stated that DPC had written the
,
manufacturer of the limitorque operated valves in the engineered
safeguards system in an effort to determine whether there is a j
more positive method that can be use(. for staking the locknuts.
A reply is expected within a few days according to the licensee representative.
i j
l i
i
.