IR 05000346/1979028
| ML19262A579 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 10/25/1979 |
| From: | Baker K, Heishman R, Streeter J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML19262A574 | List: |
| References | |
| 50-346-79-28, NUDOCS 7912070079 | |
| Download: ML19262A579 (12) | |
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U.S. NUCLEAR REGULATORY COMMISSION
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OFFICE OF INSPECTION AND ENFORCEMENT
REGION III
Report No. 50-346/79-28 Docket No. 50-346 License No. NPF-3
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Licensee: Toledo Edison Company Edison Plaza 300 Madison Avenue Toledo, OH 4365'
acility Name: Davis-Bessa Nuclear Power Station, Unit 1 Inspection At-Toledo and Oak Harbor, OH Inspection Conducted: August 13-17, 27-31, September 10-14 and 24-28, 1979 PMbDP
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Inspectors:
J.\\ F. Streete %
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r (August 13, 14, September 24 and 26-28, 1979)
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Approved By:
. Heishman, Chief
Reactor Operations and Nuclear Support Branch Inspection Summary Inspection on August 13-17, 27-31, September 10-14 and 24-28, 1979 (Report No. 50-346/79-28)
Areas Inspected:
Routine, announced inspection of previous unresolved items and items of noncompliance; startup test results review; startup test program status; onsite review committee; organization and administration; records and document control; steam generator tube leak procedure; licensee action to improve quality and timeliness of the resolution of identified concerns. The inspection involved 168 inspector-hours onsite by two NRC inspectors.
Results: Of the eight areas inspected, no items of noncompliance were iden-tified in six areas. Two items of noncompliance were identified in two areas (Infraction - disabling of undervoltage sensing relays - Paragraph 2; Infrac-tion - failure of SRB to review violations of technical specifications -
Paragraph 6).
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DETAILS
1.
Persons Contacted B. Beyer, Assistant Station Superintendent P. Carr, Maintenance Engineer R. Crouse, Vice President, Energy Supply s
- C. Daft, Operations QA Manager
- B. Hill, Engineer (B&W)
S. Jain, Engineer AD. Lee, Test Program Manager (B&W)
D. Miller, Operations Engineer F. Miller, Senior Engineer
- T. Murray, Station Superintendent E. Novak, Power Engineering Superintendent S. Quinnoz, Technical Engineer L. Roe, Chairman, Company Nuclear Review Board L. Simon, Operations Supervisor The inspectors also interviewed other licensee employees, including members of the administrative, technical, and operations staff.
- Denotes those attending the exit interview.
The NRC Pesident Inspector, L. Reyes, was also in attendance.
2.
Licensee Action on Previous Inspection Findings (Closed) Unresolved Item (346/78-12; Reference IE Inspection Reports No. 50-346/78-27, Paragraph 2, and No. 346/78-30, Paragraph 2):
HPI performance during September 24, 1977, event.
The licensee concluded that the probable cause of the delay in the clearing of the low flow alarm for HPI Leg 2-2 during the event was a combination of instrument error and a sticking stop-check valve (HP 56). However, there is no data to conclusively establish whether or not the delay was real and, if real, to establish the causes.
The licensee performed a bounding safety analysis which assumed the indicated delay to be a real delay. The purpose of the analysis was to determine if a real delay of the magnitude experienced would be indicative of the HPI system not performing as assumed in the small break ECCS analysis. The analysis compared the flow versus pressure experienced during the event with the flow versus pressure assumptions made in the small break ECCS analysis. The licensee determined that there could be a short period of time (several seconds) when the combination of the HPI Legs 2-1 and 2-2 were delivering a lower flow rate than assumed in the small break ECCS analysis before the stop-check in Leg 2-2 opened. However, the licensee also determined that the integrated quantity (gallons) of water injected through the 1522 188-2_
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combination of HPI Legs 2-1 and 2-2 always exceeded the integrated
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quantity assumed in the ECCS analysis. This was because the HPI system delivers flow much earlier (at a higher pressure) than was assumed in the ECCS analysis.
The licensee concluded from his review that the HPI system never experienced degraded performance outside the ECCS analysis. The licensee's review and conclusions were reviewed and concurred with by the Company Nuclear Review Board and the NSSS vendor (Babcock and
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Wilcox).
In addition to the above licensee analysis, the licensee also reviewed the results of HPI system periodic surveillances and preoperational testing and the results of an April 29, 1978, safety injection event.
The test results indicated satisfactory HPI system performance.
