IR 05000333/1990006
| ML20058B751 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 10/11/1990 |
| From: | Meyer G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20058B712 | List: |
| References | |
| 50-333-90-06, 50-333-90-6, NUDOCS 9010300412 | |
| Download: ML20058B751 (29) | |
Text
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. U.S. NUCLEAR REGULATORY COMMISSION l Region I l Report No.: 90-% Docket No.: 50 333 License No.: DPR 59 Licensee: New York Power Authority P.O. Box 41 Lycoming, New York 13093 Facility: James A. FitzPatrick Nuclear Power Plant - location: Scriba, New York Dates: August 12 through September 22,1990 Inspectors: W. Schmidt, Senior Resident Inspector R. Plasse, Jr., Resident Inspector - N Approved by: ' . lenn W. Meyer, Section hief, Date Projects Section 1B, DRP INSPECTION SUMMARY This inspection report discusses routine and reactive inspections of plant activities durmg dcy and backshift hours including: plant operations, r adiological protection, surveillance and maintenance, emergency preparedness, security, engineer!ng and technical support, and quality assurance and safety verification.
INSPECTION RESUITS The inspector identified a violation concerning ine ective corrective action to an unphi:.ned rr isolatior. of RWCU and identified a non-cited deviation regarding ADS accumulator pressure. An Executive Summary and an Outline of Inspection follow.
9010300412 901018 PDR ADOCK 05000333 O PDC f ' .
. -. , . g -. - i ! i; EXECUTIVE SUMMARY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSPECTION REPORT NO. 90-06 - Operations: Operations continued to perform well during plant m 1utions.
Operator control during the B recirculation pump runback and the ripair of the - #4 TCV was effective and resulted in return to full power operation without c incident.
L _ Radiological Protection: The inspector closed a previous item concerning the maintenance of the PT$3 system.
L Surveillance and Maintenance: Adequate compensatory measures were taken to prevent additional failures of an expansion boot during periodic testing of the < ' ESW system.
NYPA's declaration of the B and D EDG inoperable due to a high vibration alarm on-the ventilation fan was conservative.
- Emergency' Preparedness: NYPA conducted a practice emergency drill, which met its objective.
Security: Adequate performance continued in the security area. The inspector - reviewed the lighting of the protected area fence in the evening and found it acceptable.
Engineering and_ Technical Support: The _ voluntary. shutdown to resolve calcula- ~ !. tional concerns ini the ldPs exemplified a conservative safety perspective, f NYPA was pursuing Generic letter 89-10 concerns _ over the ability of MOVs to l - function under design basis conditions. The SPDS was found to be adequate. _ A L.
non-cited deviation was identified-dealing with' the inability of ADS to have ' performed as described by NYPA in submittals r(quired by the TM1 Action Plan.
The inspector found NYPA's corrective actions in response to the -identified , errors in feedflow calibrat_ ion to be satisfactory, On the steam leak detection - system, the inspector _ identified a, potential design problem regarding the - absence of. alarming of the loss of control powe_r for the circuits, an apparent Technical Specification error regarding - numbers of.instrut,sents, and concerns regarding drawing references.
- Safety: Assessment / Quality Verification: The inspector identified - a violation Tor ineffective corrective action and an unresolved item for an improper method-of' correcting a drawing error regarding - an unplanned - RWCU 11 solation (LER . 90-21). The inspector concluded that LER 90-21 did not correct' problems _ con- ' cerning non-adherence-to tagout procedures, poor work practices during instru-U - ment isolation, inappropriate use; of design drawings, and poor communication.
D Based on review of LER 90-012-01, _the inspector identified concerns 'in the safety _ evaintions for degraded ESW components, , < , a.'.. m_
.. -. h OUTLINE OF INSPECTION JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSPECTION REPORT NO. 90-06 .. 1.
Operations (MC 71707,93702) 1.a Operator performance during plant evolutions.
~ 1.b 'NYPA corrective actions in response to B recirculation pump runback. ' 2.
Radiological Protection (MC 71707) .+ '2.a (Closed) Unresolved Item 89-08-01.
Adequacy of maintaining PASS
- operational.
3.
Surveillance.and Maintenance (MC 61726,62703,92702)
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3.e-Expansion boot rupture during performance of ESW surveillance test.
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3.b'. Declaration of B and D EDGs inoperable due to high vibration alarm on diesel ventilation fan.
3;c Review of NYPA's control of test procedures from the control room.
. . Resolved Item F-2 of Inspection Report 88-05.
3.d Repair'of #4 TCV.
4.
Emergency Preparedness (MC-71707) s 4.alNYPA practice drill.
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. Security (MC71707) l 5.a - Routine walkdown of. protected area fence at. night, p ~ 6.- EngineeringLand-TechnicalSupport(MC 90712 92700, 92702, 71710) ' '6.a Voluntary plant shutdo_wn lto ' resolve calculational-concerns in the ' E0Ps. - 6.b (0 pen): Generic : Letter 89-10: NYPA -action ;in response; to generic, . concerns in determining MOV performance _in design basis conditions.
, ..6.c Capability-: of RHR i heat : exchangers; to. handle 1 the a flow: of both LRHR L - .- - pumps.
F-1. ' y ' 6 d '(Closed) TMI Action Plan, Item _I.D.2;: Safety Parameter Display System , C" and Generic: Letter 89-06.
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N -. [ [ Outline of Inspection (Continued) L r 6.e (Closed) _ Unresolved Item 90-04-06; Review of LER 90-19 for ADS . L - accumulator check va he design error. Non-cited Deviation.
6.f (Closed) Unresolved Item 96~11-01; NYPA completion of calibration of process computer feedwater flow evv;;. 6.g- (Closed). Violations 90-11-02 and 90-11-04; NYPA to complete correc- -tive actions to prevent exceeding 100% core thermal power and correct c flow transmitter errors.
" 6.h NYPA letter on three phase bolted short issued which completed a commitment to submit a supplement to LER 89-12. Resolved Item F-4 of Inspection Report 89-12.
,' 6.1 Inadequate monitoring for loss of power on -steam leak detection [ circuits.
F-2 Apparent TS error for numbers of instruments.
!! L 7.
-Safety Assessment / Quality Verification (MC 30703).
7.a Inspector review of LER 90-021; RWCU automatic isolation when I&C-technician lifted leads to support ~ maintenance.
Inadequate verifica-tion of as-built drawing and use of ECN to' correct errors found out- ' side a. modification; Unresolved Item 90-06-01.
