IR 05000321/1980006
| ML19312E801 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 03/28/1980 |
| From: | Kellogg P, Rogers R, Sauer R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19312E784 | List: |
| References | |
| 50-321-80-06, 50-321-80-6, 50-366-80-06, 50-366-80-6, NUDOCS 8006090547 | |
| Download: ML19312E801 (8) | |
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- p neuq'o, UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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l Report Nos. 50-321/80-06 and 50-366/80-06 I
Licensee: Georgia Power Company 270 Peachtree Street
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Atlanta, GA 30303
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Facility: Hatch Docket Nos. 50-321 and 50-366 License Nos. DPR-57 and NPF-5 Inspection at Hatch site near Baxley, Georgia f
Inspectors:_
d. Tm /v M S/Y//BO R. F. Roger i
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Date Signed W8 3l21l80 a e_ v
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R. C. Sauer Dat6 Sig' ned Approved by:
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P. J. Kellogg, Acting Sed. ion Chief, RONS Branch Da'e Signed t
SUMMARY Inspection on February 19-21, 1980
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Areas Inspected
This special, announced inspection involved 46 inspector-hours on site in the
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areas of small break loss of coolant accident procedures and training.
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Results
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' 'e Of the two areas inspected, no items of noncompliance or deviations were identi-
fied in one area..One deviation was found in one area (deviation - failure to e
ensure the stations' licensed operators have committed to memory the immediate
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actions of the facility's emergency procedures, paragraph 7).
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DETAILS
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1.
Persons Contacted Licensee Employees
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- M. Manry, Plant Manager
- T. V. Green, Assistant Plant Manager
- S. X. Baxley, Superintendent of Operations
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- H. W. Dyer, Operations Supervisor t
- R. T. Nix, Superintendent of Maintenance
- C. Coggins, Superintendent of Engineering Services
- D. McCusker, QC Supervisor
- C. R. Miles, Jr., QA Field Supervisor
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- C. E. Belflower, QA Site Supervisor
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- P. E. Fornel, Senior QA Field Representative
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- D. F. Moore, Supervisor of Nuclear Training
- D. A. Lee, Methods and Training Specialist
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Other licensee employees contacted included nuclear control center operators, I
nuclear shift engineers and nuclear plant supervisors.
NRC Resident Inspector
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- R. F. Rogers
- Attended exit interview 2.
Exit Interview The inspection scope and findings were summarized on February 21, 1980 with those persons indicated in Paragraph I above. The licensee was informed of the decision to issue the Notice of Deviation for failure to ensure the
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station's licensed operators have committed to memory the immediate actions of the facilities emergency procedures in a telephone conversation on March 7, 1980.
3.
Licensee Action on Previous Inspection Findings
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Not inspected. -
4.
Unresolved Items Unresolved items were not identified during this inspection.
5.
Small Break Loss of Coolant Procedure Review The inspectors compared the following licensees' small break loss of coolant accident (SBLOCA) emergency procedures to the operator guidelines developed
by the General Electric Operating Plant Owner's Group as approved by the
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NRC Bulletins and Orders Task Force in its letter to the owner's group dated October 26, 1979:
HNP-1902 Rev. 7 Pipe Break Inside Primary Containment HNP-1904 Rev. 5 Primary Coolant System Pipe Break Reactor Building.
In addition, the procedures were reviewed as to technical content, clarity in terms of individual actions and precautions, and procedural flow with respect to timely initiation of all operator actions.
Procedure HNP-1902 reflected the majority of the elements contained s.
within the approved guidelines, however, due to the placement of the emergency procedure within the confines of the normal operating proce-dure subdivision (1000 series) as permitted by HNP-ADM-9 section H.2 and the licensee's interpretation of the ANSI N18.7-1976 requirements
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for emergency procedure format, the contents of the guidelines (imme-l diate and subsequent actions) were combined and placed under the all-encompassing heading of operator actions. The text for operator actions therefore became one of comprehensiveness versus simplicity and presented a serious problem for the operators who must memorize these actions as detailed in section 7 of this report.
Based on procedure review and licensed operator interviews the inspec-tors identified the following procedural comments to HNP-1902 for licensee evaluation:
(1) The procedure should specify the actions the operator must take without normal off-site power being available in just a sentence or two instead of repeating the actions required if normal offsite power were available except this time, diesel generator start, bus tie in and load sequencing are interlaced within the text.
(2) Resolution is required as to what temperature the suppression pool water is to be maintained when RER is placed in the torus spray mode af ter water level in the reactor vessel is restored (see paragraph 6.(1)).
(3) The caution, warning reactor water level reading inaccuracies as
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a result of high drywell temperature affects, in section C.2.q i
should also be incorporated into section C.l.b and expanded to indicate the affects are slow since the thermal time constant of the yarway reference leg is approximately 20 to 30 minutes.
(4) Subsequent operator instructions section D.1, should not allow the operator the option to secure Emergency Core Cooling Systems (ECCS) without first obtaining the shift foreman's approval.
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b.