The data from the April 29, 1978, event yielded inconclusive information.
(Closed) Unresolved Item (346/79-21-02):
Control Room visitors.
The Station Superintendent issued Standing Order No. 27 to all plant personnel on August 9, 1979, which restricted Control Room visitors to those personnel having specific job-related business.
Inspector observations and discussions with Control Room operators indicated that Standing Order No. 27 was effective in minimizing unnecessary traffic.
(Closed) Unresolved Item (346/78-30-07):
Possible operation at power on auxiliary feedwater system. This matter related to a concern that, assaming a spurious SFRCS actuation and no operator action, the reactor would not be automatically tripped and would continue to operate on the auxiliary feedwater system.
The postu-lated scenario was:
(1) the reactor was operating at some power level greater than the heat removal capability (** 9%) of the auxiliary feedwater system, (2) a spurious SFRCS actuation occurred, (3) no operator action, (4) RCS pressure and temperature would begin to rise as occurred during the Scptember 24, 1977, event due to mis-match between energy produced by the reactor and energy removed by the auxiliary feedwater system and main steam safety valves, (5)
actuation of the power operated relief valve which would prevent a reactor trip on high reactor pressure, and (6) equilibrium RCS conditions reached where the PORV would be cycling about its set-point.
This scenario is no longer possible since the PORV setpoint was changed such that it is above the high pressure reactor trip setpoint.
The inspector examined all modes ot plant operation to determine if there were any other conditions where, in the absence of operator action and assuming an SFRCS signal, the plant would continue to operate on auxiliary feedwater. The inspectors concluded that the 1522 189-3-
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new PORV setpoint or the new anticipatory trip circuitry would pro-
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vide an automatic trip under all situations with the exception of one case. The exception was when the plant was operating with the electric startup feedwater pump.
The startup feedwater pump is normally only used during startup and shutdown prior to putting the steam driven feed pumps in service and it has a capacity to support reactor operation at a-2% power.
If the plant was operating at less than 2% power on the startup feedwater pump and a spurious SFRCS occurred, auxiliary feedwater would be initiated. This would probably
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result in a rapid cooldown of the RCS and a reactor trip on low RCS pressure. However, if a low pressure trip did not occur, continued operation in that condition did not appear to represent a safety concern.
There is only one condition as described above where unintentional continued operation on the auxiliary feedwater system could occur.
It is unrealistic to assume the plant would be in that situation and have a spurious SFRCS actuation and not have the operator trip the reactor upon SFRCS as required by procedures. However, even if those assumptions are made, the inspectors concluded that possible operation on the auxiliary feedwater system did not represent a safety concern.
(Closed) Unresolved Item (346/79-13-06):
Safety tagging procedure AD1803.00, Rev. 8.
The inspector reviewed the approved changes made to this procedure on August 11, 1979. The procedure establishes formal control on the removal of equipment from service by removing fuses etc. prior to racking breakers out.
(Closed) Unresolved Item (346/78-30-03):
Disabling of undervoltage relays to prevent tripping of 4160 VAC essential buses.
The licensee submitted to NRR on March 23, 1979, a proposed change to the tech-nical specifications. Until the change is reviewed by NRR, he licensee will continue to use the interim action described in IE Inspection Report No. 50-346/78-30, Paragraph 5, Page 9.
The inspector determined that the licensee's past actions in disabling the relays constituted noncompliance with TS 3.3.2.1.
Item 1 of the Notice of Violation for this report describes this matter.
However, since the licensee has taken corrective action to prevent recurrence, a response to that item is not required.
(Closed) Noncompliance (346/78-15; Reference IE Inspection Reports No. 50 346/78-17, Paragraph 2 and No. 50-346/78-30, Paragraph 2):
CNRB review of Startup Test Results. The CNRB had completed its review of nuclear safety-related startup test results on September 12, 1979.
There were no unresolved CNRB comments which affected the CNRB acceptance of the test results.
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(Closed) Unresolved Item (Reference IE Inspection Reports No.
50-346/78-06, Paragraph 4, No. 50-346/78-26, Paragraph 5, and No.
50-346/79-04, Paragraph 2): OTSG thermal hydraulic occillations.
The plant has a thermal-hydraulic oscillatory condition which origi-nates in the once-through steam generator (OSTG).
