ML20035F179
| ML20035F179 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 03/31/1993 |
| From: | Hopkins P, Lesser M, William Orders, John Zeiler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20035F176 | List: |
| References | |
| 50-413-93-07, 50-413-93-7, 50-414-93-07, 50-414-93-7, NUDOCS 9304210010 | |
| Download: ML20035F179 (23) | |
See also: IR 05000413/1993007
Text
{{#Wiki_filter:.. , . , [Ng UNITED STATES ,( f., NUCLEAR REGULATORY COMMisslON , REGION 81 y ).g g $ 101 MARIETT A ST RE ET. N.W. .g '" ^f AT L ANT A, GEORGI A 30323 , %...../ Report Nos.: 50-413/93-07 and 50-414/93-07 Licensee: Duke Power Company 422 South Church Street Charlotte, N.C. 28242 Docket Nos.: 50-413 and 50-414 License Nos.: NPF-35 and NPF-52 Facility Name: Catawba Nuclear Station Units 1 and 2 Inspection Conducted: February 7, 1993 - March 13, 1993 Inspector: / - Resid 3pector Dat,n Signed p W. T. Orders, Senio Inspector: , . & / s P. C. Hop' kins, Residen Inspec'tg D/teSigned p /\\. / Inspector: < (# J. Zeiler,' Resident nspectoVg fates'igned ? O i[- 3hl A3 V / ~ / Approved by: Fc4 M. S. Lesser, Section Chief Date Signed Reactor Projects Branch 3A Division of Reactor Projects SUMMARY Scope: This routine, resident inspection was conducted in the areas of review of plant operations; review of potential leak path outside containment; surveillance observations; review of Nuclear Service Water System inoperability; review of SSPS and ESFAS surveillance I test problems; maintenance observations; licensee event reports; and follow-up of previously identified items. Results: One apparent violation was identified and is being considered for escalated enforcement. The apparent violation involves the inoperability of the Nuclear Service Water (RN) System due to several RN pump discharge valves that would not open against pump discharge pressure (paragraph 6). 9304210010 930402 PDR ADOCK 05000413 G PDR _ i
. . 2 One violation with three examples was identified for failure to- follow procedures involving 1) the inadvertent actuation of the undervoltage relays on Unit 2 emergency bus 2ETB resulting in a B train blackout (paragraph 5.c), 2) the failure to drain the Spent Fuel Pool transfer tube prior to removing the transfer tube blind flange resulting in the loss of 6000 gallons Spent Fuel Pool inventory (paragraph 8.d), and 3) inappropriate operator actions during a pipe break event in the Residual Heat Removal System (paragraph 10.a). One Non-Cited Violation was identified involving the failure to implement compensatory actions prior to removing the B train Residual Heat Removal pump room equipment hatch (paragraph 8.e). Two Unresolved Items were identified involving 1) a concern over the timeliness of information handled via the licensee's Operating Experience Program (paragraph 7.a), and 2) the adequacy of Engineered Safety Features Actuation circuitry testing (paragraphs 7.a and 7.b). One Inspector Follow-up Item was identified involving the adequacy of the plant fire brigade staffing (paragraph 3.b). One licensee strength was identified in the maintenance area involving work activities associated with the repla:ement of numerous valves in the Nuclear Service Water System (paragraph 8.c). . . . . .
_ _ _ - _ _ _ _ _ _ . . REPORT DETAILS 1. Persons Contacted Licensee Employees S. Bradshaw, Shift Operations Manager J. Forbes, Engineering Manager
- R. Futrell, Regulatory Compliance Manager
- T. Harrall, Safety Assurance Manager
- J. Lowery, Compliance
- W. McCollum, Station Manager
W. Miller, Operations Superintendent
- M. Tuckman, Catawba Site Vice-President
Other licensee employees contacted included technicians, operators, mechanics, security force members, and office personnel. NRC Resident Inspectors
- W. Orders
- P. Hopkins
- J. Zeiler
- Attended exit interview.
Acronyms and abbreviations used throughout this report are listed in the last paragraph. 2. Plant Status Unit 1 Summary Unit 1 operated at essentially 100 percent power for the entire report period with no major problems. Unit 2 Summary Unit 2 began the report period in Mode 5, in day 9 of a scheduled 65 day E0C-5 refueling outage. On February 10, the unit entered Mode 6, when the licensee began detensioning the reactor vessel head. Defueling of the reactor core commenced on February 14 and was completed on February 16. On March 1 the licensee began reloading the core and completed that activity on March 4 without incident. On March 12, the reactor vessel head was set and Mode 5 was entered. Major outage activities initiated or accomplished during this report period included; Moisture Separator Reheater tube bundle replacement work, ice condenser work, valve maintenance and testing, steam generator eddy current testing, Diesel Generator A and B inspection and maintenance, and ECCS flow balance testing. _ _ - _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _
_ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . -__- . , 2 3. Plant Operations Review (71707) a. General Observations The inspectors reviewed plant operations throughout the report period to verify conformance with regulatory requirements, TS and administrative controls. Control Room logs, the Technical Specification Action Item Log, and the R&R log were routinely reviewed. Shift turnovers were observed to verify that they were conducted in accordance with approved procedures. The complement of licensed personnel on each shift inspected, met or exceeded the requirements of Technical Specifications. Further, daily plant status meetings were routir.ely attended. Plant tours were performed on a routine basis. The areas toured included but were not limited to the following: Turbine Buildings Auxiliary Building Units 1 and 2 Diesel Generator Rooms Units 1 and 2 Vital Switchgear Rooms Units 1 and 2 Vital Battery Rooms Standby Shutdown Facility During the plant tours, the inspectors verified by observation and . interviews that measures taken were proper and procedures were ! followed to assure that physical protection of the facility met current requirements. Areas inspected included the security organization, the establishment and maintenance of gates, doors, and isolation zones in the proper conditions, and access control badging. In addition, the areas toured were observed for fire prevention and protection activities and radiological control practices. The inspectors also reviewed PIRs to determine if the licensee was appropriately documenting problems and implementing corrective actions. b. Adequacy of Plant Fire Brigade Staffing In Inspection Report 91-26, the inspectors documented a review of Catawba's fire brigade staffing levels. The fire brigade staffing meet current requirements. However, the inspectors were concerned l about the adequacy of fire brigade staffing due to the identification of apparent deficiencies at other facilities in which the staff was under-manned to combat simultaneous operational transients and the activation of the fire brigade. At that time, a review confirmed that Catawba exceeded TS requirements and commitments for fire brigade staffing levels. The licensee's staffing plan included provisions to call in personnel to augment on-site staff and utilize personnel from the . .. .. _ _ . .. .