Inadequate review to determine causes'and corrective actions; Viol tion 90-06-02, 7.b (Closed) Apparent ~ Violations 90-04-01 and 90-04-02, Violations withdrawn.
, 7.c- (0 pen) Unresolved Item 90-02-06-Inspector review of-LER 90-12-01; ! Supplemental LER discussing results of NYPA's evaluation of: ESW-deficiencies.
' iI 7.d (Closed) Unresolved Items 89-11-01 and 90-02-06; Consolidation of TS.
' issues -which need further review when proposed amendment is submit = g ted.-- Resolved Item F-1 from Inspection Report 89-09; F-3 g{ -8.; 'Other Inspections and Enforcement Conferences ! 9. - Exit Interview- .f Attachment A - Acronyms-L s . S, o, t ]' [.:r- ' vl
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i t l i DETAILS ! 1.
Operations The Unit operated at rated power until August 21, when NYPA completed a I voluntary reactor shutdown af ter discovery of nonconu rvatisms in their i calculation used to determine the E0P primary containment pressure limit (PCPL) curve. On August 28, upon completion of a modification to upgrade the primary containment vent and purge valves to meet the requirements of ! , the PCPL curve, the unit was returned to service (see Section 6.a).
On' ' , [[/ August 30, the Unit was limited to 90*4 power by the.failing closed of the .
- 4 turbine control valve (TCV) due to a servo valve failure (see Section
! r 3.d).
On September 4, the servo valve was repaired and the unit returned i to full power. On September 4, the B recirculation pump ran back to mini- ! ,
mum speed because of a failed limit switch on its associated discharge valve.
The unit was maintained at 36*4 power during evaluation of the cause for the runback (see Section 1.b).
NYPA returned the unit to full power operation on September 5, and operated there throughout the ' remainder of the inspection period, a.
The inspector found that the operators performed well during start-ups, shutdowns,. surveillance testing, and plant transients.
Opera-
tors were. attentive, to duty and maintained a questioning attitude during non-routine' evolutions.
' b.- -The control room operators performed well when the B recirculation i pump experienced an automatic runback to minimum speed because of a failure of ~ its discharge valve open. limit switch.
The low speed limiter functioned ~ as designed to limit the 'scirculation pump to . minimum flow with-either the discharge valve not fully open or feed- , > ' water flow less than'20*i.- Operator ' actions included control rod -insertion -and A recirculation flow; reduction to. preclude. cperation in theJ instability region, in
accordance with Reactor' Analyst. Procedure (RAP) 7.3.16, Plant Power > Changes.
NYPA maintained. power at 36% while troubleshooting the cause of.the automatic runback, ' g ' The inspectors found that NYPA operators, engineers, and maintenance personn'el performed a co.rrect-and thorough evaluation of this prob- .i l em.- -A temporary modification, including 10 CFR 50.59 safety evalu. , ation, installed a jumper to disable the failed limit switch.- The'
safety-. evaluation' documented that the purpose of this interlock was L" . solely for pump protection and that if. the. discharge valve closure -
occurred,~ the. recirculation MG set would trip when the valve reached' 10*4 open. In addition, NYPA. performed MOV current trace diagnostics ' .to_ verify the. discharge valve was open.
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,v . i ~ v NYPA completed temporary changes to the recirculation system oper-ating procedure (OP-27) and annunciator response procedure (ARP) 09-04-03-39 associated with the B recirculation pump low speed limiter to reflect the disabling of the runback.
H 2.
Radiological protection a.
(Closed) Unresolved Item (89-08-01): The use of the post accident sampling system (PASS) during emergency drills indicated that the system was being maintained properly to perform its intended func-tion.
The PASS operating procedure (PSP)-17 stated that the PASS system would be used at least quarterly to take one of the required samples (i.e., reactor coolant, primary containment atmosphere or secondary containment atmosphere) on a rotating basis. The inspector has observed that NYPA conducts these sampling procedures in conjunc-tion with their quarterly emergency preparedness drills. The inspec-tor closed this item.
3.
Surveillance and Maintenance a.
NYPA determined a potential water hammer problem in the ESW system resulted in an expansion boot rupture and took appropriate short term actions. While performing the newly enhanced ESW check valve testing (ST-BR ESW Check Valve Test (IST)) for the second time, the expansion boot supplying water to the control room chiller condenser (70RWC-2A) ruptured, and some water sprayed through the cracked boot..This expansion boot had ruptured and was replaced during the previous per-formance of this ST.
Following this rupture NYPA determined the-potential cause was water hammer due to partial ~dr41ning of the line to-the chiller condenser af ter ESW was secured. Tt ts problem was not identified previously because prior to the der "opment of the. ESW concerns, this portion of the system was not in sd with ESW flow.
NY>A ~.nade temporary changes. to the ST and operating procedures as shoi t term actions. The ST was changed to sr scify isolation of the subjact coolers prior to' starting of the ES) pumps and then throt- ' tlit:g-open of the blocking valves to minimize the water hammer effect during testing.
The operating procedure ~ for ESW (OP-21),- and abnor-mal-procedure for Loss of RBCCW (AOP-11) were changed to require that an operator monitor the chiller condenser lines forf an expansion joint rupture following initiation of ESW.In the event of a rupture - the procedures; direct the operator to isolate the chiller condenser . and supply ESW : directly ton the control room air handling units.
- These compensating actions were found to be adequate by the inspec-tor. HYPA was continuing to evaluate-a long term solution to the, ' ' water hammer. problem.
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On September 18, NYPA conservatively declared the B and D EDGs
inoperable due to a high vibration alarm received on the associated ' diesel room ventilation supply fan (92FN-10).
Performance engineer- - ing determined, by local measurements, that the f an vibration read-
ings and fan differential pressure were comparable to the other ! diesel ventilation fans.
NYPA concluded that tne vibration alarm , actuated either because of a faulty vibration detector or slight damper oscillation.
NYPA declared the B and D EDGs operable on > . September 20.
Performance Engineering was continuing to determine . L[ whether the ventilation system was performing properly and inclued ' '
these fans in the routine vibration monitoring program. The inspec- ! tor found NYPA's actions adequate.
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On September 19, the inspector observed that I&C technicians properly
functionally checked several of the reactor water cleanup (RWCU) high ' area temperature switches.
This testing was conducted in accordance - with Instrument Surveillance Procedure (ISP)-49.
The technician, ' ! ' observed locally at the 9-21 panel, had a copy of the procedure for reference, and was being directed in his actions by the-test con- ! troller in the control room area.
