IR 05000346/1979021
| ML19208D631 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 08/09/1979 |
| From: | Baker K, Heishman R, Streeter J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML19208D632 | List: |
| References | |
| 50-346-79-21, NUDOCS 7909290126 | |
| Download: ML19208D631 (12) | |
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U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT
REGION III
Report No. 50-346/79-21 Docket No. 50-346 License No. NPF-3 Licensee: Toledo Edison Company Edison Plaza 300 Madison Avenue Toledo, OH 43652 Facility Name: Davis-Besse Nuclear Power Station, Unit 1 Inspection At: Davis-Besse Site, Oak Harbor, OH and Toledo OH Inspection Conducted: June 4-5, 20-21, and July 16-20, 1979 Inspectors:
8!9[79
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- d K. R. Baker (July 16-20, 1979)
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ly-20, 1979)
O!9[~7 9 lb &
L. G. McGregor (July 16-20, 1979)
8 /9 [79
/f Approved By: R. F. Heis an, Chief il 9 j 77 Reactor Operations and Nuclear Support Branch Inspection Summary Inspection on June 4-5, 20-21, and July 16-20. J979 (Report No. 50-346/
79-21)
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Arcas Inspected: Routine, announced inspection of licensee action on previous unresolved items and items of noncompliance; operating logs; battery electrolyte surveillance; use of computer in core thermal margin 1055 121
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determination; RCS pressure-temperature limits; status of power ascension test program; startup test results; Auxiliary Shutdown Pat. ' instrumenta-tion and controls. The inspection involved 114.5 inspector-hours on site by four NRC inspectors.
Results: No items of noncompliance or deviations were identified.
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.i DETAILS 1.
Persons Contacted B. Beyer, Assistant Station Superintendent R. Chesko, Assistant Operating Engineer C. Domeck, Project Engineer A. Horvath, Electrical Foreman S. Jain, Engineer
- D. Lee, Test Program Manager J. Lehnert, Group Foreman tJ. Lingenfelter, Nuclear and Performance Engineer L. Miller, Operations Engineer F. Miller, Senior Engineer F. Moss, Electrician
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Murray, Station Superintendent V. Opfer, Shift Foreman R. Smith, Computer Engineer
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J. Zell, Test Leader The inspectors also interviewed other licensee employees, including members of the administrative, technical, and operations staff.
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2.
Licensee Actions on Previous Inspection Findings (Closed) Inspector Followup Item (Reference IE Inspection Reports No.
50-346/78-17, Paragraph 4, and No. 50-346/79-04, Paragraph 2): Power Engineering review of the results of TP 401.01. As of July 19, 1979, the comments of Power Engineering had been resolved.
(Closed) Unresolved Item (346/78-30-03.): Technical soecification change. The licensee submitted to NRR on March 23, 19: 3, the tech-nical specification change concerning bypassing undervoltage protection when starting Reactor Coolant Pumps or Circulating Water Pumps.
(Closed) Unresolved Item (346/79-04-02): NRR review of the process computer software. The licensee determined that the process computer software had not been reviewed by NRR. The licensee stated that the functir _ computer requirements for analyzing input data from core instrut.entation were described in B&W Specification CS-3-90/NSS-14.
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Prior co initial operations this specification was sent by B&W to Bailey Meter Company who developed the computer software to imple-ment the specification.
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l (Closed) Noncompliance (34 6-79-04-01): Failure to follow procedures.
The inspector veriffud in discussions with involved personnel that the importance of folJswing pr)cedures had been emphasized to them. The licensee issued revision 2 of procedure PT 5175,00, " Differential Rod Worths Measurements at Powec", on March 3, 1979, to allow the Test Leader the flexibility to determine the stabilization of parameters prior to measurements.
(Closed) Noncompliance (346/78-16; Reference IE Inspection Reports No. 50-346/78-17, Paragraph 6, and No. 50-346/78-30, Paragraph 2):
Capability of incore detectors to determine core thermal conditions.
The licensee issued Revision 3 of procedure ST 5033.01, "Incore Instrumentation Channel Check", on May 16, 1979, which established improved criteria for determining incore detector failures and for establishing conservative replacemaat values for failed detectors.
