IR 05000259/1973013
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UNITED STATES I
- g ATOMIC ENERGY COMMISSION o
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I! Qh;I DIRECTORATE OF REGULATORY OPERATICNS c,+,[fg/ yy acciou n - suit e eis zao re acwr nec sr ne er, soar w.cs,
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Ar ta ur a, cconcia acaos RO Inspection Report No. 50-259/73-13 Licensee: Tennessee Valley Authority 818 Power Building Chattanooga, Tennessee 37301 Facility Name: Browns Ferry 1 Docket No. :
50-259 License No.:
DPR-33 Category:
B2 Location: Decatur, Alabama Type of License: 1098 Mwe, BWR (G-E)
Type of Inspection: Routine, Unannounced Dates of Inspection: September 11-14, 1973 Dates of Previous Inspection: August 16-17, 1973 Principal Inspector:
W. S. Little, Reactor Inspector Facilities Test and Startup Branch Accompanying Inspector:
E. S. Cantrell, Reactor Inspector Pacilities Test and Startup Branch
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Other Accompanying Personnel: None Principal Inspector:
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W. S.'Liffle, Reactor Inspector
'Dat6 Faci Test and Startup Branch g li/
[ _7 Reviewed By:
C. E. ffu'r$hy, Chi /f 4) ate'
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Facilities Test and Startup Branch 8604300309 860317 PDR FOIA MORROW 85-782 PDR L
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RO Rpt. No. 50-259/73-13-2-SUMMARY OF FINDINGS I.
Enforcement Action A.
Violations 1.
Certain activities under your license appear to be in violation of AEC requirements. These apparent violations are considered to be of Category II severity.
a.
Contrary to paragraph 2.2.B. of the Technical Speci-fications, TVA had not verified the correctness of
the relief valve limiting safety system setpoints before the reactor was started up.
(Details I, paragraph 7)
b.
Contrary to Criterion XV of Appendix B to 10 CFR 50, a nonconforming head 8asket was installed on the drywell by TVA.
(Details I, paragraph 5)
c.
Contrary to paragraph 6.3.A. of the Technical Specifi-cations and TVA Standard Practice BFA-25, the procedure for the operating hydrotest (BFGOI 100-7) did not require verification of the removal of jumpers used in the test.
(Details II, paragraph 6)
d.
Contrary to paragraphs 3.5.F. and 4.5.F. of the Technical Specifications, the reactor pressure was not immediately reduced to less than 122 psig or was the HPCIS immediately demonstrated to be operable upon failure of the reactor'
core isolation cooling system (RCICS).
(Details II, paragraph 5)
2.
Certain activities under your license appear to be in
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t violation of AEC requirements. These apparent violations are considered to be of Category III severity.
a.
Contrary to Criterion VI of Appendix B to 10 CFR 50,
- ne control room copy of the Technical Specifications
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was not up to date.
(Details I, paragraph 6)
b.
Contrary to paragraph 6.2. A.7. of the Technical Specifications, the Safety Review Board (SRB) minutes did not indicate that the startup program had been adequately reviewed, as required by the SRB Charter.
(Details I, paragraph 3)
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RO Kpt. No. 50-259/73-13-3-
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Contrary to paragraph 6.3. A. of the Technical Specifications, the temperature of the reactor vessel was not monitored as required by the procedure (BFGOI 100-7) during the operating hydro test.
(Details II, paragraph 6)
d.
Contrary to paragraph 6.3. A. of the Technical Specifications, the Master Hot Functional Test Instruction (11RFTI) did not have the proper approvals to perform the test prior to beginning.
(Details I, paragraph 2)
Contrary to paragraph 6.6.A. of the Technical e.
Specifications, several items subsequently reported as abnormal occuerences were not entered in the shift engineer's journal.
(Details II, paragraphs 2.a.,
2.b., 2.e. and 4)
f.
Contrary to paragraph 6.2.B.4
' *he Technical Specifications, the Plant Op s Revie?r Committee (PORC) did not investigate at cted 3* % it. crease in the "A" recirculation pump-spe<
(Details II,
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paragraph 3)
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Contrary to paragraph 6.7.2. A. of the Technical Specifications, the Director of Regulatory Operations was not notified that the required flow was not met in tests of the RCICS and that reactor pressure was not immediately reduced below 122 psig and that the HPCIS was not immediately demonstrated to be operable. '
(Details II, paragraph 5)
II.
Licensee Action on Previously Identified Enforcement Matters
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A.