The inspector believed that this
method of test-control was effective, based on this and other pre- . vious observations.
The inspector considered that this closed item 'l F-2-from Inspection Report 88-05.
f ,m d.
On September. 4, NYPA completed repairs to the #4 TCV. Power was pre-viously limited to 90% due to the failure of #4 TCV to open due to an inoperable' servo valve. NYPA' replaced the servo valve without inci-r dent and operators. returned the Unit to full power.
NYPA detailed y preplanning was evident and contributed to' this smooth evolution.
' 4.
Emergency Preparedness , . . i a.
On September'21, _NYPA conducted an emergency preparedness drill. The , inspector observed portions of the TSC manning evolution and found' t
that.it was properly conducted.. The. drill included a station black-
[ out. and to the extent possible : the, conditions-of a ' blackout were ' ' -simulated (i.e., loss of-normal 11ghts', loss of Gaitronics), i ' a c ~5.
Security ' v '
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' The = inspector ' performed a routine walkdown of the ~ protected areas i t . fence during the' late ev'ening.
No.' discrepancies were noted. Exist- ' ing lighting ' was adequate and operational and routine security: , patrols were in progress.
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Engineering _and Technical Support i
a.
The inspector determined NYPA's actions in handling potential errors L in the E0Ps exemplified a conservative safety perspective.
On . August 21, NYPA initiated a voluntary reactor shutdown while resolv-U ing calculational concerns with the E0P primary containment pressure limit-(PCPL). Previous analyses of the drywell vent valves had used ' incorrect design assumptions. NYPA concluded that the errors could potentially impact the operators' ability to open the drywell vent i butterfly valves during very low probability severe accident scen- ' arios.
The shutdown was voluntary because the drywell vent valves would have operated under all design basis conditions.
The PCPL curve was based on the ability of the drywell vent valve to open against 44 psig at the top of the drywell. NYPA determined dur-
ing re view of calculations that the valves might not open with this
pressure applied.
NYPA reviewed and revised the valve operations " , calculations and determined that a modification to replace the shaft-
'to-disc taper pins with higher strength pins would correct the prob-lem. NYPA performed this modification on both the drywell and torus ,' . vent valves resulting in. valves that would open and close with maxi- '~ mum containment pressure of 44 psig and 50 psig, respectively.
The inspector reviewed-the modification package and the associated safety evaluatio*. and found them to be acceptable.
In addition to revising the calcu,ations associated with the drywell vent valves. NYPA performed a detailed review of all calculations ! - which had been used during E0P generation.
Some minor errors were ! identified and corrected, but resulted in no changes to the E0Ps.
b.. Generic Letter 89-10: The inspector reviewed NYPA's response.to this
generic-letter, which reauested. establishment' of a program to.' ensure the operability of all safety-related MOVs under design basis condi-tions.
Specifically addressed was -the ' ability 'of-' MOVs to isolate during outside primary containment - steam line break (HPCI,.RCIC, 'RWCU).
NRC sponsored blowdown testing' of' several MOV designs have indicated that stem. thrusts required. to assure valve closure ' were-well in excess of.the' predicted stem thrusts based on valveLand oper-' . ator design formulas.
This research showed that ASME Section XI' '; testing was nottable to identify the MOV operability problems-under _ design basis: conditions, since this testing was normally doneLunder < , static conditions.
In Ltheir response to the generic letter, NYPA stated that they were legally bound toLSection XI, 'as required by:10 CFR 50.55(a), which would_ remain their; basis-for determination of MOV- . l' s-i - r I '
_. _ _ _ _ > ., ., 'O ... operability.
By letter dated August 13, 1990, the NRC disagreed with NYPA's operability position stating that if NYPA determined that an MOV was unable to perform its safety function as a result of the MOV program established by the generic letter, the MOV was inoperable and must be treated accordingly.
The inspector reviewed the results of NYPA's VOTES testing, during the 1990 refueling outage, on the HPCI, RCIC, and RWCU containment isolation valves.
These valves were replaced in 1988, as part of NYPA's containment leakage improvement program, with Anchor / Darling.
double disc gate valves. It should be noted that this specific valve design was not tested during the NRC sponsored blowdown testing. The inspector reviewed the MOV design sizing analyses, target thrust calculations, and the measured thrust for these valves. The VOTES testing indicated that several of the valves were developing thrusts that were above the design values and could indicate a valve disc friction factor of.3 which was greater than the design value of.2.
The NRC blowdown testing on other valve designs, indicated that under blowdown conditions due to additional dynamic loading a valve disc
friction factor of.5 may be more representative.
NYPA-stated that these valves would function under design basis con-ditions because double disc gate valves would have a better perform- ", ance than the valves in the NRC sponsored testing.
Further,-these valves have not been specifically tested in a blowdown condition so no data was available to show whether this type. valve will exhibit.
n any increase in valve disc friction factor due to the addition of the dynamic loading.
NYPA initiated a purchase requisition for Anchor / L - Darling to provide an analysis and material upgrade for these valves L to assure design basis function 'up to a.5 valve disc friction fac-tor..NYPA was following industry research on this valve design and planned to take appropriate actions if -any additional blowdown test results became available, which indicate an increase in ' valve disc L" - f riction factor.
Inspector-review determined that NYPA was improving the overall?MOV F program.4 The performance of thrust calculations and establishment of b torque switch settings was in accordance with vendor recommendations.
The inspector did identify some safety _ related MOVs?with incorrectly listed torque and limit switch settings in the MOVJmaintenance pro- , cedure (MP-59. 21),: : Tabl e. 10.1.
The -MOV engineer. understood. the.
i errors in the table and stated that the : correct settings would be maintained because any' switch setting changes were concurred in by_ < the MOV' engineers performing' the associated calculations to support < the baseline testing.. In addition,. any-discrepancies found in the field were.' resolved by the engineer. - The MOV engineers were in the - process of correcting these errors.
The inspector planned to con-tinue review of. NYPA's actions to address. this issue and improve
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On September 14, the inspector informed NYPA of a potential concern, initial *y identified at Millstone Unit 1, over the capability of the { RHR heat exchangers to handle the flow of both RHR pumps. NYPA con-i tinued to review this issue. The inspector plarmed to review NYPA's actions in a subsequent report.
F-1 t .- d.
(Closed) TMI Action Plan Item I.D.2, Safety Parameter Display System
(SPDS) and Generic Letter 89-06.
The inspector reviewed the Fitz- , Patrick SPDS and determined that it was adequately implemented and ! ,_ [L met the requirements of Generic Letter 89-06 and NUREG-0737, Supple- , ment 1.