- - _ _ _ - _ _ _ _ _ _ _ e e 3 maintenance, chemistry, and IAE divisions at the plant in case of emergencies. This program was identified as a strength in NRC Inspection Report No. 413, 414/90-06. It was noted in a review performed during this report period that the total number of personnel qualified to a agment the fire brigade had decreased and that the majority of those qualified are working on the day shift. This could result in the operations staff being ill-prepared to handle certain operational / fire transients, especially on night shift. The licensee has realized that the issue needs to be re-addressed and is in the process of evaluating the issue. Pending completion of that review and the implementation of proposed improvements, this issue will be carried as an Inspector Followup Item, IFI 413,414/93-07-01: Review Adequacy of Fire Brigade Staffing. No violations or deviations were identified. 4. Review of Potential Leak Path Outside Containment (71707) During this report period, the inspectors were notified of a problem identified at Beaver Valley Nuclear Station involving a potential leak path outside containment during a LOCA due to the over-pressurization of a portion of the charging pump suction piping. At Beaver Valley, a check valve is located downstream of the VCT in the suction piping of the charging pump. The Seal Water Heat Exchanger return line is located between the VCT and this check valve. In the sump recirculation mode after a LOCA, with the RHR pumps supplying flow to the suction of the charging pumps, a relief valve, set at 150 psig, in the Seal Water Heat Exchanger return line could lift, if the aforementioned check valve leaks by. This relief valve discharges to the VCT. If the VCT were to become full, it would overflow to the Liquid Hold-up Tanks located outside containment. The inspectors reviewed P&ID drawings of the licensee's Chemical and Volume Control System and determined that the system was configured similar to Beaver Valley in that a check valve (NV-229) is located downstream of the VCT, and the seal return line is connected upstream of this check valve. Unlike Beaver Valley, however, the seal return relief valve (NV-222) has a relief setpoint of 220 psig. This is above the design shutoff head of the ND pumps, therefore, this relief valve should not open even if the check were to leak by. Although the specific Beaver Valley problem did not appear to exist, the inspectors identified two other check valves located in the charging pump suction line that would have similar consequences as the Beaver l Valley problem if they were to leak by. One of these valves, NV-233, is located in the Boric Acid Transfer Pump discharge line to the NV suction line, and the other, NV-234, is located in the Reactor Water Makeup Pump discharge line to the NV suction line. Upstream of both of these check l
. . _. -- _-- ___ i . . , 1 4 l , valves are relief valves with a setpoint of 150 psig. With the relief setpoint lower than the design shutoff head of the ND pumps, if either of the check valves leak by in the sump recirculation mode, a leak path to the VCT would be established. The inspectors reviewed the licensee's
inservice testing pregram and found that neither the check valves, nor i the relief valves, were currently being tested. , The inspectors notified the licensee of these findings. Since Unit 2 , was defueled and ECCS flow balance testing was scheduled during the
outage, arrangements were made to monitor VCT level for indications of j check valve leakage during this testing. On February 24, ECCS flow balance testing was performed on Unit 2; no significant VCT level l increase was noted, indicating that the check valves were not leaking ! by. By the end of the report period, the licensee was still evaluating , their long term corrective actions for this problem. The licensee is considering adding the check and relief valves to their inservice I testing program due to the valve's newly discovered safety function. The inspectors will continue to monitor the licensee's corrective actions for this problem.
No violations or deviations were identified. ! 5. Surveillance Observation (61726) a. General J During the inspection period, the inspectors verified that plant i operations were in compliance with various TS requirements. i Typical of these requirements were confirmation of compliance with , the TS for reactivity control systems, reactor coolant systems, ' safety injection systems, emergency safeguards systems, emergency i l power systems, containment, and other important plant support ' systems. The inspectors verified that: surveillance testing was performed in accordance with approved written procedures, test instrumentation was calibrated, limiting conditions for operation were met, appropriate removal and restoration of the affected equipment was accomplished, test results met acceptance criteria . and were reviewed by personnel other than the individual directing l the test, and any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personnel. b. Surveillance Activities Reviewed The inspectors witnessed or reviewed the following surveillances: ! PT/1/A/4200/01E Upper Containment Personnel Air Lock Leak Rate Test PT/1/A/4200/26 Containment Spray System Valve Inservice Test PT/1/A/4250/03B Auxiliary Feedwater Motor Driven Pump 1B Performance Test . , - wr ..iv- g y,y-- --p y,-w, . . . . -%., a www ,-w--,- -w-- y---
j - - l i 5 PT/1/A/4450/01C Auxiliary Building Filtered Exhaust Filter Train (IA) Test PT/2/A/4350/02B Diesel Generator 2B Operability Test PT/2/A/4350/10 Diesel Generator (2B) Operating Parameters c. Unit 2 Train B Blackout On the morning of February 21, Unit 2 was defueled in day 23 of a planned 65 day refueling outage. The B train D/G was out of . service, undergoing routine engine inspection / maintenance. ' At 9:57 a.m., an IAE technician was performing load sequence testing and inadvertently actuated the undervoltage relays on Unit 2 emergency bus 2ETB resulting in a B train blackout. The actuation was caused when the technician failed to follow the applicable procedure, IP/2/A/3670/018, Load Sequencer Timer Cal, , and actuated the load sequencer test box prior to placing the
sequencer in test. Since the B train D/G was already inoperable, there were no major loads being supplied by ETB and the effects o.f - the blackout were minimal.