This condition manifests itself as a 0.25 HZ + 10 to 15 psi oscillation in steam pressure which occurs most markedly between 40 to 75% power. Feed-back mechanisms result in similar oscillations in feedwater flow,
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steam flow, electrical power, reactor core power, and reactor coolant temperature. Analysis and observations by the NSSS vendor at other similar plants show that +'
.nagnitude of the core power oscillation increases toward end of core life due to the increase in tae absolute value of negative temperature coefficient. The NSSS vendor has fo"au on other reactors that adjusting moveable steam generator orifice plates can add additional damping and reduce the magnitude of the oscillations. This has been recommended to the licensee by the vendor.
The NSSS vendor has reviewed the oscillations and concluded that the effects on components and controls are a nuisance only and do not reduce the safety of those systems.
The licensee's Station Review Board in Meeting 682 on September 11, 1979, and Meeting 683 on September 14, 1979, reviewed the item. The Company Nuclear Review Board reviewed the item in Meeting 45 on September 12, 1979. The licensee concluded that the facility could operate safely with the oscillation. The licensee committed to follow NSSS vendors recommen-dation to adjust the orifice during the first refueling to dampen the oscillations.
The inspectors observed operation of the plant at a power range 40 to 70% and at full power to determine the effects on plant control.
The following documents relating to the oscillation were reviewed:
B&W letter dated September 14, 1979 (DB-79-164), B&W 1etter dated September 5,1979 (DB-79-156), B&W memo dated October 2,1978 (NSS-14/T3.4), SRB minutes for meetings 682 and 683, and CNRB minutes for Meeting 45.
The inspectors identified no concerns which had not been addressed and resolved.
Based upon information available and observations it appears the oscillatory problem does not represent a safety concern.
(Closed) Unresolved Item (Reference IE Inspection Reports No. 50-346/
78-13, Paragraph 12, and No. 50-346/77-06, and licensee's responses dated May 17, 1977 and February 8, 1978): Preventative maintenance program for instrumentation and control system calibration. The licensee has established administrative controls in Administrative Procedure AD1844.01, Rev. 2, Preventative Maintenance to assure periodic recalibration of instruments and controls which are not controlled by the Technical Specifications but are safety-related including those covered by Section XI of the ASME Code. The 1522 i91-5_
inspectors review of the above shows the licensee has completed the items contained in his responses and resolved the concerns expressed in IE Inspection Report No. 50-346/78-13.
(Closed) Unresolved Item (346/79-21-03). Auxiliary Shutdown Panel instrumentation and controls: The licensee has placed a supply of spare bulbs at the shutdown panel. The licensee reviewed all non-operable switches on the auxiliary shutdown panel and control room SFAS panel.
If they were not labeled " spare," information tags were
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provided.
Two facility change requests (FCR 79-307 and 341) were initiated to remove the spare switches.
(Closed) Earesolved Item (346/79-21-01).
Operating logs:
On September 8,1979, a memorandum was issued to all licensed operators from the Operations Engineer in an attempt to improve the quality of Operating Logs. The memorandum expands upon the requirements and provides examples of the log entries required by AD1839.00. Adherence tothisguidanceandcontinuedemphasgsbymanagementshouldresult in an improvement in the logs.
(Closed) Unresolved Item (346/78-30-02) Reference IE Inspection Reports No. 50-346/78-30 and No. 50-346/79-04: Natural Circulation Test. The inspector reviewed the completed test package and per-formed independent calculat. ions using data from the tests. These calculations confirmed the licensee's results that at a core power level of 3.8 to 3.9% natural circulation flow was about 4.6 to 5.05%.
Acceptance criteria for the test was 1.55 to 1.60% minimum flow at this power range.
The licensee performed the test using various levels in the steam generator. The tests performed at the 35 inch level was to be used to support manual control of steam generator level using auxiliary feedwater (with no small break).
The licensee desired to control at this lower value to prevent unneeded reduction in pressurizer level caused by the rapid automatic filling of the steam generator to about 90 inches. Due to operating constraints (steam generator level)
and level swings using manual control the average level during the test was 38.9 inches; minimum level reached was 35.2 inches. Data from the test suggests that ability to achieve and maintain adequate natural circulation is not influenced by steam generator level, but is dependent upon auxiliary feed water flow as long as some level is maintained in the generator. The licensee's emergency procedures require the level be controlled at 35 inches. This was discussed with the licensee and the licensee modified the following procedures to require maintaining steam f,enerator level above 35 inches with a 35 inch minimum; EP1202.16,
" Loss of Steam Generator Feed," Rev. 10, EP1202.33, " Emergency Operations of NSS," Rev. 8, EP1202.24, " Loss of RC Flow," T-4096, EP1202.24, " Steam Supply System Rupture," T-4097, and SP1106.06, " Auxiliary Feedwater System,"
Rev. 9.