The NRC staff accepted the FitzPatrick SPDS in two steps. A , safety evaluation report (SER) was issued, on March 18, 1988, which o stated that the system was acceptable except that NYPA had to either add or address why they were not going to monitor; 1) containment ' radiation, 2) containment combustible gas concentration, 3) contain- - ment isolation status and 4) source range neutron monitoring. Subse-quent to-this SER, Generic Letter 89-06 was issued, which requested i that'NYPA review their SPDS system against an included checklist and ' certify that the system monitored the required parameters.
NYPA certified, on June 29, 1989, that the FitzPatrick SPDS raet all the - requirements of NUREG 0737, Supplement 1, including those addressed . above. - except for source range neutron monitoring.
The NRC staff , determined based on this submittal that the SPDS was satisf actory.
The issue.-of.the source range monitors still remained open pending
NRR and industry initiatives and evaluations.
l " The inspector reviewed the FitzPatrick $PDS using the checklist pro-
vided 'in the generic letter as a guide and found_ that the system was adequate and monitored-proper parameters.
Based on this review the inspector closed this TMI item and generic letter.
- e.
(Closed) Unresolved Item (90-04-06): The inspector reviewed LER , t 90-19 on.'a design error regarding the ADS accumulator, supply check b valves.
The design error resulted in the installation of check ' , valves with 25 psig-cracking-pressures, such that an unaccounted. 50 psig. pressure drop, occurred across. the two valves, NYPA determined that ADS was able at all time to meet its ' requirer safety function b (i;e.,: one actuation of the; SRVs.to depressurize E reactor). This 't p statement appeared to be true provided the pneumatic supply pressure was maintained above 105 psig (normally-120 psig), NYPA:did document' ,. that the system. prior to the 1986 modification which installed the-1* backup nitrogen supply wculd; not have met the commitment's documented-l by NYPA to meet TMI: Action Plan Item II.K.3.28, (i.e. 2 SRV cycles at ' 70% of. containment design pressure of 62 'psig).
In addition, prior theJ 1990 refueling outage, any pneumatic system failure -(pressure ' L . regulator, system leaks ~ greater than regulatory capacity) which would , . . P .iU _
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< have caused a slow decrease in pneumatic supply pressure would not have been identified until the pneumatic supply pressure reached 85 psig (low pressure alarm setpoint). NYPA documented that accumulator pressure might not have been sufficient to allow a single operation of the ADS valves, if header pressure decreased below 105 psig for a sustained period of time.
The time element was moortant since the check valves would have maintained the pressure in the accumulators above 55 psig (i.e., 105 minus the 50 psid from the check valves) needed to perform one operation of the SRV for some unknown period of time.
From discussions with operators the inspector determined that' ~ ADS pneumatic supply pressure could drop to approximately 100 psig for short periods during MSIV testing.
The 25 psid cracking pressure check valves were installed from ' initial construction until discovered in 1990.
This design error caused NYPA to deviate from commitments made in response to TMI Ar tion Plan Item II.K.3.28.
The safety significance of this condi-tion was very minor because the period of time with pneumatic , pressnre below 105 psig was minimal and no system failures causing a pretsure decrease to the low pressure alarm setpoint were documented.
Further, NYPA committed in the LER to revise the FSAR to better + . reflect the design basis for ADS.
Based on these actions this instance was evaluated as an non-cited deviation in accordance with-NRC's Enforcement Policy.
Previously Unresolved Item 90-04-06 was , reclassified as a deviation and closed.
f.
(Closed) Unresolved ' Item (90-11-01): NYPA adequately. completed the ' recommendations of GE Sil'No. 452, Supplement 1 for performing a loop.
calibration from the feed flow transmitter to-the~ process computer's o " printed output. This was needed to eliminate potential bias in the process computer calculations. NYPA elected to perform 'the loop calibra* ion via a method slightly dif ferent than-the SIL recommends, but the inspector agreed that NYPA's method 'of. applying the calcu-lated. differential pressure to the transmitter - and. verifying the - results at the computer point' performed the same function. -If the f loop : tolerance. was not within the acceptable range, troubleshooting to adjust individual components would have been ' conducted.. This.
' ' -method ;would identify and allow correction of potential bias in the . resistance circuit,' which was the major. concern : off this' section. ofE . - the SIL. This 1 tem was closer'. ' g.1 ~(Closed) Violations-(90-11-02) and (90-11-04):.NYPA agreed with the . violations in :their May 9,1990 response.
The' inspector reviewed NYPA's implementation of.GE SIL No. 452, Supplement I recommendations . andfound the actions. acceptable.
NYPA verified all, similar , LRosemount: transmitters in the plant were. properly compensated for static pressurization and any ef fect had been included in the :asso-- ciated instrument' setpoint.
NYPA determined the failure to perform-M cu
._ _ _ _ _ - _ _ L 4{. ' t-G l l static pressure compensation on the feedflow transmitters was an ' isolated occurrence.
The transmitters were replaced by a minor mod-ification that did not get adequate engineering review, indicating weaknesses in the administrative controls for minor modifications.
NYPA technical services department took additional precautions to ensure all modifications have a qualified responsible engineer , closely intnived with the development of calculations and post modif-ication ts.ing.
Onsite and corporate engineering were in the pro-
cess of ev.npleting a formal change to the modification control manual '
' to ensure adequate engineering reviews.
The corrective actions appeared adequate and these items were closed.
h.
On September 4,1990, NYPA issued a letter to the NRC staff dealing !. with the three phase bolted shnrt issue while testing the emergency diesel generator (EDG) in parallel with the main transformer.
This letter requested that the NRC waive the design requirements for-the l three phase bolted short and-thus allow the EDGs to be tested in parallel with the main transformer.
This submittal satisfied NYPA's commitment to submit a supplement to LER 89-12, and allowed closure of item F-4 from Inspection Report 89-08.
! 1.
During the review of the steam leak detection steam described in Section 7.a. the inspector noted that portions of the steam leak i detection system 'would not fail conservatively (i.e., generate an isolation) following a loss of power nor would there be any indica-tion to' the operators that this portion of the system was not oper- , ational. The inspector also identified a minor deficiency in the TS.
L Further, the referencing of controlled drawings by other controlled drawings and the drawing information in the caster equipment list ( (MEL) was not consistent. The' concerns are as-follows.
The steam leak detection system.was initially designed to meet -- IEEE Standard 279.