Procedure HNP-1904 similarly reflected the majority of the elements
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contained within the approved guidelines, however the following incon-
sistencies were found:
(1) The procedure contained no automatic actions.
(2) All guideline considerations, except for the control of vessel level with available high pressure injection systems, were con-
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tained within the procedures' operator actions section. The exception, which is an immediate operator action as defined by the guidelines, was placed in the subsequent operator actions
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section of the procedure.
The inspector identified the above inconsistencies to the licensee
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for resolution (321/366/80-06-01).
The inspectors had two additional comments applying to both procedures:
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(1) The placement of the SBLOCA procedures within the cenfines of the normal operating procedure subdivision (1000 series) in order to j
provide a means of manual operation of the Emergency Core Cooling and other automatic initiated systems appears to confuse the intent of an emergency procedure - that being readily accessible.
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The inspector's concern is that all emergency procedures should be assembled under the same heading or the Hatch designated 4000 series subdivision. The licensee is requested to investigate this issue (321/366/80-06-02).
t (2) Ensuring support systems to ECCS equipment are also operating when the various ECCS Components are placed into operation to
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provide core cooling. This practice would assure ECCS Component availability for continued long-term operation by prevention of component failures or component isolations as a result of radio-active discharges from the system. Examples of this item include:
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(a) Availability of the Standby Gas Treatment System and High i
Pressure Coolant Injection (HPCI) room coolers to support HPCI.
(b) LAvailability of Plant Service Water to cool the RHR (LPCI)
and core spray pump bearing and room coolers.
6.
Small Break Loss of Coolant and TMI Lessons Learned Training.
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The inspectors reviewed the training the licensed personnel received on the small break LOCA procedures required to be completed by December 31, 1979
as indicated on page 5 of Enclosure 6 to Darrell G. Eisenhut's letter to all operating nuclear power plants dated September 13, 1979. Inspection of the formal classroom training for shift and non-shift licensed operators
indicated the training was not complated by the January 1,1980 deadline i
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but was conducted during the period February 15 through February 19, 1980.
Though performance of the required training has been met the inspectors
expressed a concern to the licensee representative that the facility should
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improvise a method to ensure that all future NRC stated requirements are
completed in a more timely manner.
Inspector review of the February classroom training indicated the two-hour instruction covered the following topic areas:
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The Bulletin and Orders Task Force Audit of Small Break Loss-of-Coolant
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Accident Emergency Procedures and Operator Retraining Report dated January 8,1980.
b.
General Electric's Services Information Letter SIL No. 299 dated
t July 25, 1979, detailing high drywell temperature effects on reactor vessel water level instrumentation.
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NEDO-24708 Section 3.1.1.2, Operator Guidelines
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HNP-1902 Pipe Break Inside Primary Containment.
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HNP-1904 Primary Coolant System Pipe Break Reactor Building f.
HNP-1905 Pipe Break in Turbine Building
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HNP-1906 Pos( Accident Venting h.
HNP-2405 Drywell Floor Drains Sump Trouble At the conclusion of the lectures, a 45 minute examination was administered to the operators. The inspector reviewed 50% of the graded examinations.
Based on the quality of the responses made by the operators to the test
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questions the inspector had the following comments on the material presented:
(1) Resolution is required as to the maximum bulk temperature the suppression pool should be maintained to prevent excessive loads to the pool boundary and structures during safety / relief valve
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discharges. The value presented by lecture.is unlike the value i
specified in the HNP-1902 procedure or technical specification section 3/4.6.
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(2)
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The operators were instructed to utilize the cold reference leg I
type of level indicators (such as the normal operating range and fuel zone level indicators) should an unusual condition of very
high drywell temperature occur following an accident condition
within the drywell. This is only true on BWR 5's, BWR 6's and
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some late BWR 4's since their sensing lines are routed out of the drywell. Plant Hatch does not have this type of arrangement.
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Further, the level indicators covering the fuel zone region will i
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condition of which is present during calibration of the instrument.
The above comments were provided to the licensee for resolution
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(321/366/80-06-03).
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In addition to the classroom instruction, the licensee included a walk through of the procedures with a shift supervisor or training coordinator during the period February 11 through 13, 1980.
The inspector reviewed the training outlines, hand-outs, and materials associated with the above training areas.
In addition, training records were reviewed to insure that all licensed personnel had attended training sessions that had been completed prior to and during the inspection.
Based upon the review of the training program as defined above the inspector determined that the licensee's training program was adequate.
7.
Small Break Loss of Coolant Accident - Operator Interviews
'l The inspectors interviewed eight licensed operators, which included one
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a staff (off-shift) SRO, two shift supervisors. (SRO), one shift foreman, two SRO's on shift but not acting as shift supervisors, and two shift reactor operators.
The licensed operator interviews.*were performed to determine the adequacy of the appropriate procedures from a functional standpoint and the effec-tiveness of the training program. The following areas were covered:
Understanding what constitutes a small break LOCA.
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Differentiation between an LOCA and other depressurization events.
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Familiarity with the SBLOCA procedures.