Violations Prompt Reporting of Design and Construction Deficiencies TVA's reply to the violations was received on September 19, 1973.
Their corrective action will be followed up during the next inspection.
B.
Safety Items None l
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R0 Rpt. No. 50-259/73-13-4-
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III. New Unresolved Items 73-13/1 Review of Critical Instrumentation Valve Lineup Compared With As-Built Conditions Details I, paragraph 4.
73-13/2 Vacuum Breaker Failure Vacuum breaker between the reactor building and the suppression chamber did not function. TVA will submit an abnormal occurrence report.
(Details I, paragraph 4)
73-13/3 RCICS High Steam Flow Pressure Switch Setpoint Improper setpoint on instrument that senses high steam flow in the RCIC turbine steam line. TVA will submit an abnormal occurrence report and RO:II will follow up during the next inspection.
73-13/4 Failure of HPCI to Reach Rated Speed
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TVA abnormal occurrence report BFAO-739W dated September 18, 1973, describes failure of a fuse in an inverter power supply.
This appears to be a generic problem. Corrective action is incomplete and followup will be continued during the next inspection.
IV.
Status of Previously Reported Unresolved Items 73-12/1 Relief Valves Awaiting arrival of documentation at the site.
(Details I, i
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paragraph 7)
73-12/2 Failure To Limit Refueling Floor Vacuum TVA abnormal ~ occurrence report BFAO-734W dated August 17, 1973, describing the occurrence and cor7ective action was
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reviewed and the inspector has no further questions.
73-12/3 Main Steamline High Flow Trip Setpoint Error TVA abnormal occurrence report BFAO-735W dated August 23, f
1973, was reviewed and the inspector has no further questions.
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R0 Rpt. No. 50-259/73-13-5-
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I 73-12/4 HPCI Pressure Switch Malfunction TVA abnormal occurrence report BFAO-736W dated August 24,
1973, was reviewed and the inspector has no further
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l-questions.
(Details II, paragraph 2.a.)
73-11/1 Liquid Radwaste Automatic Shutoff
Resolved.
(Details I, paragraph 11)
l 73-6/1 HPCI and RCIC Systems Water Hammer Initial tests indicate that the problem is corrected.
Final resolution swaits completion of startup testing.
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(Details I, paragraph 12)
I 73-6/8 Preventive Maintenance Schedule Resolved.
(Details I, paragraph 13)
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73-5/2 Cask Decontamination Tank j
No change.
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Valve Wall Thickness (AEC Letter to TVA dated June 30, 1972)
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i Awaiting TVA report.
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V.
Design Changes
None VI.
Unusual Occurrences
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A.
Improper Setting tow Pressure Coolant Injection System (LPCIS)
Break Detecticn Logic Corrective action described in TVA letter dated September 18, 1973, i-was verified during this inspection and the inspector has no further j
questions.
(Details II, paragraph 2.e.)
B.
HPCIS Steam Supply Valve Failure TVA abnormal occurrence report BFAO-738W dated September 12, 1973,
reported the failure of valve FCV-73-16 to operate.
Investigation i
showed that the cause of failure was a gear installed backwards in the valve operator gear box. Followup by the inspector at the
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R0 Rpt. No. 50-259/73-13-6-site indicated that the gear had apparently been installed backwards prior to Unit 1 licensing. The inspector emphasized that the use of detailed written maintenance procedures could help prevent such recurrences. The inspector had no further questions on this item.
C.
Core Spray System Low Flow TVA abnormal occurrence report BFAO-737W dated September 12, 1973, reported the failure of the core spray system Icop 1 to deliver required flow. The inspector followed up on this occurrence during this inspection and has no further questions.
(Details II, paragraph 2.b.)
D.
Valve Operator Mounting Bolt Failure
The TVA report dated July 30, 1973, concerning the failure of mounting bolts for valve operators on suppression pool spray valves FCV-1-74-72 and -58 has been reviewed and the inspector has no further questions.
