IR 05000324/1980004
| ML19323B655 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 03/20/1980 |
| From: | Kellogg P, Riley B, Sauer R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19323B653 | List: |
| References | |
| 50-324-80-04, 50-324-80-4, 50-325-80-04, 50-325-80-4, NUDOCS 8005140034 | |
| Download: ML19323B655 (7) | |
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UNITED STATES d00514o O~$8')
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NUCLEAR REGULATORY COMMISSION o,
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REGION II
101 MARlETTA ST., N.W SUITE 3100 g
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Report Nos. 50-324/80-04 and 50-325/80-04 Licensee:
Carolina Power and Light Company 411 Fayetteville Street Raleigh, NC 27602
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Docket Nos. 50-324 and 50-325 License Nos. DPR-71 and DPR-62 Inspection at Bruns ic if e r Southport, North Carolina Inspected by:
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11 ate Signed Approved by:
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Je 70 P. J.[/Kel/96g, A 1.g(g-Section Chief Dat Signed
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SUMMARY Inspection on February 5-8, 1980 Areas Inspected This special, announced inspection involved 58 inspector-hours on site in the areas of small break loss of coolant procedures and training.
Results o
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Within the areas inspected, no items of noncompliance or deviations were identified.
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o DETAILS i
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Persons Contacted
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Licensee Employees t
- J. M. Brown, Operations and Maintenance Superintendent D. C. Cooper, Shift Specialist
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J. A. Padgett, E&RC Supervisor E. H. Norwood, Senior Specialist-Training
- R. M. Poulk, NRC Coordinator
- M. A. Jones, Training Supervisor B. Parks, TMI Brunswick Site Manager Other licensee employees contacted included nuclear control center operators,
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nuclear watch engineers and nuclear plant supervisors.
i Other Organizations
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R. Noble, General Electric B. Griffith, United Engineers
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NRC Resident Inspector J. E. Ouzts
- M. J. Davis (In Training)
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- Attended exit interview 2.
The inspection scope and findings were summarized on February 8,1980 with those persons indicated in paragraph I above. The inspectors reviewed with the licensee the areas of the inspection and stated that the findings were clear in these areas, but did identify one item to be followed upon subse-quent inspections.
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3.
Licensee Action on Previous Inspection Findings Not inspected.
4.
Unresolved Items No unresolved items were identified during this inapection.
S.
Small Break Loss of Coolant Procedure Review
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The inspectors compared t.he following licensees' small break loss of coolant accident (SBLOCA) emergency instructions (EI) to the operator guidelines developed by the General Electric Operating Plant Owners' Group as approved by the NRC Bulletins and Orders Task Force in its letter to the owner's group dated October 26, 1979:
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EI-1.2 Rev 10 Rupture Inside Drywell EI-1.3 Rev 0 Small Break Outside Drywell
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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.
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Instruction EI-1.2 reflected the majority of the elements contained a.
within the approved guidelines, however the placement of certain
f immediate operator actions such as verification of the automatic
transfer of the High Pressure Coolant Injection (HPCI) system suction i
and manual transfer of the Reactor Core-Isolation Cooling (RCIC)
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suction from the condensate storage tank to the suppression chamber;
the manual control of injection flow water rates to the high pressure injection systems in order to stabilire level; and the depressurization of the vessel should vessel pressure increase above the shutoff head i
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of the low pressure systems being used to maintain level were considered
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subsequent actions and placed accordingly in the instructions. The licensee implemented the guidelines as guidelines as addressed in their THI-2 lessons
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learned short term requirements (NUREG-0578) response to the office of Nuclear Reactor Regulation dated December 31, 1979.
Based on procedure reviews and licensed operator interviews the inspectors recommended the following procedural changes to EI-1.2 for licensee evaluation:
(1) Develop a curve relating drywell temperature to an associated correction factor per level instrument reference column or identify
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that a fixed conservative value be subtracted from indicated
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1evel should drywell temperature increase greater than 135 F for
a specified time interval.