(Closed) Unresolved Item (346/79-04-04); Possible positive moderator tempercture coefficient above 95% rated thermal power. The inspector thcught there was a possibility of a positive temperature coefficient above 95% rated thermal power based upon changes in nuclear instrument indications which accompanied Tave changes during the performance of TP 800.05, " Reactivity Coefficients at Power".
However, more accurate heat balance determinations during the same interval indicated essen-tially stable power during the Tave changes. The nuclear instrument variations are attributed to a "downcomer effect" which result in changes in shielding of the nuclear instruments as the water (T-cold)
in the reactor vessel c'.owncomer region changes. The results is a tL direct variation of ir licated power with T cold. The inspector con-cluded that the licentee had determined that the moderator coefficient was negative above 90 power and was in compliance with TS 3.1.1.3.
(Closed) Unresolved Item (346/79-04-03): Compliance with commit-ments to NRR concerning testing to be performed at 100% power. In a letter to NRR dateo June 8, 1978, the licensee committed to con-ducting TP 800.05 "R. activity Coefficients at Power", at approxi-mately 100% power fcilowing startup from the Summer 1978 outage.
The testing was actually performed at 90% power.
The licensee viewed 90% to be "approximately" 100% and in compliance with his commit-ment for TP 800.05.
Since TP 800.20 was only intended to be used in conjunction with other tests such as TP 800.05 and never by itself, the licensee believed he had met the intent of the commitment for TP 800.20. The licensee performed a safety evaluation of this matter and concluded there were no safety issues. Although the inspector stated he believed 90% power to be a liberal interpretation of the
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words "approximately 100% full power", he concurred that the intent
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of the NRR commitments had been satisfied and that testing at the
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90% power level did not involve safety issues.
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(Closed) Unresolved Item (346/78-14; Inspection Report Nos.
50-346/78-17, Paragraph 2, and 50-346/78-30, Paragraph 2): Stopping HPI pumps following a LOCA. The licensee issued Revision 3 of pro-cedure EP 1202.06, " Loss of Reactor Coolant and Reactor Coolant Pressure", on January 6,1979, to amplify restrictions against securing HPI pumps following a LOCA. The revised procedure required a minfmum decay heat flow be established or the leak be isolated
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before securing the HPI pumps, even if pressurizer level is re-established before the requirements are met.
These requirements have since been replaced by other restrictions approved by NRR for all PWR facilities.
(Closed) Unresolved Item (346/78-30-04): Core lift analysis.
The licensee received a letter dated February 13, 1979, from the reactor
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vendor which indicated that the core lifting problem had been reviewed
in light of the burnable poison rod removal and that the net holddown force had increased due to the poison rod removal.
The letter also indicated that the potential for core lifting below 500 F was con-sisgent with FSAR Section 4.4.2.7 because four pump operation below 500 F is not considered a normal mode of operation and'is not allowed by the technical specifications. The inspector determined that the core lifting issue had been previously reviewed by NRR for another plant and that review appeared generically applicable to all B&W plants.
The licensee plans to consider placing the temperature ingerlock which prevents starting a fourth reactor coolant pump below 500 F on a refueling interval surveillance schedule.
(Closed) Unresolved Items (346/78-30-01 and 346/78-13): Need for s-revised small break analysis. Prior to the TMI-2 event of March 29, 1979, NRR had determined that no additional information was required from TECO concerning the licensee's rationale for not revising the small break analysis to remain in compliance with 10CFR 50.46 (K).
Subsequent to the THI-2 event, certain information was submitted to and reviewed by NRR concerning small breaks. The information on small breaks required by NRR is described in Enclosure 1 of the letter from H. R. Denton to L. E. Roe dated July 6, 1979.
(0 pen) Noncompliance (346/78-15), Inspection Report No. 50-346/78-17, Paragraph 2, Infraction 2.c: CNRB review of power escalation test re ts.