VII. Management Interview On September 14, 1973, the inspection results were discussed wich the following:
Tennessee Valley Authority (TVA)
Division of Power Production (DPP)
H. J. Green - Plant Superintendent J. B. Studdard - Operations Supervisor R. G. Metke - Plant Results Supervisor W. A. Roberts - Maintenance Supervisor i
J. C. Hodgins - Engineering Aide
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Division of Construction (DEC)
E. Hilgeman - Construction Administrator M. M. Price - Assistant Plant Manager Division of Design (DED)
L. D. Weber - QA Coordinator Of fice of Engineering Design and Construction (OEDC)
G. M. Tolson - Supervisor, Quality Standards
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RO Rpt. No. 50-259/73-13-7-Division of Power Resource Planning (DPRP)
P. J. Namanns - Nuclear Engineer General Electric Company (G-E)
J. E. Stice - Site Manager R. E. Spencer - Operations Manager The inspector discussed the potential violations, unresolved items and corrective actions described in the previous sections of this report with those listed above.
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RO Report No. 50-259/73-13 I-l DETAILS I Prepared by:
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W. S. Liter'e
'Da t'e Reactor Inspector Facilities Test and Startup Branch Dates of Inspection: September 11-14, 1973 Reviewed by:
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C. E. Wurpfiy, Chieff 6ataf ' i Facilities Test and Startup Branch 1.
Persons Contacted i
Tennessee Valley Authority (TVA)
Division of Construction (DEC)
R. T. Hathcote - Project Manager J. T. Walker - Mechanical Engineering Section Supervisor M. M. Price - Construction Engineer J. Roberson - Mechanical Engineer Division of Power Production (DPP)
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H. J. Green - Plant Superintendent f
T. Bragg - Nuclear Engineer i
R. Metke - Plant Results Section Supervisor
W. A. Roberts - Maintenance Supervisor General Electric Company (G-E)
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R. E. Spencer - Operations Manager 2.
Startup Test Program The inspector reviewed the startup tests conducted prior to the inspection.
The official records of the following were examined:
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RO Report No. 50-259/73-13 I-2-l f
" Master Hot Functional Test Instruction" (MHFTI)
STI-4, " Full Core Shutdown Margin Test"
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STI-5, " Control Rod Drive System"
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STI-6, " Source Range Monitor Performance and Control Rod Sequence" l
STI-35, " Recirculation and Jet Pump System Calibration"
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STI-90, " Vibration Measurements"
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The official test records were examined relative to the requirements of the FSAR, TVA administrative procedures, Technical Specifications j
and AEC regulations.
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f All tests to be conducted at 300 F (N50 psig) were completed. Tests
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I of the high pressure coolant injection (HPCI) and the reactor core
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isolation cooling (RCIC) systems had been conducted at 150 psig (%338 F).
It was found that the site authorization to perform the MHFTI had
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not been signed prior to initation of the tests contrary to TVA
administrative procedure BFA21, " Initial Startup Program," and
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paragraph 6.3.A. of the Technical Specifications.
3.
Safety Review Board (SRB)
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The minutes of the SRB were reviewed to evaluate whether it is
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operating in accordance with the Technical Specifications and the
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l TVA Operational Quality Assurance Manual (0QAM). Technical i
Specification 6.2.A.7 states that the charter shall identify i
the responsibility of the SRB in conducting reviews. The
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i charter states in OQAM,.Part I, Section 6.1.III.B.2. that the SRB j
l will assure adequacy of startup program, plans, and procedures
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requirements, the SRB minutes do not indicate that the startup program l
was reviewed during the past two years, and the only time it was i
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reviewed was in December 1970.
It was also noted that the SRB i
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i had not reviewed the preoperational test program since December 1970.
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4.
Vacuum Breakers Failure A DPP representative reported to the inspector on September 12, 1973,
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that the vacuum breakers between the suppression chamber and the reactor building had failed to function properly. The operator l
j manually opened the vacuum breakers at a AP f -0.6 inch of water.
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The reason for the failure was that the differential pressure switches
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R0 Report No. 50-259/73-13 I-3-PDIS-64-20 and -21, which actuate the vacuum breakers, failed to operate. TVA stated that their preliminary investigation indicated that root valves in the sensing lines from the suppression chamber to the AP switches were closed. They have found that these valves are not listed in the valve lineup, apparently because they were omitted from the drawings. Following the above occurrence, TVA reconfirmed the calibration and correct operation of the pressure switches.
Because of this problem, TVA said that before they restart the reactor, they will check all critical system instrumentation to make sure that valve lineups are consistent with the as-built conditions. The inspector told DPP representatives that this would be an unresolved item until the check is complete and any required corrective action taken.
5.