By 12:30 p.m., 2ETB had been re-energized and aligned to normal. The load sequencer initiation constituted an ESF actuation which
resulted in a 4 hour notification and requires the submittal of an
LER. The inspectors reviewed the procedure which was being used at the , time of the event, and interviewed the personnel involved. The ' procedure requires that all preliminary requirements be performed before the sequencer is taken to the TEST mode which aligns the sequencer for individual timer calibration. The personnel involved had successfully performed all procedure steps through preliminary requirement 10.1.7. At this point, the personnel realized that the sequencer did not have control. power available - which precluded continuation of the test. It took approximately 1 1/2 hours for the operations group to return power to the sequencer. When the test personnel resumed his activities, he forgot to resume testing at the place in the procedure where he. had secured. Instead of resuming the test at step 10.1.8, he omitted steps 10.1.8 through 10.1.13 and began the individual timer calibrations which are detailed on enclosures to the procedure. Thus, when the technician energized the sequence test box, the sequencer responded appropriately by opening the incoming breaker. The sequencer would have been placed in TEST in Step 10.1.10. Had the personnel followed the procedural requirements, ' the event would not,have occurred. The failure to follow IP/2/A/3670/01B is identified as one of three examples which collectively constitute Violation 414/93-07-02: Failure to Follow Procedures. One violation was identified. . = - . . . _ _ _ . _ . . . . _ __
i I ! j . . l 1 i 6 i . < l ' 6. Nuclear Service Water System Inoperability (61726) j Event Summar_y: l i On February 25, 1993, Unit I was operating at 100 percent power and Unit 2 was defueled, in a scheduled refueling outage. At approximately 3:00 i p.m. that afternoon, the licensee was performing routine testing of Nuclear Service Water (RN) pumps IB and 2B when it was discovered that ' the respective discharge valves for those pumps, IRN38B and 2RN38B would i not open against pump discharge pressure. Since the discharge valves on the A train pumps, valves 1RN28A and 2RN28A are identical, the licensee concluded that they were most probably inoperable also. Accordingly, the licensee placed Unit 1 in the action requirements of TS 3.0.3 at
5:45 p.m. As corrective action, valves 1RN28A and IRN38B were modified such that they are 20' open when in the " closed" position. This modification reduced the differential pressure experienced by the valves l and consequently, the i.orque required to open them. After the t modifications were completed, the valves were successfully tested and declared operable. The unit exited TS 3.0.3 at 10:05 that evening. The licensee ultimately concluded that the torque switch settings on I valves IRN28A and 1RN38B had been decreased to a non-conservative value in August of 1992 when they were adjusted in an attempt to conform with the guidance of Generic Letter 89-10. The licensee also concluded that the torque switches for valves 2RN28A and 2RN38B had not been adjusted to a value of 3.0 in July 1989 when it was determined that a value of 1.5 was non-conservative. The licensee adjusted the torque switches to l the proper value, successfully tested the valves against pump discharge . ! ' pressure and returned them to service. Testing confirmed that all four of the valves had been incapable of ! opening against the discharge pressure of the pumps at various times , since plant start-up. This in turn asserts that both RN locps were j inoperable simultaneously for considerable lengths of time when one or
both units were ir. Modes 1 through 4. I Backuround: i The RN System consists of two loops, with two pumps in each loop. The { discharge of the two pumps in each loop form a header that supplies ' coolant to both units. These headers can be cross-connected. During normal operation, one RN pump is in operation providing cooling water to both units through cross-connected headers. Heat loads supported by the RN system include but are not limited to the following: l - Diesel generator cooling water heat exchangers ! - Containment spray heat exchangers - Component cooling water heat exchangers - Auxiliary feedwater, safety injection and centrifugal charging pump lube oil coolers - Auxiliary feedwater, safety injection, residual heat removal and centrifugal charging room cubicle coolers - _ _ . _ __ . . _ _ . _ . .- _._ __._., . . _ , . . -
. _ _ __ . . , a 7
The RN pump discharge valves receive an open signal from a set of contacts in the respective pump motor breaker. They open when the associated pump is started and close when the pump is stopped. The valves do not receive a direct ESF signal and they have no control switches in the control room. The valves themselves are 30" butterfly valves supplied by BIF/ General ' Signal Corporation. BIF supplied the complete vaive including the motor
actuator. When the valves were originally installed at Catawba, the valve actuators were set up with a torque switch setting of 1.5, the lowest setting. Valve IRN28A was adjusted to a value of 3.0 sometime in 1985 when it was found that the valve would not open. The circumstances , l behind this event are not known, but the other three valves in question were not evaluated at that time. In July 1989, as a result of PIR-0-C89- 0376 which documented the failure of another BIF butterfly valve to l open, the open torque switch setting for the three other valves in ! question were to have been set to manufacturer's maximum allowed value l of 3.0. The as found evidence of this event indicates that 2RN28A and 2RN38B were not reset. In August 1992, the torque switch settings for valves IRN28A and IRN38B
were reduced to 2.0 in an attempt to conform with the guidance of ! Generic Letter 89-10. The as found evidence of this event indicates that the engineering analysis which was performed to derive the torque switch setpoints was flawed, in that it was found that the valves would not open against pump discharge pressure. P The torque switch setting history for the four valves in question is as follows: 'orque Switch Setpoints 1RN28A 2RN28A 1RN38B 2RN38B 1976-July 89 3.0 1.5 1.5 1.5 July 89-Aug 92 3.0 1.5 3.0 1.5 Aug 92-Feb 93 2.0 1.5 2.0 1.5 Feb 93-Present 2.0 3.0 2.0 3.0 A review of this information indicates that all four of the valves were inoperable for extended periods of time when one or both of the units operated in Modes 1 through 4. During these periods of time, both loops of the RN system were also inoperable. Safety Sionificance: The licensee performed an evaluation of the safety significance of this event by analyzing the station's response to selected design basis events and by reviewing Catawba's PRA. The period considered was from August 1992 through February 1993. The time was chosen based on data which indicated that during that period, three of the four RN pump discharge valves (IRN28A,1RN38B and 2RN38B) were unable to open against pump discharge pressure. (Testing indicated that RN pump 2A and it's . .. -. ..
. - . _ . . -_ . j -
, t , I I i 8 l l r ' associated isolation valve 2RN28A may have been capable of operating with a torque switch setpoint of 1.5). The four design basis events considered were; f 1. Failure of the RN system with one or more units at full power. l 2. Loss of all offsite power (LOOP) with failure of the RN system. , ' 3. LOCA with failure of the RN system. 4. LOOP and coincident LOCA on one unit with failure of the RN system. i For the purpose of this evaluation, the licensee considers the " worst case" RN failure to be one that would depressurize the RN system such that the IA,18, and 2B RN pump discharge valves would not be capable of opening. Subsequent testing indicated that RN pump 2A and the , ' associated isolation valve 2RN28A may have been capable of operating to pressurize the shared RN supply header if one of the other three pumps failed while running. Therefore, a failure of RN pump 2A while no other , pumps were running would have been the worst case single failure. l ! 1. Failure of the RN System , During the period spanning August 1992 through February 1993, RN
pump 2A was operating with no other RN pumps running for a total ' of 998 hours. The licensee concluded that, if RN pump 2A had failed during this time, a total loss of RN would have occurred. The licensee assumed that the operators would have been alerted to i the event by annunciators in the control room. Operators would have tried to restore RN per AP/0/A/5500/20, " Loss Of Nuclear
Service Water," which would have instructed them in part to start i l the three idle RN pumps. RN pumps lA, IB, and 2B would have j started and dead headed against the discharge isolation valves ' preventing them from opening due to high differential pressure. j The licensee assumed that the operators would be aware that the i valves had not opened and that an operator would have been dispatched to the RN pump structure. The licensee assumed that the most limiting parameter of the event would have been an increase in temperature of the KC system which cools the reactor coolant pump motor bearings. The licensee concluded, based on equipment heat-up rates and the time required for the operators to manually align the valves, that RN could have been re-established before causing any further plant transients. l l 2. LOOP With Failure of the RN System l l During the time in question, the licensee concluded that if RN pump 2A had failed to restart during a LOOP event, a total loss of , ' RN would have occurred. Immediately following the LOOP, the main generator, the reactor, NC pumps, and turbine would receive trip signals. The four D/Gs i 1 . . .. ,- . .. -- - - . . .. ~
. . . . . . - _-___ --- . .