The inspector could find no control room indication that was able to inform the operators that the power , to the-steam.-leak detection circuit had been. lost when the , ' specif_ic leads were lif ted.
This appeared to be a needed func-i tion per IEEE,279, Section 4.13, Indication of Bypasses.
In ' this specific case or in. the case ofifailure of a fuse supplying.
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power to the temperature switches, there. was no way for oper- ! ators to know that the instruments were bypassed; or that they would not function because of loss.of power. NYPA was reviewing this issue' and the.. inspector planned to follow the. resolution in a subsequent report. F-2 . & I
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Technical Specification Tabic 3.2.2 lists six (6) total, three -- (3) per cnannel, high temperature isolation instruments as pro-vided by plant design for RWCU isolation. Since the implementa-tion of modification 82-53, there have ber eight (8) instru-ments, four per channel. NYPA planned to cor-rect this error by a TS change submittal; see section 7.d.2 below.
The numerous drawings needed to evaluate the steam leak detec- -- tion system (control room panel 9-21) were not consistently referenced by each other and by the plant's MEL. The steam leak' ' detection elementary diagram indicated the terminal power sup-plies for all the associated instrumentation, but did not reference the more specific panel wiring diagrams.
The panel wiring diagrams for wiring coming into the 9-21 panel and for internal wiring did not reference each other.
The MEL did not provide a list of all controlled drawings needed to evaluate the 9-21 panel when the 9-21 panel was accessed.
Further, the MEL-information for various temperature switches was not consistent (i.e.
some temperature switches listed loop diagrams while others did not).
7.
Safety Assessment / Quality Verification a.
LER 90-021-00: The inspector identified several-additional causes, other' than those' identified by NYPA-in this LER, for. the June 28, 1990, reactor water cleanup (RWCU) division l' isolation-valve ' closure.
The. inspector conducted'-his evaluation following-NYPA's investigation and submittal of the LER. T_he additional causes included: failure to have adequate control of safety related equip-- . ment when providing electrical isolation, ' failure - to ensure that ' revised controlled NYPA drawings were updated tocthe. as built condi-tions following modifications, and-failure to properly review a draw-ing change for possible effects on other drawings.
_ ' As background,- the event occurred with the unit in cold shutdown.and the RWCU system in operation. The isolation occurred when.an instru-ment and control (I&C) technician relanded leads.that he had removed for electrical.-isolation 1 to allow replacement u a~ RWCU division 2 steam leak detection temperature switch (12TS-102B).
The technician' . used a: loop drawing (i.e.~ detector, instrument, relay / contact, ~ fuses,. . - and-power suppl )s to determine what leads - needed - to be lifted to . / electricallylisolate the temperature switch.
This drawing was; in , '
- error; it' indicated that the
- division ' 2 switch was powered from a-division'1 power supply.
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The inspector agreed that the loop drawing was in error. NYpA deter-mined that the loop diagram haJ not been properly updated by a con-
~ tracto' # . wing a 1934 modification (082-053). However, there were hunierous otner reasons why the isolation occurred, b The inspector identified deficiencies in the control of equipment
including failure to use the tagout procedure, use of an inappropri-ate drawing to determine electrical isolations, and poor understand-ing of divisional interfaces.
Specifically; Instrument and control standing order (1C50)-12 t'iowed tech- -- _.
nicians to lift leads to complete troubleshooting ar.d mainten-ance.
This appeared to be inconsistent with the unit tagout procedures.
There was a miscommunication between operations and 1&C. Oper- -- ations believed that the technician would use IC50-12 to dis-j connect leads at the temperature switch, replace the switch and reland the leads.
They did not understand that the technician was going to disconnect leads in a control room panel to provide isolation.
The lif ting of the leads for electrical isolations did not get -- operator review and tags were not hung, as required by the unit _ tagout procedure.
A voltage frisk was not performed af ter isolation to ensure that -- the isolation was ef fective before starting work.
This was a requirement of the tagout procedure. If this had been done, the , technician would have noted that the isolation used was incorrect.
. ) The use of a loop diagram to identify isolation was inappropri- -- _ ate.
The inspector reviewed several loop diagrams and identi-fied that the power supply leads indicated only tell the user where the ultimate power was coming from and did not indicate whether the instrument' was the only ' component supplied from those terminals.
The l&C technician chose to lift leads rather than to pull fuses -- that could have isolated the instrument in question.
Further, the.1&C technician apparently did not notice that he -- - lif ted leads in the division 1 portion of the 9-21 panel, while __ the switch he replaced was in the division 2 portion.
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..,4 e . , , ! Il ' ! If the I&C technician had been following the tagout procedure, the chance of such an event would have been minimized because of the nor- , mal attention paid to electrical isolations by the operating staff, i - Operators do not use loop drawings to determine isolations and would have used the elementary drawing that reflected the as built i , 1 - condition.
! , l There were drawing errors on other controlled drawings that indicated l.
the supposed as built conditions following modifications.
When a
' technical services engineer identified the drawing errors for this specific temperature switch, all applicable drawings, including the . loop, were not checked to ensure that they were correct.
[;: Specifically p ! '~ The steam leak detection system elementary drawing (791E472)' . -- prior to May 1990, did not show the as built conditions follow-ing modification 82-53 completed in 1984.
The NYPA review of ' the completed drawing was not adequate to ensure that the-as- ' built condition reflected th, installation of 12-TS-102B to the division 2 power supply and several other changes made during that modification.
A NYPA engineer identified and corrected the drawing errors on -- this elementary diagram in May 1990. While performing trouble- ,'
- shooting for modification 89-084 -(removal of the RCIC and HPCI steam leak detection high temperature isolation time delay relays), th9 responsible engineer noticed - that the elementary drawing was in error with respect = to the-power supply of-12-TS-1028.-. This discovery was outside th9 bounds of the modif-ication. However,. the engineer included the required change on
.a' modification engineering change notice (ECN) that - also cor-rected modification related errors. The engineer only corrected p ' the deficiency in drawings that were 'already identified.in the scope for the modification and did.not review the loop diagram, , i.
=The inspector concluded that the~ review processes used by NYPA to g
- ensure that drawings were properly updated. to the' as built conditions were = inadequate,, as
- was the process used-to update a drawing with an
. ECN. when the condition being changed - was. outside the - modification ~ scope.
It: appeared,that once the field as built drawing was sent to White Plains for corporate revision, there was.no verification that-an engineer reviewed the : completed ' drawing-to ensure they. reflected . the as builto conditions.