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The licensee has initiated a facility change request to install an
?.utomatic dual level setpoint control system for steam generator level control on auxiliary feedwater.
It will automatically control at a level of about 40 to 42 inches (with appropriate allowances) if there is no small break and at the auto essential setpoint if '.here is a small break.
3.
Inspection of Test Results
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a.
Test Results Reviewed TP 800.05, " Reactivity Coefficients at Power" (testing at 40%
and 75%)
TP 800.04, " Natural Circulation Test" (testing at 4%)
TP 800.08, "ICS Tuning at Power" (testing at several power levels)
TP 800.11, " Core Power Distribution" (testing at 15%, 40%, 75%
and 100%)
TP 800.23, " Unit Load Transient Test" (testing at several power levels)
TP 800.31, " Vibration and Loose Parts Monitoring (testing at several power levels)
TP 500.03, " Initial Radiochemistry Test" (testing at several power levels)
b.
Purpose of Review (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 signature:
or initials were contained in the "as-run" copy of the p rocedure.
(5) Verify documentation existed for QA involvement in testing.
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(6) Verify test results had been approved.
(7) Verify documentation existed for reviews by the SRB and ChPB and appropriate corrective action was taken on any comments.
c.
Findings The inspector did not identify any items of noncompliance or
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deviations and judged the test results to be acceptable.
Listed below are some specific inspector findings.
(1) TP 800.31, " Vibration and Loose Parts Monitoring" Test data (magnetic tapes and discs) are not stored in the facility's record storage room.
The licensee stated proper storage for this data would be found.
Data regarding some plant parameters appears to be incomplete or incorrect. The missing data does not affect the test results. The licensee acknowledged the inspectors comments.
A few copies of reduced data contained in the completed test package were impossible to read. The licensee replaced these with readable copies.
(2) TP 500.03, " Initial Radiochemistry Test" The test established base line chemistry data, monitored activity buildup.
Data indicates good fuel clad integrity and no leakage from the reactor coolant system to the secondary side of the steam generator or component cooling system.
Cs-137 was not detected in the radiochemistry tests. A graph (Attachment 16 to TP 500.3, " Expected Concentrations ofCs-137FromSurfaceContaminationofFuelEgements"}
indicates that the levels should have been 10 to 10 uCi/cc.
This would be in the detectable range.
No inves-tigation was made as to why none was detected. The licensee has since detected Cs-137 in his coolant and the hPC's Confirmatory Measurement Program has verified his ability to correctly quantify the levels of Cs-137 (IE Inspection Report No. 50-346/77-05). The licensee stated he would review the matter to determine if the graph was correct or there was a problem in the analysis at the time.
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4.
Status of Startup Test Program The licensee has completed the major portion of the test program.
Portions of some tests remain open to clear various deficiencies which have been evaluated by the licensee as not affecting safe opersiion of the facility.
The following tests arc under review by Power Engineering:
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Nuclear Safety Related 170.05, " Control Room Heating, Ventilation and Air Conditioni g"
- Resolving deficiency regarding fan balancing.
401.11, " Instrument System Ground Grid Pre-Op test" - Awaiting facility change to correct deficiency.
800.00, " Power Escalation Sequence Controlling Procedure - Resolve deficiencies in other related tests.
Non Safety Related 210.07, " Sampling System Non-Nuclear Areas" 280.02, " Turbine-Generator Acceptance Test" 281.01, " Turbine-Generator Lube Oil System" 291.01, " Excitation System Acceptance Test" 276.01, " Condensate System Acceptance Test" The following tests have not been completed:
Nuclear Safety Ra ated 267.01, " Station Drainage" - Safety related areas completed, testing remains in non safety related areas.
No, Safety Related 130.02, " Pump House Crane" 130.05, " Turbine Room Cranes" 130.07, " Intake Str acture Gantry Crane" 261.01, " Cathodic Protection System" 265.01, " Circulating Water System" 274.01, "MSR Acceptance Test" The following tests have been completed and are in review:
Nuclear Safety Related 261.02, " Freeze Protection and Electric Heat Trace" - Heat tracing on nitrogen needs additional changes to resolve deficiency 1522 195-9-
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600.35, " Piping Dynamic tests" - Awaiting formal closecut of deficiency from Bechtel Corp.