The CNRB had not reviewed any test results since the initial CNRB review on September 22, 1978. However, the CNRB noted in the minutes of Meeting #33 on February 23, 1979, that'the committee had more results to review. As stated in the September 12, 1978, letter from the licensee to.RIII, CNRB review of test results will be completed"......within 30 days of the completion of the Power Escalation Sequence Testing and resolution of associated test defi-
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ciencies" " Completion of the Power Escalation Sequence Testing is scheduled for August 7, 1979, and resolution of any associated test deficiencies should occur approximately August 10, 1979. Therefore, the CNRB reviews should be completed approximately September 9,1979.
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(Open) Unresolved Item (Inspection Report Nos. 50-346/78-27, Paragraph 2, 50-346/78-30, Paragraph 2, and 50-346/79-04, Para-graph 2): High pressure injection system parformance during September 24, 1977, event. The licensee had completed an analysis to show that the delay in clearing the low flow alarm on leg 2-2 was not unanticipated. The licensees review of a April 29, 1978, event for comparison purposes yielded inconclusive results. The inspector reviewed the licensee's analysis and instrument cali-bration records and stated that more information was necessary to support the licensee's ennelusion that the delay of leg 2-2
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did not indicate inoperability of the HPI system. Specifically, the inspector stated that the maximum instrument error of the low flow alarms should be used in conjunction with the time the
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first leg (2-1) cleared, the RCS pressure decay rate, the HPI
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line loss curves, the HPI pump shutoff heads, and the ECCS flow assumptions. The licensee agreed to continue his evaluation of this matter.
3.
Operating Logs Inspection Report 50-346/78-17, Paragraph 5.b.,
indicated that the licensee was reviewing the need for improving his documen-tation in logs of inoperable safety-related equipment. The licensee stated during the present inspection that his review indicated log entires should be'more comprehensive to assure that the logs accurately and completely reflect major equip-
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ment (including all safety-related equipment) status changes.
The licensee plans to issue revised guidelines regarding the upkeep of logs. The inspector reviewed procedure AD 1839.00,
" Station Operations", Revision 5, and the Unit and Reactor Operator's Logs and also determined improvements were desirable in the quality of log entries. This is an Unresolved Item (346/79-21-01) pending the licensee implementing improved lor.-
keeping practices.
4.
Battery Electrolyte Surveillance The inspector reviewed surveillance records for station batteries IN, 2N, and 2P for the period May 2-July 4, 1979, to determine if the electrolyte specific gravity measurements were corrected to full electrolyte level per TS 4.8.2.2.a.2.
The inspector verified that a correction factor of 0.0015 was being applied for each 1/16" deviation from the full mark. The inspector did not identify
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any items of noncompliance or deviations and had no further ques-tions concerning this matter.
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5.
Core Thermal Margin Determination in Process Computer The inspector reviewed with the Computer Engineer the software relating to core thermal margin determination in Bailey Meter 855/50.
The inspector notad that all the software was initially programmed by B&W, and the licensee was responsible for making occasional software changes when notified by B&W. The inspector examined and verified software Change Notification, SCN-76-01, dated June 22, 1978.
The inspector noted that the licensee was not keeping computer printouts which verified that software changes were made and that the changes were correct. The licensee stated he would begin keeping those printouts for record purposes.
No items of noncompliance or deviations were identified.
6.
RCS Pressure-Temperature Limits Emergency Procedure EP 1202.06, " Loss of React.or Coolant and Reactor Coolant Pressure", contains Figure 2B, " Emergency Pressure / Temperature Limit Curve Without Forced or Natural kCS Circu'. tion".
The Technical Specifications contain Figure 3.4-3,
"Reaucor Coolant System Pressure-Temparature Limits for Cooldown."
Figure 2B establishes allowable combinations of RCS pressure and core outlet temperature whereas TS Figure 3.4-3 establishes s.
allowablecombgnationsofRCSpressureandTc_. Actual Tc may be as much as 150 F lower than core outlet temperature under emer-
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gency/ faulted RCS conditions where there is no forced or natural RCS circulation and there is high pressure injection and/or makeup addition. With the large difference, use of EP 1202.06 Figure 2B could result in exceeding tha limits of TS Figure 3.4-3.