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Operator knowledge of appropriate related procedures.
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Confirmation that the appropriate procedures immediate actions were e.
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memorized.
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Understanding the procedure's subsequent actions and precautions that ensure plant safety.
Recognition of the importance of the primary and backup heat sinks.
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Ability to determine break locations.
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Walk-thrcughs of the procedures including system-related aspects of the procedure to ensure that the licensed operator actions could be
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performed (see also paragraph 8).
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Knowledge of transient response characteristics necessary to guide the
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licensed operator to the correct procedure.
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The ability of the operator to recognize level variances and their
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meaning.
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Recognition of possible instrumentation abnormalities including those encountered during the TMI transient and environmental considerations.
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The understanding of how Emergency Core Cooling Systems (ECCS) initiate i
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and how they function to place the reactor in a safe shutdown condition.
Understanding the underlying causes of TMI and how these causes can be n.
related to a BWR.
Based on the operator interviews in the above areas the inspectors judged the SBLOCA training adequate. The inspectors did note however, that the operators were unable to recite all the immediate operator actions specified by the subject procedures. Contributing causes for this problem are:
The licensee's interpretation of the ANSI N18.7-1976 requirements for
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procedure adherence (section 5.2.2).
The standard states in part that
" procedural steps for which actions should be committed to memory include, for example, immediate actions in emergency procedures". The licenseg representative stated at the exit interview that the facility's licensei personnel were noc required to memorize the operator action steps detailed in all of the station's emergency operating procedures, only for those steps specifically defined to be memorized in the Annunciator Response procedure HNP-2001. Inspector review of the procedure identified that only three events require operator memoriza-tion of steps to be taken, they are:
reactor scram with MSIVs open; reactor scram with MSIVs closed and turbine trip. These events repre-
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sent only two of the twenty-six events listed in Regulatory Guide 1.33 which complements ANSI N18.7-1976 and details the emergency procedures the facility must have established and implemented as required by Section 6.8 of the individual unit's technical specifications.
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l The licensee's interpretation of the ANSI N18.7-1976 requirements for
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emergency procedure format and content (section 5.3.9.1).
The operator
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actions presented in the procedures identify all the operator actions required to place the plant and containment in a near-normal condition vice identifying the actions the operator must take as soon as possible to protect the core and to reduce the loss of primary coolant inventory.
Additional actions to bring the plant and containment back to a stable condition should be identified in the subsequent actions. The net result of the licensee's interpretation is the expansion of the immediate operstor actions to pages beyond the concept of simple and direct actions to prevent or mitigate the consequences of a serious condition.
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Placement of the SBLOCA procedures in the normal operating procedure
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subdivision (2000 series) tends to promote expansion of the procedure to be all-inclusive and general. For additional details see paragraph i
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The training and retraining program for licensed personnel as detailed
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in HNP-ADM-200 does not provide guidance that emergency procedure imunediate operator actions are to be memorized.
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Failure of the licensee to ensure that the station's licensed operators have committed to memory the immediate operator actions of its designated
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emergency procedures constitutes a deviation to NRC commitments
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(321/366/80-06-04).
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The inspectors further identified that the licensee's requalification program should be enhanced to cover the following additional areas:
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(1) Heat transfer and fluid flow fundamentals i
Saturated temperature conditions in the reactor vessel are
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obtained through conversion of reactor done pressure versus
reactor water cleanup system or recirculation system suction leg temperature readouts.
Significance of safety-relief valve thermocouple readout.
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Determination of superheated conditions.
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(2) Drywell high temperature affects on level instrumentatica - degree of variance and time frame associated to cause the affect.
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(3) Alternative methods of determining if the core is adequately r
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cooled should level instrumentation be lost.
(4) Verification that support systems are operable in order to support
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ECCS components.
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(S) InstructionastotheoperationoftheADSlogicafterreactor f
vessel blowdown is accomplished.
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(6) Instruction as to the meaning and reliability of the fuel zone level indicators during accident situations.
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8.
Small Break Loss of Coolant Accident - System Considerations
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The inspectors reviewed system-related aspects of procedures to ensure that
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operator actions subsequent to an SBLOCA could be performed. System con-
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siderations in the following areas were reviewed:
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Instrumentation to carry out operator actions in the SBLOCA procedure.
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Understanding of the power operated safety relief valve (SRV) position indication system (pressure switch on the individual SRV tail pipe)
including thermocouple monitoring.
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Equipment response to safety injection reset.
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Safety injection effects on containment isolation.
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Real time consideration of SBLOCA procedure actions, including adequate y
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time to remove one RHR (LPCI) division for containment spray / torus
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cooling.
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Instrumentation verified for environmental effects (for the conditions,,(
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prevailing at the time of the accident), power supply (with loss of
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offsite power and a single failure in the most limiting instrument
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bus), and redundancy (in sensor and readout device).
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No problems were identified with system considerations a through e.
Item f could not be asolved prior to the inspector's departure from the site.
This item is considered open pending further inspection by the Regional
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Office (321/366/80-06-05).
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