1 9 would have started automatically and would have powered their respective essential buses. Each of the three idle RN pumps would have started and dead headed against their respective isolation
valve, preventing them from opening, j The licensee concluded that for this event, the limiting factor j would have been temperature increase of the D/G engine cooling system which is cooled by RN. The D/Gs would have started and run , without RN cooling for a limited time until either manual trip or i engine failure. The operators are instructed by procedure to trip l the D/Gs manually after 2 minutes if RN cooling has not been '! established. l
The operators would not have been able to establish RN cooling to ! ' the D/Gs within 2 minutes. The licensee assumed that the operators would have evaluated the status indicators available in the
control room, would have recognized the loss of RN, secured the ' D/Gs and then addressed the resulting Blackout. l , The licensee concluded that if the operators secured the D/Gs, the reactors could have been maintained in Hot Standby until one or l more RN pump discharge valves would have been opened. Once an RN ! ' pump disciarge valve was opened, the associated D/G could have
been restarted. l The licensee also concluded that if the operators did not l recognize / react to the loss of RN, it is likely that the D/Gs , would have run until they failed. In this situation, the operators ! ' would have entered the loss Of All AC Power procedure and would ! have established NC pump seal injection from the SSF. In this ! situation, the licensee concluded that the units could have been ' ' mainta:ned in Hot Standby until offsite power was restored or l portable generators were brought on site. > 3. LOCA with failure of the RN system. f ! ! During the time in question, the licensee concluded that if RN i pump 2A had failed during a LOCA on an operating Unit, a total loss of RN would have occurred. The licensee assumed that the operators would have been alerted to
the loss of RN by control room annunciators, would have been aware l that valves 1RN28A,1RN38B and 2RN38B had not opened. In this situation an operator would have been dispatched to the RN pump structure. , The licensee concluded that the limiting condition for this event l is temperature increase of the KC system which cools various ECCS l l components. ESF response would occur as designed despite the loss of RN. The licensee concluded that increasing KC system , l temperature would not have presented a problem until ND was , ! l I l , .. . . . . - - . . - - - - - -- - -
.. . . 10 aligned for containment sump recirculation. There should not have j been any difficulty in restoring RN prior to that time. 4. LOOP and Coincident LOCA on One Unit With Failure of the RN I System. < The results of the licensee's analysis of this accident scenario > were not complete at the end of this report period and will be l documented in the LER. ! PRA Review: The licensee's engineering staff determined that this event placed the RN System in a degraded state and that, the core-melt frequency was
increased by a factor of approximately 2. This result may change based
on the licensee's ongoing evaluation of the coincident LOCA transient. ' Safety Analysis Conclusion: The licensee's safety analysis concluded that during the seven month period beginning August 1992 through February 1993, the CNS Nuclear , ' Service Water system could not have responded automatically to mitigate the consequences of Design Basis events.
'
This event is identified as Apparent Violation 413,414/93-07-03: - Inadequacy in Design, Engineering and Procedure Implementation Resulting ' ! in the Inoperability of the RN System. This violation is being
considered for escalated enforcement action in accordance with the NRC i Enforcement Policy. Accordingly, no Notice of Violation is being issued at this time. - 7. SSPS and ESFAS Surveillance Test Problems (61726) i a. Containment Spray ESFAS Test Inadequacy j On September 15, 1992, personnel at the South Texas Nuclear Station identified that the portion of the ESFAS circuitry associated with the Containment Spray System channel between the ' process instrumentation and the ESF actuation and logic instrumentation was not being tested properly. During the monthly ACOT for Containment High-High Pressure, the input relay test contacts for each channel were opened. Upon completion of testing these contacts were returned to the closed position. This testing did not verify continuity of the circuitry upon closure of the ' contacts. Failure of the contacts to close following testing may not be detected unless this continuity check is performed. The failure of these contacts would render the Containment Spray circuitry inoperable. Unlike the rest of the ESFAS which normally de-energizes to trip, the Containment Spray circuitry energizes to actuate. Because of this design, the vendor (Westinghouse) provided a separate test i t , - - , -, ..-r, , , ,,,-m - -r- -e ---e-r --v- , .,.e ,me-vn---~m--,.e --- -
. . l .
i ! i 11 circuit to verify continuity of the above-mentioned contacts. South Texas personnel did not recognize the significance of this test circuit, and, as a result, this continuity check was not being performed. On February 16, 1993, engineering personnel at Catawba were notified of the Containment Spray test inadequacies at the South ' Texas Station via the licensee's OEP. A review of associated test procedures was completed at which time it was determined that testing to verify continuity of the previously described portion of the Containment Spray circuitry was not being performed. l Testing was initiated to verify operability of the Unit 1 . Containment Spray channels. The results of this testing ' determined that all channels were operable. Since Unit 2 was in Mode 5, having previously been shutdown for refueling, the licensee planned to test the Unit 2 Containment Spray circuitry later in the outage. The licensee indicated that procedures governing the testing of the Containment Spray circuitry would be revised to incorporate the continuity check. i On February 17, while performing the Unit I continuity checks as discussed above, the licensee identified a problem with the test circuitry for two of the spray channels. Testing was modified slightly through the circuitry card in order to adequately conduct
the continuity check. Preliminary investigation indicated that
l the circuitry card for these channels may have been mis-wired. l This was based on discrepancies noted between the vendor supplied i i schematics of the circuitry and a visual verification of the cards. At the end of the report period, the licensee was discussing these apparent discrepancies with Westinghouse. Until the cause of the apparent mis-wiring on the cards can be resolved and reviewed by the inspectors, this issue will be carried as part of URI 413, 414/93-07-04: Review of SSPS and ESFAS Testing Inadequacies. < i During the inspector's review of the issue, it was also noted that i the licensee's Corporate Office received information regarding the circuitry problem from South Texas on October 23, 1992. The initial evaluation and determination of corrective actions by personnel in the licensee's OEP was completed on November 9,1992; however, it was not until February 16, 1993, that the site engineering personnel were forwarded information regarding the problem. The licensee's OEP program requires the evaluation and development of corrective actions, including distribution of the material to the group assigned responsibility, within 90 days. i The inspectors noted that once the information was received at the site, timely corrective action was implemented; however, the overall 90 day requirement was not met. The inspectors determined that further review of the timeliness of material handled via the ' licensee's OEP was necessary. Until this review can be completed, this issue will be carried as an Unresolved Item. This item is documented as URI 413,414/93-07-05: Review of OEP Timeliness. i ._ __ . - ___i
, __ _ _ _ l ' . .