The' revision - section-of the drawing has : blocks = for the initials -of an engineer and a verifier, but they were , ~ crossed out.- The inspector reviewed the modification ' procedure - for-use of _ ECNs. 'and determined that ' there was no specific review to g.
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ensure that drawings other than those already identified for a modif-ication be reviewed to ensure that an ECN incorporating an item out-side of the modification package did not effect another drawing.
For I this reason the inspector concluded that the more appropriate method to change the drawing error, to ensure that drawings identified out-side a modification, was a drawing change request (DCR).
DCRs get ' reviewed to ensure that all drawings are updated to correct a given deficiency.
The inspector considered that the review process for ensuring the accuracy of plant drawings, which state that they were as built, and the process of updating drawings to correct conditions outside the scope of a modification represented an unresolved item.
UNR 90-06-01 Identification of these causes indicated that NYPA had not completed an adequate review to determine and correct the cause of this event.
This represented a violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action and 10 CFR 50.73, Licensee Event Report, which required that NYPA perform a review of unexpected engineered safety feature actuations to determine and document the causes and correc- , tive actions needed to prevent recurrence.
VIOLATION 90-06-02 b.
(Closed) Apparent Violations 90-04-01 and 90-04-02. On August 21, an enforcement conference was held in NRC Region I to discuss the find-ings of Inspection Report 90-04, with respect to the ESW system. The enforcement results of the conference, the list of the NRC and NYPA participants and portions of NYPA's presentation material were transmitted to NYPA by a separate letter dated October 9, 1990.
Based on this letter, apparent violations 90-04-01 and 90-04-02 were withdrawn.
c.
(0 pen) Unresolved Item 90-02-06, Review of LER-012-01. LER 90-012-01 was a supplemental LER discussing the results of NYPA's evaluation of ESW deficiencies discovered during the 1990 refueling outage. In the cover letter of the LER, NYPA stated that the submittal was voluntary because all components would have performed as designed based on engineering evaluation of the as-found conditions.
The inspector reviewed two calculations dealing with determination of the effect of the as-found condition on cable / electric bay unit coolers and on the operability of swing check valves that were found stuck open.
Calculation JAF-90-095: Comparison of flow resistance of 30*4 -- restriction in ESW' piping to throttled cooler balance valve.
This. calculation showed that for each case the 30*4 blockage observed in the piping was of sufficient length to not be neariy equal to the flow resistance of the downstream throttle valve.
The inspector agreed with the calculation but it did not demon- _ strate that flow to the coolers would have been sufficient in ' the as-found condition.
The inspector discussed this with licensee representatives who agreed with the inspectors conclus-ion and are reperforming the calculatio.____ _._______ _ __ p.
{
! ! Calculation JAF-090-093: Area required to close stuck open l -- swing check valves.
The inspector found that this evaluation I was extremely simplistic. The normal design for a horizontally , mounted swing check valve is that gravity alone should provide
the initial closing movement. Further, NYPA did not have actual
test data to conclude that 150 lbF would have caused the valves to go shut. This number was derived after the fact by asking a . maintenance man if he believed that the valves would have gone ! shut if ke stood on them.
Further, t'4 applied differential pressure was assumed to be 60 psid.
he inspector expressed i i concern that the assumptions did not all appear to be rigorous i engineering assumptions.
The licensee acknowledged the inspec- ' tors concern and agreed to review the assumptions used and i reperform the calculation if necessary.
Further, statements were made in the t.ER, that a quarterly flow test , had been performed to dernonstrate the operability of the ESW inlet i piston check valves.
The testing referred to was a flush from a i downstream drain connection.
Actual flow through the valves, as ' would have been required to meet _ ASME Section XI was not measured.- . It should be noted that NYPA took exception to performing such flow i ' test in their IST program and received approval to perform the open and inspect method for determination of operability, ' p-In summary, NYPA's justification that ESW would have performed as designed in the as-found conditions was not complete. NYPA intended
to continue this evaluation.and to. provide the results to the: ' inspector. This item remained open.
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The inspector' discussed several outstanding Technical Specification (TS) issues with the. acting Director of Nuclear Licensing.
The fol- " . lowing ~ issues, addressed in-previous reports, were being' tracked by ' , , NYPA for proposed TS change.
The inspector consolidated these pre-vious issues and planned to review the remaining' issues in a subse- . < ' 'quent inspection report.
F-3 1) NYPA - submitted - a proposed amendment to. NRR on April 2.1990, to - ! provide more information in' TS 3.5.F.2 for ECCS systems needed ' during' cold shutdown 'does not specify what ~ systems are required , when maintenance that has the potential to drain the vessel was . y being performed. 'This action closed item F-1 from Inspection i Report 89-09.
', 2): The instruments:that' provide primary containment isolation.(PCI): ' functions for HPCI and RCIC were included with the instrumenta-- , tion-that controls and ' initiated the ECCS ' system (TS table - 3.2-2) not withLinstruments that cause PCIs.(TS table 3.2-1).
'! Since PCI : instruments may' be required -to be operable when ECCS -
Linstruments are not, the distinction is important..NYPA was in
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- the process of developing a proposed amendment to address this I ' y,, issue.
The inspector-elso confirmed that NYPA planned to cor-
E rect the number of high temperature steam leak detection instru- ~ ments installed for RWCV in this submittal (this omission was _ m' - , ' noted in Section-7.a above). This closed item F-1 from Inspec- " ti.on Report 89-09 l ' e 3) (Closed) Unresolved Item (89-11-01): TS 3.5.0 states that the ADS SRVs shall be - operable whenever reactor pressure is greater- ' +% than"100 psig. To meet SRV vendor recommendations against low' ~
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. pressure SRV t_ests, NYPA performs ADS SRV retests at approxi-
mately 940 psig with the mode switch in run.
NYPA was. in the j m process of' developing a proposed amendment to address this '
, issue.
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4 4) NYPA:was-in-the process of developing revised ESW flow rate.TS >; 'to -reflect existing commitments with respect to ESW system- - ~ operability.
g s E '8.- -Mid-Cycle Meeting and Other Inspections l a w .. . a . I a.
. On-September 14, Ta mid-cycle meeting was held in NRC Region lI to ' - ' discuss NYPA's self initiation and the NRC's assessment: of NYPA's , 's- = performance to that date in SALP cycle nine. A list of the NRC'and L .NYPAxparticipants and: portions. of NYPA's handout are ' Attachment- (1)- -! m - to;this. report.
, =q b.
_ Inspection Report 90-20, E0P Review, August 20 -thru August 24,.1990.