NRR reviewed EP 1202.06 prior to lifting the shutdovn order of May 16, 1979, and concluded that the TS Figure 3.4-3 limits were developed for normal and upset conditions and are not applicable to a depressurization and cooldown following a LOCA. NRR concurred that core outlet thermocouples could be used by the operator to control the degree of subcooling in the RCS per EP 1202.06 Figure 2B.
The NRR positions are documented on Pages 21 and 22 of the NRR safety evaluation supporting the decision to lift the shut-down order.
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On July 13, 1979, the licensee submitted to NRR a proposed change to the technical specifications which would make it clear that TS Figure 3.4-3 is not applicable to emergency / faulted conditions such as a LOCA and to add a new figure which would be applicable to those conditions. The proposed new figure is consistent with EP1202.06 Figure 2B.
In view of the above actions by the licensee and NRR, the inspector stated that use of EP1202.06 Figure 2B during emergency / faulted RCS conditions as described in the procedure would not be considered te be in noncompliance with TS Figure 3.4-3.
7.
Control Room Visitors On several occasions during the inspection, the inspector observed as many as ten people in the operator area of the Control Room.
It appeared that many operating personnel tended to congregate for lunch in the Control Room. The inspectors stated that unnecessary person-nel should fird other places to congregate for lunch and other such activites. The inspectors stated that during an emergency such unnecessary personnel in the Control Room could be a distraction to operators and could physically delay operator actions. The licensee acknowledged the inspector remarks and stated he intended to issue a memorandum by August 10, 1979, to give guidance on limitation of Control Room access. This is an Unresolved Item (346/79-21-02) pend-ing the licensee's implementation of improved Control Room access
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controls.
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Status of Power Ascension Test Program The licensee had not yet completed at - power transient testing. The transient testing remaining to be completed consist of a load swing from 90% to 40% to 90% power with four reactor coolant pumps (RCPs)
operating and a load swing from 60% to 30% to 60% power with three RCPs operating. These swings will be accomplished in accordance with TP 800.23, " Unit Load Transient Test".
Data will also be taken during the swings to complete TP 800.08 "ICS Tuning at Power". The licensee stated that the testing would be completed by August 7,1979, and the licensee review of the test results would be completed by August 10, 1979.
In addition to TP 830.08 and TP 800.23 testing, the licensee has not yet completed the 100% power portion of the following non-safety
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related tests:
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TP 272.01, " Main Feed Pumps and Turbine Acceptance Test"
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TP 280.02. " Turbine-Generator Acceptance Test" TP 291.01. " Excitation System Acceptance Test" These tests are scheduled to be completed August 7,1979.
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The licensee stated that completion of the five tests mentioned above
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will constitute completion of the Power Escalation Sequence Testing.
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9.
Reaccor Trip Test L
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The licensee committed on FSAR page 14-108 to measure unit response i
during cnd af ter a deliberate reactor cg; turbine trip from power.
(The reactor / turbine trip circuitry as recently modified provides for an immediate reactor trip following a turbine trip as well as the pre-vious provision for an immediate turbine trip following a reactor tr.f o.)
Although not deli 5erate, the licensee obtained enough information foi-
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lowing the January 12 trip to demonstrate that all acceptance criteria for the reactor trip test (TP 800J14) were met.
The licensee completed a safety evaluation (" Safety Evaluation for FCR 79-258") in which it
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was concluded that the January 12, 1979, reactor trip could be used to satisfy the FSAR commitment on page 14-108. The inspector concurred that the January 12, 1979, reactor trip satisfied the licensee commit-
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ment.
10. Auxiliary Shutdown Panel Instrumentation and Controls The inspeccers examined the Auxiliary Shutdown Panel and noted that
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several indicating lights were burned out and many control switches
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l were labeled " SPAR 2".
The inspectors stated that during an emergency situction these conditions could be confusing to operators. The licensee took prompt action to replace the burned out bulbs and 'to
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place a supply of bulbs near the panel. The licensee also verified that the extra control switches were correctly labeled " SPARE".