I l 12 b. SSPS Train A, Phase B Isolation Test Problem l e During this report period, the inspectors were-informed that a
wiring error had been discovered in the SSPS test circuitry l associated with the Phase B Containment Isolation logic at the ' Braidwood Nuclear Station. Due to this mis-wiring, when the Phase B Isolation circuitry was tested, the Containment Spray Actuation j logic was actually being tested. As a result, the TS required monthly Surveillance for testing the Phase B Isolation circuitry was never properly performed. On March 1, the inspectors notified the licensee of this mis- wiring problem. The following day, the licensee completed their review of circuitry drawings and visually checked the SSPS circuitry on both units. It was discovered that the drawings for both units were correct; however, the visual verification determined that the wiring errors existed in the SSPS Train A ! circuitry of both units. The licensee entered TS 4.0.3 for Unit I allowing 24 hours to correct the wiring discrepancy and perform the TS surveillance for the Phase B Isolation circuitry. Testing was satisfactorily completed within the allowable 24 hours and the
licensee exited TS 4.0.3. On March 9, the Unit 2 SSPS Train B I circuitry was corrected and then tested with satisfactory results. i At the end of the report period, the licensee was discussing these wiring discrepancies with Westinghouse. Until the cause of the discrepancies can be resolved and reviewed by the inspectors, this issue will be carried as part of URI 413, 414/93-07-04: Review of SSPS and ESFAS Testing Inadequacies. , 8. Maintenance Observations (62703) a. General l Station maintenance activities of selected systems and components were observed / reviewed to ensure that they were conducted in accordance with the applicable requirements. The inspectors verified licensee conformance to the requirements in the following areas of inspection: activities were accomplished using approved , l procedures, and functional testing and/or calibrations were ' performed prior to returning components or systems to service; quality control records were maintained; activities performed were accomplished by qualified personnel; and materials used were properly certified. Work requests were reviewed to determine the status of outstanding jobs and to assure that priority was assigned to safety-related equipment maintenance which may affect system performance, b. Maintenance Activities Reviewed The inspectors witnessed or reviewed maintenance activities associated with the following maintenance Work Requests (W0s): i l J . . - - .- - -. - - - . - - . - ..
.. - i . . 13 , 92082533-01 Remove / Repair Flange Gasket on the Nuclear Service Water Header 93001434-01 Remove / Repair 4160V 2ETB Switch Gear Breaker (s) 92090884-01 Repair Diesel Generator Metering , 92070838-01 Inspect / Repair D/G 2B Frequency Module 92083918-01 Routine Testing of 2B D/G Relays 92088718-01 Preventative Maintenance on Limitorque Operator to Valve 2NV-872A , 92083935-01 10 Year Maintenance Inspection of Diesel Generator 2A 92082536-01 Replace Valve JN-53B 92072296-01 Perform ISI Hydro on B Train Nuclear Service Water Train System c. Nuclear Service Water Valve Replacements , During this report period, the licensee completed replacement of nine B Train RN valves and 12 A Train valves. The valves were replaced, in part, due to their excessive leakage rates; the new ' valves are fabricated of stainless steel with improved isolation capability. The B Train valve replacements were conducted between ! February 18 and 20, and the A Train replacements were conducted March 2 through March 4. Since the RN System is shared between units, Unit I was placed in a 72 hour action statement in
accordance with TS 3.7.4. , In order to accomplish the successful draining of each RN train, twelve teams, consisting of over 100 employees, were used to assist operations personnel. This maintenance was well coordinated and expeditiously completed. The inspectors noted that there was good pre-planning of the activity, with adequate consideration of contingency plans. The inspectors observed good j cooperation and communication between maintenance engineering, ' operations, and craft personnel. This activity is identified as a strength in the area of maintenance.
d. Loss of Unit 2 Spent Fuel Pool Inventory Incident On February 9, 1993, Unit 2 was in Mode 5, in day 11 of a scheduled 65 day refueling outage. In preparation for refueling, maintenance personnel were directed to remove the fuel transfer tube blind flange to allow filling of the refueling cavity. The refueling cavity and the spent fuel pool refueling transfer canal l are connected by the fuel transfer tube. The tube is fitted with a blind flange on the reactor building end and a gate valve, KF- 122, on the spent fuel pool end. While unbolting the blind flange, water leaked from the flange area indicating that the transfer tube was filled with water. This leakage was unexpected since KF-122 was previously closed prior to filling the spent fuel pool refueling transfer canal. l . . . _ _ _ _ - ._ ,_ _
, . - -. _ .. !
. . ! ! 14 Attempts to re-tighten the flange to stop the leak were unsuccessful. Efforts were immediately initiated to install the weir gate that separates the spent fuel pool from the spent fuel pool refueling transfer carmi. Within approximately 30 minutes i the weir gate was installed and inflated, which prevented further inventory loss from the spent fuel pool. KF-122 was then opened ' slightly and then re-closed. It was noted that more turns were required to close the valve than to open it, indicating that the i valve had not been fully closed. Leakage stopped following the , full closure of the valve. It was subsequently determined that approximately 6000 gallons was drained from the spent fuel pool, but the minimum spent fuel pool level as required by Technical Specifications was maintained. Since the refueling cavity drains were open at the time of the l event, water leaking from the flange drained into lower l containment. , During review of the incident, the inspectors noted that-the
removal of the blank flange by maintenance was initiated without ' operations personnel first performing procedure OP/2/A/6200/13, i Draining, Filling and Purification of the Refueling Cavity, Enclosure 4.13, Removal of Transfer Tube Blind Flange. The purpose of this enclosure is to ensure the controlled removal of j l the blind flange and requires that the transfer tube be drained of l water prior to requesting maintenance personnel to remove the '
flange via a predefined work request. l , l Due to a weakness in the coordination of work activities, i operations personnel authorized the approval to perform the work request for removing the blank flange without ensuring that this activity was adequately controlled. Had this procedure been l performed as required, this event should have been prevented. The failure to perform OP/2/A/6200/13, Enclosure 4.13 is identified as one of three examples which collectively constitute Violation 414/93-07-02: Failure to follow Procedures. e. Failure to Establish Compensatory Actions for ND Hatch Removal On January 28, with Unit 2 in Mode 1, operations personnel in the WCC responsible for coordinating maintenance activities, , authorized the removal of the concrete equipment hatch located i above the 2B ND pump. This hatch functions as a boundary for ventilation, radiation, fire, and flood protection. The hatch removal was planned in order to install pump / motor lifting equipment in the ND pump room. This equipment was to be pre- staged in the room to prepare for upcoming maintenance on the ND pump during the refueling outage that was scheduled to commente the next day. The hatch was removed at approximately 9:00 a.m. Shortly thereafter, it was recognized that a Compensatory Action was i _ _ _ . - - - _ _ _ . _ , . . . .. . - _ . _ . _ , _ . _ _ , . -
. - . . , i 15 required for the hatch removal due to the potential adverse effect ' on the Auxiliary Building Ventilation System. At 11:40 a.m., the hatch was re-installed. The licensee performed a past operability evaluation for the period during which the hatch was removed. The evaluation . included the results of subsequent testing to ' confirm that with the hatch removed, the Auxiliary Building Ventilation System was i still capable of meeting its design basis function of maintaining i' a negative pressure in the ND pump room. The results of this evaluation determined that the loss of boundary protection functions provided by the hatch did not degrade any safety related equipment or instrumentation. The inspectors reviewed the evaluation and determined that the licensee had adequately addressed the past operability concerns for this incident.