800.01, " Shield Survey - tvaluating need for installation of additional shields or changes to FSAR.
800.03, " Site and Station Radiation Survey - Same as 800.01 s
Non Safety Related 266.03, " Chlorination System Acceptance Test" 270.01, " Auxiliary Boiler and Auxiliary Steam" 270.02, " Auxiliary Boiler Oil System" 277.01, " Condensate Polishing Demineralizer" 290.03, " Station Cooling Water System" The licensee and inspectors did not view the incomplete porticas of the test program described above as resulting in unsafe operation of the facility.
5.
Licensee Review and Resolution of Identified Concerns The inspector met with the Vice President, Energy Supply, on August 14, 1979, to discuss action the licensee was taking to improve the timeliness and quality of the licensee's review of identified problem
.ind potential problem areas.
The inspector stated it was his under-standing, based upon his attendance at the May 31, 1979, management meeting documented in IE Inspection Report No. 50-346/79-12, than an important function of the new Nuclear Services Departmant will be to assure that identified problem areas are promptly resolved in a quality fashion. The Vice President, Energy Supply, confirmed the inspector'
2nderstanding. The licensee's actions in this area are partially in response to a RIII concern expressed in Paragraph 2.d of IE Inspeccion Report No. 50-346/79-03 relating to management problems in grasping and addressing technical problems in a timely fashion.
6.
Onsite Review Committee The charter for the Davis-Besse Station Review Board (Rev. 2 dated October 3, 1978) and minutes for meetings 569 (January 9, 1979)
through 683 (September 14, 1979) were reviewed to determine if the onsite review functions were being conducted in accordance with Technical Specifications. The following items were noted:
The charter is e brief compilation of Technical Specification a.
requirements.
It does not address the methods by s'lich the committee will fulfill its responsibilities, handle adminis-trative items, use of secretary, time frame for minutes distri-bution, minute approval method, qualification requirements for 1522 196
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alternates, prohibitions against reviewing and voting on items if they were generated by that member, etc.
This was discussed with the licensee. The licensee agreed to revise the charter.
b.
With the exception of item c below a review of the minutes and discussions with members reveal that the committee was fulfilling the requirements of Technical Specification 6.5.1.
Technical Specification 6.5.1.6.c requires the SRB to investi-
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c.
gate all violations of the Technical Spacifications including preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Vice President -
Energy Supply and to the Chairman of the Company Nuclear Review Board. There was no mention in the minutes of fulfilling this responsibility. There was no other system the SRB had for fulfilling this responsibility.
Specifically, the SRB did not do the above for the violations contained in IE Inspection Report No. 50-346/79-04. This appears to be an item of noncom-pliance (346/79-28-01) with Technical Specification 6.5.1.6.c and is considered to be an Infraction.
7.
Organization and Administration The onsite organizational structure was reviewed and found to be as described in Figure 6.2-2 of the Technical Specifications.
The qualifications of Operations Engineer, Operating Supervisor, two Shift Foremen, and Technical Engineer were reviewed and found to conform to the Technical Specifications and ANSI N18.1-1971.
The position of Lead I&C Engineer is presently vacant.
The Maintenance Engineer is serving as the Lead I&C Engineer until the position is filled.
No items of noncompliance or deviations were identified.
8.
Records and Document Control A preliminary review of auministrative instructions, records, record storage, and record storage facilities was conducted to obtain information for use in future inspection activities.
9.
Steam Generator Tube Leak Procedure The procedure EP 1202.57, " Steam Generator Tube Leak," Rev. 2 was reviewed.
The procedure does not specify under what conditions the facilities Emergency Plan will be activated, specific procedures to minimize releases during cooldown, radiation monitoring require-ments, how to quantify release rates from various turbine building paths or safety valves, grab samples to be taken, and steam generator isolation requirements for minor leak case. The problem areas in 1522 197
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the procedure were discussed with the licensee. The licensee stated
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the procedure would be revised.
This item is considered to be an Unresolved Item (346/79-28-02) pending the licensee's revision of the procedure and review by RIII.
10.
Unresolved Items s
Unresolved Items are matters about which more information is required to ascertain whether they are acceptable items, Items of Noncompliance, or Deviations. An Unresolved Item disclosed during the inspection is discussed in Paragraph 9.
11.
Exit Interview The inspectors met with licensee representatives (denoted in Para-graph 1) at the conclusion of the inspection on September 28, 1979.
The inspectors summarized the purpose and the scope of the inspection and the findings.
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