Drywell Head Gasket In reviewing the GRB minutes, the inspector found that a special meeting had been called because Browns Ferry had installed the wrong gasket in the drywell head. The minutes stated that TVA had previously calculated and reported verbally to the Directorate of Licensing that this original G-E supplied gasket would begin to deteriorate at a dose of 5 x 106 rads. Because of this, TVA had verbally committed to L that they would use a "Nordell" gasket 8 rad. The SRB which was calculated to withstand doses up to 10 determined that it was safe to operate the plant as long as the power versus time relationship was less that 10% power for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, after which time this gasket must be replaced. This determination was based on calculation for a loss-of-coolant accident assuming a TID-14844 fission product release. The SRB minutes stated that it was TVA's understanding that all BWR's were using the
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orginial G-E supplied gasket which TVA considers to be inferior to the "Nordell" gasket. TVA shut down the reactor on September 12, 1973, and committed to the installation of the "Nordell" gasket during
the shutdown. The inspector told the DPP representatives that this occurrence was considered to be indicative of a breakdown in TVA's quality control program contrary to the requirements of the TVA standard practice BFA6, " Quality Control of Materials and Parts," and of 10 CFR 50, Appendix B.
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R0 Report No. 50-259/73-13 I-4-l 6.
Document Control The inspector looked at the official copy of the control room l
Technical Specifications and found that the L approved change l
in the relief valve limiting safety system settings had not been entered as a Technical Specification revision approximately
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1-1/2 weeks after the change was approved by L.
The inspector l
told DPP representatives that this was indicative of a failure
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of TVA's document control program in violation of Criterion VI l
7.
Certification of Limiting Safety System Settings The Target Rock relief valves had recently been modified and their setpoints changed as described in TVA's report to the Directorate of Regulatory Operations dated August 29, 1973. The inspector asked if TVA had onsite documentation or vendor certification that the relief valve setpoints were properly set and tested. DPP and G-E representatives replied that documentation and/or certification i
was not onsite, but was supposedly on the way to the site. The
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inspector told DPP representatives that they violated the Technical Specifications when they operated the reactor without proof that the relief valve setpoints were correct. Subsequent to the inspection, a DPP representative telephoned the inspector on September 19, 1973, that the setpoint certification had been received from G-E and that some of the setpoints exceeded the
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Technical Specification requirement. This will be reported as an abnormal occurence.
8.
Operational Quality Assurance The inspector discussed the 0QAM and implementing procedures with a DPP representative with the objective of understanding how the line organization supervisor's quality assurance responsibilities are
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defined.
It was revealed that the supervisor's responsibilities
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relative to the audit of quality related functions in his area of responsiblity is not addressed in existing procedures. A DPP representative told the inspector that TVA will give considerstion to explicitly defining the supervisor's QA responsibilities. The inspector stated that this will be considered an unresolved item to be resolved during the next inspection.
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Drywell Ventilation and Air Conditioning DPP and DEC representatives described to the inspector the problem
of maintaining the drywell atmosphere temperature at 135-150 F.
Atmosphere temperatures as high as 160 F had been experienced at
primary system temperatures considerably less than the design valves.
TVA's investigation indicated that the problem was caused by inadequate distribution of cool air supply resulting in a stagnation problem in the upper part of the drywell. TVA decided to shut the reactor down and modify the air supply ducts in the drywell.
They began shutting down on September 12, 1973, and expected to have the modifications complete within less than one week.
10.
Shock Suppressors DPP and DEC representatives explained to the inspector that during the unplanned shutdown to correct the excessive drywell air temperatures, TVA would inspect all shock suppressors within the drywell and replace all soft ware on the Bergen-Patterson shock suppressors. The replacement parts will be the latest recommended by the vendor. The DEC representatives said that Bergen-Patterson has said that it may be a year before they can make a final recommendation concerning the seal material. TVA stated that inspection results will be reported to the AEC as requested by R0B No. 73-4 dated August 17, 1973.
11.
Liquid Radwaste Automatic Shutoff The inspector confirmed with DPP representatives that the liquid radwaste system has been modified to automatically stop radwaste discharge on receiving a high radioactivity signal. This item is considered to be resolved.
12.
HPCI and RCIC Systems Water Hammer
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The inspector observed one of the KPCI tests at a reactor pressure of 150 psig, looked at the RCIC and HPCI test data and talked with DPP test personnel. Test results indicate that the modifications to the turbine steam exhaust lines in the torus have resolved the water hammer problems experienced during the preoperational tests. Additional tests are scheduled on these systems and the inspector will continue to follow up on this proble s-
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R0 Report No. 50-259/73-13 I-6-13.