The licensee is thinking about removing the handles of the spare switches a
and placing covers over others. This is an Unresolved Item (346/79-21-03) pending completion of licensee actions.
i 11. Reactor Coolant Pumo Starting Interlock on Low Component Cooling Wcter Flow Inspection Report 30-345/78-17, Paragraph 5.b, indicated that diffi-culty was frequently experienced with starting RCPs because of a
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45 gpm CCW flow interlock. The report also indicated that the inspec-
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tors asked the licensee to evaluate the advisability of having to send a man inside containment to verify the low CCW flow switch pickup-9_
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prior to starting a RCP in light of meeting T.i requirements. This se because of the possibility of containment dnaccessibility or delays in RCP starting due to a containment entry prerequisite.
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licensee stated that it was routine practice to send a man in contain-
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ment to witness the starting of RCPs when starting up from an outage.
Ecwever, the licensee also stated that it was not necessary to enter
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containment to verify the pickup of the low CCW flow switch since
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this can be done by reviewing the computer printout in the Control
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Room. Additionally, the licensee stated that the low CCW flow switch pickup could be verified at relay cabinets outside ".he Control Room and containment and could be bypassed if it was necessary.
The inspec-tor stated that it would be helpful to operators if a statement was placed in the RCP starting procedure to remind operators of the CCW
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interlock difficulty. The licensee acknowledged the inspector's remark. The inspector had no further questions concerning this matter.
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Inspection of Test Results a.
Test Results Reviewed TP 800.05, " Reactivity Coefficients at Power" (testing at 90%)
TP 800.13, " Unit Load Rej 'ction Test" (testing at 100%)
TP 800.20, " Rod Worth Measurements" (testing at 90%)
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TP 800.25, " Shutdown from Outside the Control Room" TP 800.26, " Loss of External Load Including Offsite Power Test" e
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Purpose of Review
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(1) Verify test changes were properly approved, annotated in procedure, completed, and did not change test objectives.
(2) Verify test deficiencies were properly documented and resolved.
(3) Verify cognizant engineering function had evaluated test results against acceptance criteria.
(4) Verify data sheets were complete and appropriate signatures
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or initials were contained in the "as-run" copy of the procedure.
(5) Verify documentation of QA involvement in testing.
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(6) Verify that the test results had been approved.
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Findings The inspector did not identify any items of noncompliance or deviations and judged the test results to be acceptable. List-ed below are some specific inspector findings.
(1) TP 800.25 - Reference Inspection Report 50-346/78-30, Para-graph 8.
(a) The referenced report states "However, if emergency procedures are modified to include the actions taken there appears to be no impact on the acceptability of the test."
The inspector found that the emergency procedure (EP 1202.33) contained instructions in Step 4.7.9 to ~ anually control the Atmospheric Steam m
Vents. The inspector had no further questions con-cerninc this matter.
(b) The referenced report states " Based on the requirement that the reactor can be shutdown and maintained in hot standby outside the control room and that there was communication from the control room regarding the makeup tank level the test results were unacceptable.
Further review by the Project Inspector is proceeding to determine if the test is acceptable. Computer logs and data associated with the test are under further
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review." The inspector determined that the cocimunica-tion from the Control Room regarding makeup tank level during the test didnot invalidate the test results.
However, the communication did indicate a need to revise the emergency procedure to assure that the test results and the revised procedure (Revision 7 of EP 1202.33, dated June 25, 1979) assure that the plant can be shut-down and maintained in hot standby condition from outside the Control Room.
The inspector also reviewed computer logs and data associated with the test and found that there were no indications of unacceptable equipment performance during the test.
(2) TP 800.05 and TP 800.20= Reference Inspection Report 50-346/
79-04, Paragraph 4.b.
All Power Engineering comments had
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been satisfactorily resolved.
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13.
Unresolved Items
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Unresolved Items are matters about which more information is re-
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quired to ascertain whether they are acceptable items, Items of Noncompliance, or Deviations. Unresolved Items disclosed during the inspection are discussed in Paragraphs 3, 7, and 10.
14.
Exit Interview The insepctors met with licensee representatives (denoted in Paragraph 1) at the conclusion of the inspection on July 20, 1979. The inspec-tors summarized the purpose and the scope of the inspection and the findings.
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