The inspectors reviewed Station Directive 3.1.14, Operability Determinations, Section 6.5, which provides guidance for ! implementing Compensatory Actions. Item 6.5.1 defines . Compensatory Actions as measures or actions taken on a temporary l , basis to replace an as designed safety function of a component or i ! I design feature such that the results of an FSAR Design Basis Event do not exceed the consequences stated in the FSAR. Based on this definition, a Compensatory Action was required for the removal of ' the ND pump room equipment hatch. This event is considered to be a violation of the requirements of , TS 6.8.1, for failure to follow Station Directive 3.1.14. After l review of the circumstances relative to the issue, the inspectors ' determined that the criteria specified in Section VII.B.(2) of the NRC Enforcement Policy were satisfied, in that, the violation was licensee identified, had low safety significance, and, appropriate corrective action was initiated prior to the end of the report period. For these reasons, this issue is documented as NCV 414/93-07-06: Failure to Implement Compensatory Actions Prior to Removing ND Pump Room Hatch. 9. Review of Licensee Event Reports (92700) l The below listed LER was reviewed to determine if the information I provided met NRC requirements. The determination included: adequacy of ! description, verification of compliance with Technical Specifications and regulatory requirements, corrective action taken, existence of l potential generic problems, reporting requirements satisfied, and the relative safety significance of each event. (Closed) LER 413/91-03: Technical Specification Violation for Emergency Personnel Hatch Being Unlatched Due to Equipment f ailure/ Malfunction. anuary 10, 1991, with Unit 1 in Mode 3, the Resident Inspectors discovered that the emergency personnel hatch between l l . _ _ - . -. - - . - - . . - -
_. . .
v 16 upper and lower containment compartments was not latched. TS 3.6.5.5 requires that this hatch be operable and closed when the j unit is in Modes 1-4. The licensee determined that the hatch may i have come unlatched due to ice condenser traffic in the area. The inspectors previously verified that licensee corrective actions t were completed to prevent recurrence of this incident. The latching arms were adjusted for proper hatch securement. A match , mark tab has been installed underneath the hatch handwheel to l serve as the location for placing the tamper seal in a more restricting configuration. Signs have been installed in the area warning personnel not to step on the hatch. In addition, a visual inspection of the tamper seal and hatch position was added to the , operator rounds for Modes 1-4. The licensee's corrective actions l were determined to be acceptable. ) No violations or deviations were identified. 10. Followup on Previous Inspection Findings (92701 and 92702) a. (Closed) URI 50-414/93-03-01: Review Residual Heat Removal Train . 2A Water Hammer Event. 1 During this report period, the inspectors completed their review ' of an event that occurred on January 31, while Unit 2 was in Mode 4, when operations personnel were in the process of placing the ND System in service. Just prior to starting the 2A ND Pump, a water hammer occurred causing a pipe break immediately upstream of a 3/4 i inch vent valve in the 2A ND suction piping. This resulted in the discharge of reactor coolant through the break into containment. l The break was isolated when operations personnel closed the ND ' loop suction isolation valves from the NC system. Operations later re-opened these valves twice for short periods during troubleshooting; each time the valves were opened, another water hammer occurred in the piping. The inspectors reviewed the results of the licensee's engineering l analysis of the event and determined that it was comprehensive and confirmed initial indications that the water hammer was induced by the introduction of hot NC water (340'F) in the ND suction piping when the system was operated with letdown established several l hours prior to attempting to place the system in service. l Pressure in the suction piping then decayed to the saturation pressure of the hot water due to a small, unidentified leak in the system. This caused the formation of voids in the HD suction piping, and the subsequent re-pressurization of the suction piping just prior to starting the N0 pump for the second time resulted in the first water hammer. Two subsequent water hammers occurred in the piping when operations personnel re-opened the ND loop suction isolation valves to the NC System. These water hammers resulted due to the rapid collapse of voids that were formed as a result of i the break. l
{ ! .. - _ . - _ _ _ _ -_ _ - . 4 <
._ _ _ . r . .
17
Based on the inspectors review of the operational aspects of the i event it was noted that the operators promptly recognized the loss of NC System inventory through the safety relief valve when the 2A ND pump was started. They then took effective action to secure , the pump and to close the ND loop suction isolation valves, thereby, terminating the event. However, the inspectors considered that the re-opening of the ND loop suction isolation valves, especially on the second occasion, was inappropriate, and represented an unnecessary risk to plant safety and a danger to i personnel that might have been in containment. This determination
was based on evidence that there had been control room indications that were not validated or monitored closely enough that could ! have been used to determine that a leak existed inside . containment, avoiding the need to re-open the valves. , 1 The inspectors identified two weaknesses, described below, that l contributed to the operation's staff failure to recognize earlier l in the event that a leak inside containment had occurred. 1. According to the operators, when the HD pump was secured, a momentary alarm indicating the opening of the ice condenser inlet doors was received. The operators reported that they i did not refer to the annunciator response procedure for this alarm since they believed that it was caused by someone i l inside containment who may have entered the ice condenser compartment, or, who may have physically bumped into one of the inlet doors. They recalled that several days earlier, i ' the alarm was received due to personnel opening the ice condenser access door. Since this alarm was not i investigated, the operators failed to recognize that it was legitimate and had annunciated due the increase in containment pressure when the pipe break occurred. i 2. The post-event analysis also indicated that the Containment Floor and Equipment Sump increased noticeably (from 8 to 11 inches) when the ND pump was secured, as well as, when the valves were re-opened. This increase was not observed by the operators. The operators reported that they were not monitoring this parameter closely due to shift turnover information that indicated that personnel might be in containment washing down equipment, rendering the indication unreliable. The operators failed to verify whether this activity was on-going; if they had, they would have discovered that no one was in lower containment. It was not until after the second opening of the valves that this information was obtained. OMP 1-8, Authority and Responsibility of Licensed Reactor Operators and Licensed Senior Reactor Operators, Sections 8.2.A.1 and 8.2.A.2, require that Senior Reactor Operators shall keep themselves informed of the plant operating status, and shall ., .__- - - -. ... _ _
_ ' s
o ! ! j control activities to insure safe, efficient operation of the l unit. Contrary to this procedure, on January 30, 1993, the Unit 2 SR0s 7 were not knowledgeable of the unit status and failed to control
- activities to insure the safe and efficient operation of the unit, in that, during a pipe break transient in the Residual Heat ! Removal System, the SR0s failed to validate or adequately monitor indications substantiating a leak inside containment, and i repeatedly re-opened isolation valves resulting in a loss of l reactor coolant into containment. ! This is considered to be a violation TS 6.8.1 for failing to ' follow procedure OMP l-8 and is one of three examples which collectively constitute Violation 414/93-07-02: Failure to Follow Procedures. b. (Closed) Violation 413,414/92-22-01: Programmatic Breakdown of Equipment Control Process. . The Licensee responded to the above violation by letter dated November 11, 1992. The Licensee has taken numerous corrective actions to rectify the problems identified in the violation. The inspectors have reviewed Operations Management Procedures, Administrative Procedures, Training Procedures and Records, Containment closure and Containment Integrity Procedures and Key ) Control Procedures to insure that the Licensee has completed the corrective actions committed to in regard to this violation. This item is closed. c. (Closed) Violation 413/92-18-02: Inappropriate Operator Response to BDMS Alarm Actuation. The Licensee responded to this violation in letter dated October 1, 1992. OMP l-8 Authority and Responsibility of Licensed Reactor Operators and Senior Reactor Operations, Section 7.2.13, describes the responsibilities of the 0ATC. Step 7.2.B.9.C requires that the 0ATC verify that the appropriate automatic actions for an alarm have taken place prior to taking recovery actions. On the day of the event, the Operator believing the Unit 1 Train A BDMS alarm to be spurious, acted peremptorily to restore normal suction to charging pump 18. The 0ATC did not immediately refer to the appropriate annunciator response procedure to verify that automatic actions wen complete. Consequently, the BDMS actuation , caused the suction source of the centrifugal charging pump to swap i from the VCT to the FWST. The 0ATC closed the suction valve from the FWST but did not reopen the suction valve to the VCT. The licensed operators have received Simulator Classroom instruction and active simulator training on the proper response to BDMS alarms and recovery actions. Appropriate procedures were enhanced to insure their ease of use. The inspectors have ,_ _- _ , -. .- . .-.- , - . . . - .