Preventive Maintenance Schedule The inspector reviewed the master preventive maintenance schedule with DPP representatives. TVA has made up this schedule based upon FSAR commitments, vendor's recommendations and internal TVA recommendations based upon their experience and the experience of others. The inspector confirmed that the schedule is in use and stated that he had no further questions at this time.
14.
Failure of Limitorque Operator Mounting Bolts TVA had reported this item as a design deficiency on June 4, 1973.
TVA submitted a final report dated July 30, 1973, on the occurrence.
The inspector confirmed during the inspection of June 26-29, 1973, that DEC had installed larger bolts as described in the report.
The inspector has reviewed the report and has no further questions.
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s-s RO Rpt. No. 50-259/73-13 II-1
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DETAILS II Prepared by::
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F. S. Cantrell, Readt4( Inspector
~ Date Facilities Test and Startup Branch Dates of Inspection:
t mber 11-14, 1973 Reviewed by:
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C. E. T'urphy, nief, Facilities
'Dat6 Test and Startup branch 1.
Persons Contacted Tennessee Valley Authority (TVA)
Division of P'ower Production (DPP)
H. J. Green - Plant Superintendent J. B. Studdard - Operations Supervisor A. Qualls - Assistant Operating Supervisor R. G. Metke - Plant Results Supervisor 2.
Abnormal Occurrences The details of the following reported abnormal occurrences were reviewed
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with respect to documentation, investigations, and reports required by the Technical Specifications:
a.
HPCI Pressure Switch Malfunction
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BFA0-736W reported to the Directorate of Licensing in a letter dated Augus t 24, 1973, - BFAO-736W was not documented in ths Shift Engineer's Journal; however, the event was investigated and reported to the Commission as required by the Technical Specifications.
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b.
Core Spray System Low Flow BFA0-737W reported to the Directorate of Licensing in a letter dated September 12, 1973, - BFAO-737W was not documented in the Shift i
Engineer's Journal; however, the event was investigated and reported to the Consnission as required by the Technical Specification (The licensee's investigation showed that the indicated low flow in the
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core spray system was only in the test loop, and that the core spray
system was always operable).
c.
HPCIS Steam Supply Valve Failure BFA0-738W repo::ced to the Directorate of Licensing in a letter dated j
September 12, 1973, - no deficiencies were noted.
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R0 Rpt. No. 50-259/73-13 II-2 d.
KPCIS Power Supply Failure
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BFAO-739 reported to Region II, Directorate of Regulatory Operaticas by telegram dated September 9,1973, - no deficiencies
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were noted, e.
Improper Setting LPCIS Break Detection Logic BFAO-7310 reported to Region II, Directorate of Regulatory
Operations by telegram dated September 7,1973, - the change in switch setting discussed in the telegram were not documented in
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the Shif t Engineer's Journal; however, the details were reviewed
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by the Plant Operations Review Committee (PORC), and an.'.nternal
abnormal occurrence report was prepared to supply the details for
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the licensee's 10 day written report to the Directorate r,f Licensing
due September 18, 1973.
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3.
Unplanned Increase Recirculation Pump Speed
The Shift Engineer's Journal on September 4,1973, noted that the speed of the A recirculation pump drif ted up 35%, and the pump was tripped at 7:45 am.
The reactor was shutdown at the time. A licensee's representa-tive stated that the pump speed controls were being adjusted during this period, and the drif t in speed was attributed to this work even though no work was in progress at the precise time. There have been no additional problems with speed control since the work was completed according to the licensee's representative. The Shif t Engineer's Journal did not show the resolution of this problem, and minutes of the PORC meetings did not show
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that the PORC reviewed this event.
4.
Log Book Review
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The Shif t Engineer's Journal (Logbook) was reviewed against the require-ments of the Technical Specification, and the Power Plant Operation
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Section Instruction for the periods August 16-17, 1973, and September 1-11,
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1973. The contents of, the Journal are specified in Letter No. 16 of the i
Power Plant Operation Section Instructions. The Shift Engineer's Journal
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settings determined by instrumentation personnel performing surveillance
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checks. Paragraph 6.6.A of the Technical Specifications specifies that
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" Records and/or logs shall be kept in a manner convenient for review as s
indicated below:
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3.
Abnormal Occurrences
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Checks, inspections, tests.