_ ' L9. LEgt11nterviewy f* L At periodicLintervals -during thel eourse of-this inspection,7 meetings were; ..- 1 held:withs seniorifacility. management:to discusscinspection scope and; find-i ' ingsc. 'In' addition, - at thei end ;of thel period',J the inspectors met _with- _. - j, ' - licensee _ representatives sand summarized 1the! scope = and findings ;of' the' i . ( inspection asitheyLare described in'this report.
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- American Society of Mechanical Engineers
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Emergency Diesel Generator - J
- E0P-Emergency Operating Procedures
- - ESW - Emergency Service Water - , -Y HPCI High Pressure Coolant Injection System - ? IST In-Service. Testing . LER-Licensee Event Report - ' i-- .MEL Master Equipment-List - MG Motor Generator + =MOV-Motor Operated Valve -
1 Main: Steam Isolation Valve MSIV.
' - LNRC Nuclear Regulatory Commission '- ., -NRR Nuclear-Reactor Regulation.
-- New, York ~ Power Authority
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.'J = 0P:t ' Operating Procedure
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PASS; ' Post Accident: Sampling. System -
- PCPL-Primary-Containment' Pressure Limit
- -- RC IC L -- -Reactor Core Isoiation" Cooling Syr'em- > RHR' - - Residual Heat Removal-System RPS1 Reactor. Protection System - RWCU- - Reactor ' Water Cleanup Syst.em - -
iSER1 Safety Evaluation.; Report ' - SIL': Service Information Letter- -
- SRV; Safety Relief Valve:
- 'Sur'veillance Test-
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Service Water-. ^ CV .TMIf ~ Turbine Control Valve - = ' ' Three. Mile Island- -- 4TS< 4-- Technical Specification.
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' ATTACHMENT 1.T0 INSPECTION REPORT NO. 90-06
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. Attendees of the September 14, 1990 Mid-Cycle Meeting in Region-I.
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- W. Kane, Deputy Regional Administrator . a ' C;.Hehl, Director.-DRP '
- J. Wiggins, Deputy-Director, DRP
' JJ,lLinville, Branch Chief, DRP . ; ( "g G. Meyer, Section Chief, DRP . ! > T DE Vito,l Project Engineer, DRP-
.W.LSchmidt,' Senior Resident Inspector i , M. Miller;--Reactor _ Engineer, DRP.
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-R..Capra,oDirector Projects Directorate, I-1, NRR . a i iD.LlaBarge, Project Manager, NRR
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JJ?Brons:,cExecutiveLV.P.,NuclearGeneration .. x' j (RGBeedle, V.P.,- Nuclear Support ' t-
TW;:'Josiger,,V.P., Nuclear 0perations ' F, zWl-;Fernandez,iResident Manager '?,
R :Liseno.: Superintendent of Power. A p =GR:Vargo,LRES? Superintendent U .- ' *, , ?Gj.Tasick,DQA Superintendent'
' " iJ;.Elmers,LActing"BWRLicensingManagerJ . ," -E, ' - , - T I} .,. I,.
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.-__ _ _ __ - _. - _ -. _ -. - - .. ....-..- - _- -. - - _... ' ,7, , .. ATTACHMENT 1 TO IR 50-333/90-06 . , MID-SALP MEETING 9/13/90 . . 1.
SELF ASSESSMENTS / MAJOR CONCERNS , 2.- SURVEILLANCE PROGRAM ADEQUACY ! 3.
PERSONNEL ERRORS ' 4.
HANDLING OF LARGE SCALE-PROGRAMS REQUIRED BY , L TECHNICAL SPECIFICATIONS 'i t i
- 5.
NRC / FITZPATRICK COMMUNICATION ISSUES 6.
GENERAL DISCUSSION ISSUES L !
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_- - - - - __ ,. *,.. , 3-ATTACHMENT 70 IR 50-333/90-06 . SELF-ASSES 5AENTS/ MAJOR CONCERNS Discussion of Issue Some comments have arisen concerning FitzPatrick's questioning attitude or ability for self-appraisal.
A pparentl HPCI anc' ESW have initiated some of this tainking.y, issues with We do not see - where this should be considered a potential problem.
- Historically, if anything, we have been accused of being too self-critical.
, - Has there at any time been an issue with PORC review, operation, self-questioning? No, in fact, through the years, performance has been well-thought of.
. - QA audits (not required by T.S.) over the past several years include: - Annual' trending report of all AQCRs < - Planned maintenance task force activities I - Drawing update program - IST, ISI, snubber programs - Reactor Analyst procedures and program PASS and effluent monitor systems - - Welding program - Effectiveness of PORC - Control room activitier including start-ups '! - EQ program . - Note, problems with HPCI, ESW were self-identified by existing programs.
- Detailed review of EQ air conditioning unit 'ablems.
' - We identified and addressed Rosemount transmitter problem-in '86.
., ! -1 Numerous changes /new programs are continually being initiated to improve overall operation (see Management ~ Goals and Performance Indicators).
Most Significant Areas of Concern and Actions in-Progress 1. - Communications & Accountability - Management-union committees - Clarification of " big-picture" responsibilities
- ContinuedLmeetings of senior management with all plant _ personnel' . I ' f f t r
__
., . ATTACHMENT TO IR 50-333/90-06 . - Expanded. Planning Department and changes to interdepartmental meetings-1 Expanded newsletter and interplant message monitor system - Clarification of procedural use , - Long-term upgrade of administrative program 2,- Radiological Program (See Corrective Action Program) , e 3.
Trending Programs - Defined, integrated root cause, and problem trending programs c, 4.
Organization / Engineering Support li !! - Comprehensive staffing and organizational review in i progress t 5.- Facilities ' - Long-term facilities program continuing . , i i h s i
. m
i b '" , . . ... _. ., * , ATTACHMENT TO IR 50-333/90-06 , PERSONNEL ERRORS Discussion of Issue Over the past two years more emphasis has been placed by management to identify personnel errors and to critique the events to determine the cause.
This certainly has raised the consciousness of both the staff and the NRC concerning this matter.
Questionc naturally arise as to whether the problem is increasing or not and the reasons behind ' How the Issue Is Being Addressed 1.
Continuing the philosophy of not overlooking errors, but . raising the level of review and consciousness and ccatinued H performance of critiques.
2.
Fourth quarter 1989, a training module-for "self-checking" was performed in.the continuing training program for operators, maintenance mechanics and electricians, I&C, technicians, and RES technician personnel (see attachment).