4 i i ) . . , ' ! l l 19 , , verified that appropriate actions have been taken or being taken to insure that recurrence of incidents of this nature in the future have been greatly minimized. This item is closed. ' 11. Exit Interview The inspection scope and findings were summarized on March 15, 1993, with those ,oersons indicated in paragraph 1. The inspector described , the areas inspected and discussed in detail the inspection findings listed below. No dissenting comments were received from the licensee. The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspectors during this inspection. ' Item Number Description and Reference IFI 413,414/93-07-01 Review Adequacy of Fire Brigade Staffing (paragraph 3.b). l VIO 414/93-07-02 Failure to Follow Procedures (three examples) (paragraphs 5.c, 8.d, and 10.a). l Apparent Violation Inadequacy in Design, Engineering and i 413, 414/93-07-03 Procedure Implementation Resulting in l Inoperability of RN System (paragraph 6). i l URI 413,414/93-07-04 Review of SSPS and ESFAS Testing Inadequacies (paragraphs 7.a and 7.b). l URI 413,414/93-07-05 Review of OEP Timeliness (paragraph 7.a). NCV 414/93-07-06 Failure to Implement Compensatory Actions Prior to Removing ND Pump Room Hatch (paragraph 8.e). i 12. Acronyms and Abbreviations ACOT - Analog Channel Operation Test BDMS - Boron Dilution Mitigation System CR - Control Room i D/G - Diesel Generator ECCS - Emergency Core Cooling System l E0C-5 - End-of-Cycl e-5 ' ESF - Engineered Safety Feature ESFAS - Engineered Safety Features Actuation System FSAR - Final Safety Analysis Report FWST - Refueling Water Storage Tank IAE - Instrumentation and Electrical IFI - Inspector Follow-up Item IP - Instrumentation Procedure KC - Component Cooling Water KD - Jacket Water Cooling (for D/G) LER - Licensee Event Report LOCA - Loss of Coolant Accident - - --- - - -- ,
_ . . o 20 Loss of Offsite Power LOOP - NC - Reactor Coolant System NCV - Non-Cited Violation ND - Residual Heat Removal NV - Chemical and Volume Control 0ATC - Operator at the Controls OEP - Operating Experience Program OMP - Operation Management Procedure OP - Operating Procedure P&ID - Piping and Instrumentation Detail PIR - Problem Investigatien Report PRA - Probabalistic Risk Assessment PRT - Pressurizer Relief Tank PT - Periodic Procedure RHR - Residual Heat Removal RN - Nuclear Service Water System R&R - Removal and Restoration SSF - Standby Shutdown Facility SR0 - Senior Reactor Operator SSPS - Solid State Protection System TS - Technical Specifications URI - Unresolved Item VCT - Volume Control Tank WCC - Work Control Center i f
- _ .- - _ _ _ . - _ .. . __ _ _ , f + e* D!CLOSUn.E 3
ms . . red ==f meefsen / Vol. sr No. us l Phr idy to Isar / Notacne - . . anosamens seed ooms==:s iac The secretary of the -- ui c Nacieer Ragnistory C
- _
Washlagton.DCatss& ATTN: ,
Dockettag and Servios Branch. l Hand deuver aa====ts tac One h%te j Mint North.11568 Rockvtus PLha. i RockvtD MD meewesa P.43 am. to 415 '
pm Fedwalwwkders. ' Copus of === may be esammed . at the NRC Pubhc ht Roost 213 1 L Strost.NW.(Lawer Level). . Washe.geon.DC . com mamanasseeunoscouracr: Jame IJoberman.Derector.Osce of ' .
EnforcumsenL UANesteer Rerulatory Conh Washisstest DC 30565 ' (301-eD H rn). euenasmsvary swoonanos Backposed - The NRC's cumma poucy en 2 enforcement manierences to addressed in Section V et the laissa reveales to the " General Statommet of and . !