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R0 Rpt. No. 50-259/73-13 II-3 The Shif t Engineer's Journal did not document abnormal occurrences involving changes in switch settings reported in letters to the Directorate of Licensing dated August 24,1973, (BFAO-736W, Para-graph 2a), and September 12, 1973, (BFAO-737W, Paragraph 2b), and by telegram to Region II Regulatory Operation dated September 9,1973, (BFA0-7310, Paragraph 2e).
5.
Reactor Core Injec sn Cooling System (RCIC)
The Shif t Engineer's Journal shows that tests of the RCIC using the test loop (with reactor pressure at 150 PSIG, and temperature at 3500F)
on September 8,1973, at 10:30 pm, and on September 9,1973, at 3:45 am,
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indicated a maximum flow of 560 GPM. Paragraph 4.5.F.1. of the Technical Specifications requires a minimum flow of at least 600 gpm during each test. Paragraph 3.5.F. of the Technical Specifications requires the RCIC to be operable with irradiated fuel in the reactor when reactor pressure in greater than 122 psig, but permits continued operation for seven days if the high pressure coolant injection system (HPCI) is operable. Para-graph 4.5.F.2 of the Technical Specification requires the HPCI be demon-strated immediately when the RCIC is determined to be inoperable.
P.aactor pressure was not reduced below 122 PSIG until 12:35 pm, September 9,1973. During a subsequent test of the HPCI in the automatic mode at 2:30 am, September 10, 1973, flow through the HPCI failed to reach the 5000 gpm that is required for the system to be considered operable by paragraph 4.5.E.1 of the Technical Specifi-
cation (BFAO-7311).
Subsequently, the orifice in the RCIC test loop was removed, and the RCIC was declared operable following a test at 6:45 pm, September 10, 1973.
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Plant records do not show that the apparant failure of the RCIC, and I
the delay in testing the HPCI and/or in reducing reactor pressure below 122 psi were investigated by the PORC. The failure of the RCIC to meet the required test flow on September 8 & 9, 1973 was not reported to
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Regulatory Operation within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as required for an abnormal occurrence.
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Operating Hydro Test
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The test procedure, BF GOI 100-7, dated July 16, 1973, and approved by the Plant Superintendent, was performed September 1,1973. The results
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were compared against the requirements of the procedure.
l Plant records indicated that only 2 of six points for monitoring tempera-ture of the reactor vessel were " called up" and recorded on the computer printout as specified in the procedure. A checkoff sheet provided with the procedure listed prerequisites for the test; however, no space was provided to show: a) the leak inspection was made, or where leaks were
found; b) approval or that the test was satisfactory, c) that jumpers installed on interlock were removed (Required by Standard Practice Proce-j dure BFA 25 Jumpers, Inhibits, and Wire Removal), d) that key valves were cycled to show operability following the test (as required by the procedure).
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, e s RO Rpt. No. 50-259/73-13 II-4 There was sufficient documentation to show that the test was performed and met the requirements of an operational hydro test, however, the available data indicates that the test was not
" adhered to" as required by Paragraph 6.3A of the Technical Specification.
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REGION II
M INSPECTION REPORT IDENTIFICATION. SHEET
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REPORT NO. JiC O
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To:
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Licensee:
4W Arc /
Inspection Dates [
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Other Inspectors M
Feeder Report Prepared By:
Date Inspector's Evaluation (continue on attached sheet, if necessary)
t'&pMraY y&am'rr* ve3 fg(2 $
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Signatiuff '
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The attached Feeder Inspection Documentation is transmitted in reviewed and approved. form.
I Comments:
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[FET R INSPECTIO:s REPORT IDENTIFIr ' ION SHEET
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REPORT 50.
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C. E. durphy, Chie'f, Facilities Test and Startup Branch
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Inspection Dates 9/11-14/73
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Licensee:
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Feeder Report Prepared By:
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.Date F. S. Cantrel'1, Reactor Inspector
Facilities T d Star Branch
. Inspector's Evaluation (continue on attacksk akeet, Np e s cecessary)
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I don't think the administration of BF-1 is as bad as the number of violations
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may indicate,%ecause we have compared what we' found against a very strict
. interpretation of the Technical Specification.
In the longrun, I think this
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.', will benefit both the licensee and Regulatory in that the licensee in his investigation of problems will be more likely to refer to the requirements
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of.the Techn, ital Sppcification as part of his investigation. ;; It also puts top management on notice that the Technical Specification must be considered in decision concerning operation of the plant.-
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F. S. Cantrewll, Reactor Inspector
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The attachr.d Teeder Inspection Documentation is thus'nitted in reviewed and
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