This module of self-checking has been added to the a'aprentice programs and has been -incorporated in management ~ oicarvation training programs.
3.
-Continuing 'rk on improving human factors of procedures.
4.
The overall philosophy on procedure use-and what is expected from personnel is being reevaluated (industry problem).. A draft document is presently'in review which clarifies many issues and provides additional guidance on use, changes, when needed, etc.
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- - - - - .. .- 0- , ATTACHMENT TO IR 50-333/90-06 .
RANDLING OF LARGE SCALE PROGRAMS REQUIRED BY TECHNICAL SPECIFICATIONS (Ex: ISI, IST, Snubbers, etc.)
Dis'cussions of Issue Several years ago, as a result of problems identified in the IST program, we commenced systematic reviews of various programs to identity and correct other problems.
These reviews aave been, " and continue to be, done in-house and also through the use of QA who use industry experts on occasion.
Results have.been self-identified problems in the snubber program, the fire program, etc.
The major cause of this oroblem appears to be the lack of formalized programs which can transcend personnel changes.
Many of the programs were correctly established and implemented by one or two individuals without complete formalized administrative programs or implementation.
When they left or were replaced, deterioration-frequently occurred as new requirements or changes - occurred.
How the Issue Is Being Addressed The-process'of systematic reviews and formalization of ' administrative programs continues.
Recent efforts include: IST-Procedures - PEP-6.0 Series (4 procedures) - 1989 & 1990 PSO-31, Revision 9 - 6/90
QA Audits - 6/89 &'5/90L ISI Procedures - PS0-31A - 4/90 QA Audits-9/90 - Fire Program Procedures - Design consolation in progress with administrative program details to follow.
QA Audit - Scheduled 4th Quarter 1990
. Snubbers Procedures - MDS0-13, Revision 0 - 7/88 Revision l - In Progress QA Audits - 6/90 ,
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'e ATTACHMENT TO IR 50-333/90-06 - SRV Program Incorporated into existing maintenance PM program this year.
E_Q Program Procedures - MDS0-18 - 8/89 WACP-10,1,11, Revision 11 - 7/90 . QA Audit 8/90, 11/89 - oth*er Performa2ce Programs Procedures - Vibration Analysis, PEP-3.1 - 10/89 Performance Monitoring / Analysis of Heat Exchangers, PEP-2.3,2 - 6/90 . t'- !. .y.
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ATTACHMENT T0-IR 50-333/90-06 . NRC/FITZPATRICK COMMUNICATION ISSUES Issue: Authority disagrees with violation or wording in report, or we question - in many cases never have a . response.
Examples 1) NRCI-89-03-(Item -01) (Response 6/21/89) Issue on violation of 10CFR50.72 notification on TCV found operating incorrcetly repaired assessed several days latcr after evaluation that event was reportable under 10CFR50.73.
We stated in response that NRC should provide clarifications and we discussed this at the last SALP, region said it would check witn headquarters on issue - no feedback (see attached).
2) NRCI-89-09'(Icem -02) (Response 11/89) f On cover page cited an example of unprofessionalism of SR0s concerning declaration of SBGT LCO.
We responded our disagreement - no feedback _(see attached).
. 3) NRCI-89-80 (Item -01) (Response 9/89) Issue on violation concerning RBCLC lines'in containment and valve isolation.
We disagreed with violation - no feedback.
4) NRCI-89-80 (Item -02) (Response 9/89)- Issue on violation due to four examples _of inadequate. ! procedures.
We disagreed.
NRC closed this open item in M.
Inspection 90-13 but never addressed agreement or disagreement of-violation.
- - . 5)_ NRCI-90-01 (Item -01)-(Response 5/90) Issue involved violation of SR0 in lunchroom being not in ' l command.
Our response agreed:that' corrective action
inappropriate, we stronglyfdisagreed with= statement SRO not o in come.and.
NRC closed open' item in Inspection 90-03,.
L however, no mention of disagreement (see-attached).
l6) NRCI-90-03 (Item -04)' (Response.8/90) ! l' L Issue addresses cover letter and report statements on
10CFR50.59 control room human factor. issue.
We disagreed , y with many of the statements.
NRC closed open item in Inspection _90-04 - no feedback on issue (see attached).- L I y . . > .- . - - -.
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ATTACHMENT TO IR 50-333/90-06 . 7)- NRCI-90-04 (No formal response as of yet) Issue involves written statement in report of HPCI testing during start-up (see attached).
Does not describe numerous tests done earlier in the day and the previous day, or the fact that numerous people were involved with the trip and why post testing was performed.
P esents wrong problem.
8) - NRCI-90-012 (Response 6/90) Issue involves fine on overexposure.
We agreed with violation but requested a review of the amount of the fine due to inaccurate statements made in the rep' ort.
The NRC responded 8/90 but never addressed many of the inaccurate statements (see attachment).
9) NRCI-90-013 (Item -01) (Response 5/90) Issue involves violation concerning fuel oil sampling limits and proposed Technical Specification change.
We disa with Level IV violation - no feedback (see attached). greed
-10) NRCI-90-017 (Item -02) (Response 7/90) Issue involves PORC review of RES procedures.
We disagreed with violation - no feedback (see attachment).
. 11) Others-11ssue with C. Morse and infamous ele ~ctrical questions.
- What should.we do with inaccuracies in inspection reports? - Many inspection reports put findings in perspective, i. e., reviewed ten mod packages and found the following problems ...: many do not.
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_ _ ,_ _ _. _ _ _ _ _ _. _ i oO m ' e . ATTACHMENT TO IR 50-333/90-06 - , . GENERAL DISCUSSION ISSUES J 1) Emergency Planning Icsue - NUREG 1210 NRC Training Manual has a philosophy.of evacuation upon declarin j a 2-mile radius quickly,g a General Emergency within whereas the present NRC flowchart has sheltering with further calculations needed prior to evacuation.
, Issue - Philosophy of downgrading of emergency classifications appears to be contradictory between i regions.
Vogtle declared a Site Area Emergency and upon restoring EDG, they downgraded to Alert.
Our information says Atlanta Region found acceptable whereas this region does not? Issue - During our December drill, the region response team is to participate.
Will a pre-drill training session be held? How do NRC personnel gain access / training in the middle of a casualty? 2) Training > Information provided to date is that upcoming new license.
exam will have a simulator evaluation performed by 1 SRO and 2 SR0s.
This is contrary to NRC guidance, our training programs method of shift alignment, etc.
- 3) Security Security inspection reportu are not being sent to the
Resident Manager.
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