- Procedure for Enducement i (Enforoneses Petsylis CFR L C est was en ' F 1&1988(Er 5F91).The amferossent senses that. eeAorceanent arGlast 4 normour he eventoihe puhuc.- , i Howwer,to Csanission has decided j to impleases a trialpreyam to I deterades whetherto meestain the - omvent petsy weetregned to
j enferommest emessenes = m edopie < l Type-Year Tytsipropus ter new peacy tot usand aBoa meet enfarossent seudurances to ba spes to - Cenessebug Open totersammese artsedamos by e5 meenbers of the publ,c. Ceresrensens peasy gesamment PeEcy 9tseenset I ammett Nesseer Rapletory i ch 7tiessJes ! Time NRCis luglementies a rwo yeat l 8'I*I8F'" # 83'" ausenaevtThe Mestser Regalatory obsereseen of enforcement co--< a.,(NRClisIsselag tis paksy The K we mor the etetsm ent en e s W u ties of a propen anddetendas wheter to two reer trial preyes te eBow seiocad esembash a perummums poBey for enforesumes osoferososo to be opea es % enderoament sthace by almembers of the confereatne sa as eacessment of l gamersi pabhc.TMs poucy statament ik % crnmas: descrthes en w stat proya, (1)Whosher thefact that the and taforums the pubbe of bow te get conde.sans was span impacted the I inforunstsen em upsetas epea NRC's abdity to esaduct a meanmsful enfarossnest -- con 8eresse endler tuplement the Mtc's =
savemM Wel is efectrve sa N l July 10 isE comments on the t=W)Wheeer en open conference (2 program are being reca(ved.$dmit the housase's parecipetion m ,' commana= en er before es cosapiedes th* *'*I"'"' of the artal propen scheduled for July (3) Wheder the NRC +-; :"'d e 1 11.1seL Comssents received after this air =iba=# assoma of resources m l date wtB be comendered ifit is pesetical i to de an. but the P-maion is abb to making the confwencs public: and ,, a ssure come6dersues only for cAmmaann (4)The estaat n!poblic interest :n , received as or belere this date. opening the enforcement conferer:ce. r ' , - - - - , , . - - , . , , . . . . , , , -n,--,,, ,,,-,..,,,,,.,,n,.,.n_,, ..,--,,c-----,a-n. .,,,n--a.. ,n,,,--
- ~ . s
. fedend Essister / Vol 57. No.12s / Frid=y kly 10.1m2 / No6ces 3m3 j ' n mru rar s e.mac op.m e m w _- w d a ,-eese mane n.bsci = p- =a scree e.u ensemens Camissememo r-- cial rescars, e!ges, bessers, pesence. eec. rex larrer ' NO
- "'
1'* 18' N - and &a.:
- dorcement coc!crtname wd not be
wG eest of the req rypes of d:s: creve persons saay be rven ed. m to h pd4ac J h h t!ccesees. Ead reposal eSco erg coczme io ben ea e ndw:1 Se enferoement coderwace (1) W be taimE asunst sa 5.AnemoamchesOpes h F'"****f" na somedamos mLk repor.al Mr.a.l., or d b acoco bash not h preetce. The acts comema codemco taket (gam.st a.n edmdual, t.:r::e os As e a as it la M 6at an wd omesme e be a mesong berwiega whethat a.n indivti::.al haa &-W es!worment coeJarance wCl be open to S NRCsad $a h W h h ps W h anoa es E d 4 eafwm-., cembrance is open for a, i emi p.3er,3,g pebe observsuse, et is not opes for a faI!ures erhart the NRC La.s ned ance M paMc paractpensa. that the indlvidualls) Lerched observanos as part cJ the esency's trial p ,,,,,en ,,g ,g op. ,.go,,, ,,,, gg pecgruxa and W h Ecocan a m d conimmase ese retaded set 0) Se UIIs
- the !Lv!bse of en h1tC this redered Reg 6meer actics that ottitnes
,ppef,,, ,gg g. ,e w g: ,p,, [[yg,,, g the i,-. f t==== vre be asksd * edmeesset osebronces are embrect to g "Em8 te futhermere med may be embre:s u Pnvecy Act trJorina coe. or ether brtng to k en----. Warums change pnar to any ressluaq . I idor nanos wt.ich codd be m aMaasd so that the NRC can schedule es edorament acuss and (1) Se appropr a sized confance ran statements of views or axpretuons of
ente enen Lrdytag The NEC aies se@ eppropness optanos made by NRC menpbyees at , , 4 medied mis.ad=inkt sec'es or $tste liaison cScers that en open edweement conferences or 6 5 " " _ _ _"" g open aassalat edermm=an comfaremos has teen bd bred, are not in*==in4=d w _ j
- **
- d and thatitis opes to pec
represset flas! determMatsaus or behafs. - ! k' "w' g ' j 1-i.ndua.Ts sam,A la addi$on. ea,os.o.t o. er.sose wn - be ,eo,ce,,es,t o,,,ds to annou.nce opeee ,ema The NRC Inten see-cn m ,,,some . > cpen to the Petes if the d-. w1E ncM* st lesat to worbes drys le '"'" 'E*P'" ' ! 1 be conduceed by talephone or the advtace of the enferr=mmet conference wG be prmied as opportsaky to codersace wtB be conducted et a g, % h suben written a===.-aw anocynounty 6 2 .x , g a se petge ee- s mee --- l r,tsevety ens 41ticensee's fee.lity. - ,
m.o,. .,d e a,,,.vd of wG , = be forwerded to the . ' ' = arie=== par and Dtreemt of the OlBoe of ht fw 'f2j Tonese C3)Tou-free doctronse beures board t'"'" **d 888d08'*80"*- _ ,. ! r_ucuere IXtector for Opersocos. i anforcement corJersocne wm not be
open ta the puhthe ta spec!al cases j wbare good oness has been shows eher d"; of the tol4se Desed as Asshden. ndEL shes te day of My test i M~*' the besett of public meseng e ersesses, the pubte mey ceg Per the Heelant Assenmeery r'-- observanos apsfast the petaatiallaps pot)ess.4rts to obtala a recordes of
! on the agenc(a enderzament ac5aa to a speamteg opes sederasseemt a====81. Oe, j particular casa, confer ===== The NRC wtB leone seeiher secrosary e(she oisundeesse, j Tie NRC wtD strtvs to rarM open r,deral Regioeur mothee efter the tolMee pt Dee, es.ssses Pted F-e ea: tes us.1 j enforcement confervoces detne abe mese g eynemes e,e asubashed. -,ma ,,,w 'wo-res.r anal prog-sea m accordsmas To assist to NRC he seektet < ! th the foDoeruns thre* sods a, to ensagensets to support ! ') Appr ~d? 13 Percer.t of al pu oboerrsches of andepoemmet g % enforomment cadenaces amane nees, enewediesle innernmeed ta 4 a parnenter enfarr====* conw by the NEC wG be open for attendtes.e shodd.oie se meseed Correctl0nB 'edad ***" casse ! pusc ob.meo o; (2) At tasat one spea adorusseset identafled la the no6ae Vol. F. No. tJe
codervoce en be hwiin aesh of ===a-aesag the apes Frwy, My tr. test 3 the onal omces; and ccederemos as later eaa tre busiases 1 p) a en!orcement codursecas dare prior to the enforemment _ l mt! be condweted w th a vndety of the M r., leuCLEAR Mt00LATORY f types of beansees. F ofOpes W Cotabases0It . To :,o4d powieca! bias in Ibe Camferename ' j selectoo procese and to arte:npt to meet T a wrewr==* a - wa rr.er.c os. a s th,,e ,oe a etaiad a .ve. ferunas enforcement conferences will conthese Candhe *iG Open Ei - . fourth alagib&e enferomment con Cordersacee;posey staternent 1 to ruw=nau be held at ee NRC mvolving oe. of ths,e cuerones of r a . licensees wG marimWy be open to the emers. Members of the pubbe be p pubthe dsrtag ebe tnal proyam. a.llowed access to the NRC reponal is aptles du==.=nt F2 A W 4 However,la cases where there is en omans se attend opea adoreemaan
- Page am's2 ts 6e W d
. or,goint edNdscatory proceedtes udth confermaram in accordamos wrth b one or amore isrterrenors enfor==at -Standard Opersting raw for leir.M 18K en page W tn ,
- "'=d cohma, ander sans.
' ! codervoces involving tarves r:1sted to N a, A.,m, Sepport for NRC nathe Ehh hae M tt W E W the subhet matter d the oopalag Heat And Meennsa pubtshed j ad}udicanos may sloo be opened.For Nove L 1971 (36 F1t M:31).Thrge reed *)ety 11.19M". > the purposee of this trul program, the procedurve provide that visiters may h* es.ues sees imune ** 4 i r --- - - - - - - - - . - - - - - }}