(2)
Indicate the alternate automatic initiation for ECCS equipment at reactor low level No. 3 indication under step 3.1.3 to be drywell pressure greater than 2 psig and less than or equal to 410 psig (versus greater than or equal to) so the operator can determine
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if ECCS equipment energized prematurely should pressure indicate
l e.g. 500 psig at the time of initiation.
f (3) Identify what portion (Immediate actions, etc.) of EI-31, Reactor Scram Instruction, is to be performed under step 3.2.2.
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(4) Add "take the reactor MODE switch out of RUN to prevent MSIV closure on low vessel pressure" to section 3.2.
i (5) The licensee should consider cross-referencing matertal when an action statement such as " initiate containment spray should con-tainment pressure exceed 20 psig" in step 4.0.4 is presented with
no additional supportive statements following.
Further, the licensee should investigate a method of improving procedure subsection identification indexing for added clarity and ease to the operator when the appropriate subsection of a multi-subsectioned
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procedure is sought. An example of this comment is the operation of the Containment Atmosphere Dilution (CAD) system addressed in step 4.0.13.
(6) Deletion of steps 4.0.8 and 4.0.9 as soon as modifications to the Containment Atmosphere Control (CAC) and CAD systems are completed.
(7) Procedural step 4.0.13 is currently inadequate irt that it does not provide a calculational sheet and method or identificatior of an individual or support group or provide reference to a procedt.re that the operator can use to determine the contairunent leak rate requested by this step.
b.
Instruction EI-1.3 is a new procedure plant Brunswici developed to specifically address small breaks outside the drywell since no procedure existed that could adequately cover this situation. Essentially the procedure mirrored the guideline procedure with additional amplification added as necessary. The inspectors recommended the following procedural changes to EI-1.3 for licensee evaluation:
(1) Symptoms 6 and 7 of section 2.0 should be expanded to allow the operator to relate to associations versus affects for this type of parameter change in a break outside the drywell. For example:
a decrease in generator output should be associated with no increase in steam flow or an increase in reactor power should be associated with no increase in generator output.
(2) A new step, step 3.2.11 should be added to the procedure to verify RHR (LPCI) and Core Spray (CS) are injecting into the reactor vessel and that water level responds accordingly on multiple indicators after vessel pressure is reduced as a result of the Automatic Depressurization System (ADS) blowdown.
The inspectors had a general comment on the two procedures, that of:
c.
ensuring support systems to ECCS equipment are also operating when the various ECCS components are placed into operation to 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 a radioactive discharge from the system. Examples of this item include:
Availability of the standby gas treatment system and HPCI room
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coolers to support HPCI Availability of RHR Service Water to cool the RHR (LPCI) and CS
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pump bearing and room coolers.
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6.
Small Break Loss of Coolant and TMI Lessons Learned Training The inspectors reviewed the training reactor operators and auxiliary operators received for the SBLOCA procedures required to be completed by December 31, 1979 as indicated on page 5 of Enclosure 6 to Darrell G. Eisenhaut's letter to
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-4-all nuclear power plants dated September 13, 1979. Formal classroom training for shift and non-shift licensed operators was given during the. pociod December 17, 1979 through December 19, 1979. The lesson outline covered the following topic areas:
a.
NED0-24708 Section 3.1.1.2 Operator Guidelines b.
EI-1.1 Leaks Inside Drywell
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EI-1.2 Rupture Inside Drywell d.
EI-1.3 Small Break outside Drywell e.
EI-8 Abnormal Reactor Water level f.
EI-15.1 Station Blackout Operation g.
EI-31 Reactor Scram h.
EI-40 SRV Fails Open i.
Level Instrumentation, especially effects of Small Breaks on level instrumentation.
The licensee evaluated the type of training needed and judged that the two hour instruction consist of the procedural changes made as a result of the small break LOCA analysis and the philosophy behind the changes. Discussions of the applicable TMI-2 lessons learned short-term requirements of NUREG-0578 and the guidelines of General Electric Company NEDO-24708 were also conducted.
Though no walk through of the procedures with the shift supervisor or training coordinator were conducted, the licensee does intend to schedule a training seminar on leaks and ruptures in March,1980. The training will include walk throughs and senarios on certain annunciation and what actions the operator should take.
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The inspectors 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 the inspection.
Based upon review of the training program as defined above ari with respect to the CP&L letter identified earlier to the Office of Nuclear Reactor Regu-lation dated December 31, 1979 detailing the scope of licensed operator training, the inspecto :s determined that the licensees training program was adequate.
7.
Small Break Loss of Coolant Accident-Operator Interviews The inspectors inter"lewed six licensed operators, which included one staff
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(off-shif t) SRO, two shif t supervisors (SRO), an SRO on shift but not a shift supervisor, and two shift reactor operators.
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-5-The licensed operator interviews were performed to determine the adequacy of the appropriate procedures from a functional standpoint and the effectiveness of the training program. The following areas were covered:
a.
Understanding what constitues a small break LOCA.
b.
Differentiation between a LOCA and other depressurization events.
Familiarity with the SBLOCA procedures.
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Operator knowledge of appropriate related procedures.
e.
Confirmation that the appropriate procedures immediate actions were memorized.
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Understanding the procedures subsequent actions and precautions that ensure plant safety.
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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-throughs of the procedures including system-related aspects of the procedure to ensure that the licensed operator actions could be performed (see also paragraph 8).
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Knowledge of transient response characteristics necessary to guide the 11 censed operator to the correct procedure.
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The ability of the operator t.
recognize level variances and their meaning.
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Recognition of possible instrumentation abnormalities including those encountered during the TMI transient and environmental considerations.
The understanding of how Emergency Core Cooling Systems (ECCS) initiate m.
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, however they recommended that the licensee's requalification program be enhanced to cover the following areas:
Heat transfer and fluid flow fundamentals a.
Saturated temperature conditions in the reactor vessel are obtained
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through conversion of reactor dome pressure versus Reactor Water Cleanup system or Recirculation system suction leg temperature read-out _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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Significance of safety-relief valve thermocouple readout.
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Determination of superheated conditions.
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Drywell high temperature affects on level instrumentation -degree of variance and time frame associated to cause the affect.
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Alternate methods of determining if the core is adequately cooled should level instrumentation be lost.
d.
Operator awareness of plant modifications i.e., the SRV position indi-cation modification. Operators are not aware of what the illuminated red lens on the SRV control switch signifies.
e.
Verification that support systems are operable in order to support ECCS components.
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Instruction as to the operation of the ADS logic after reactor vessel blowdown is accomplished.
g.
Instruction as to the meani-and reliability of the fuel zone level indicators during accident s ations.
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8.
Small Break Loss of Coolant Accident - System Considerations The inspectors reviewed system-related aspects of procedures to ensure that operator actions subsequent to a SBLOCA could be performed. System consider-ations in the following area were reviewed:
Instrumentation to carry out operator actions in the SBLOCA procedure.
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Understanding of power operated safety relief valve position indication including accoustical and thermocouple monitoring.
Equipment response to safety injection reset.
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Safety injection effects on containment isolation.
Real-time consideration of SBLOCA procedure actions, including adequate e.
time to remove one RHR (LPCI) division for containment spray / torus cooling.
f.
Instrumentation verified for environmental effects (for the conditions prevailing at the time of the accident), power supply (with loss of offsite power and a single failure in the most limiting instrument bus), and redundancy (in sensor and readout device).
No problems were identified with system considerations a through e.
Appropriate personnel were not available to discuss item f.
This item is considered open pending further inspection by the Regional Office (324/325/
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80-04-01).
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