IR 05000213/1992021
| ML20127H653 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 01/15/1993 |
| From: | Doerflein L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20127H628 | List: |
| References | |
| 50-213-92-21, NUDOCS 9301250014 | |
| Download: ML20127H653 (22) | |
Text
.,
. - -. -.. -
.... -. -. _. -
,. _. _. - - -.
.- -. -.,
r a
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-213/92-21
-;
Liecase No.
DPR-61 Licensee:
Connecticut Yankee Atcmic Power Company P. O. Box 270
,
Hartford, CT 06141-0270 Facility:
Haddam Neck Plant Location:
Haddam Neck, Connecticut j
inspection Dates:
November 15, 1992 to December 12, 1992
,
inspectors:
William J. Raymond, Senior Resident inspector Peter J, Habighorst, Resident inspector Robert DeLaEspriella, Reactor Engineer
.
Approved by:
hu / * (&1 I i5
'
.
Lawrence T Doerflein, Chie Dat'e
'
Reactor Projects Section No. 4A Areas - Inspected:
NRC resident inspection of plant operations, radiological controls, maintenance, licensee event reports, and periodic' reports,- As an initiative, tlic inspectors evaluated manual operator actions during a postulated station blackout,
. Results: See Executive Summary
.
9301250014 930115 PDR ADOCK 05000213
- G PDR-
- -
.
-
..
..
- - -.
-
-
,
,
..
..
-. - ~ -.- - -.-. - - -.
. ~. -... ~. - -. - - -....,
a-EXECUTIVE SUMM ARY
,
IIADDAM NECK PLANT INSPECTION 92-21 Elant Opna@as The inspectors reviewed risk-significant operator actions during a station blackout (SBO) event,-
The inspectors determined sufficient manpower exists to accomplish risk significant actions during an SBO_ scenario. Based on either in-plant walkdowns or the plant specific simulator runs, the inspectors noted' time critical steps in ti.e emergency operating procedure could be accomplished within the required time, The inspectors noted some human engineering deficiencies, weakness in the emergency plan implementing procedure for alternative site communications, and an inadequate basis for a special test acceptance criteria. However, none
,
of the deficiencies were significant in the mitigation of a postulated SBO. Finally, the inspectors determined the auxiliary operator training program placed appropriate emphasis on the emergency -
operating procedures and was a notable strength.
'
One non-cited violation was identified regarding the failure to properly perform surveillance on the core deluge valves in the low pressure safety injection system (section 2.3).
Maintenaltet The in'xctors noted that the inric acid system reliability was lessened due to *e need for recurrent corrective maintenarce of the boric acid filter. Further plant action is needed to identify the cause for the filter degradation..Notwithstanding the repetitive maintenance activities, the inspectors found operability of the system was appropriately maintained.
Safety Assessment and Ouality Verification
<
The inspectors determined that licensee actions to review the 10 CFR 21 report on Foxboro modules were appropriate to: determine applicability.
The inspectors also determined an-appropriate operability decision was performed.
It appears that CYAPCo has not communicated well with the plant staff about the PEP, including.
the status of its implementation, how the program is expected to benefit them, and to re-emphasize the PEP is an " enhancement" to already good performance, i
,
.,
-
,
,,
_
~,
- _.-
--2
,_
.
.n-
.
~ _.
. _
. _. _.. _. - _ _ _..,. _. -
_ _ _ _.... _ _ _ _ _
. _
._ -
. __
,
TABLE OF CONTENTS-
1.0 '
SUM M AR Y OF FACILITY ACTIVITIES....................,,
,
l=
,
.
2.0 PLANT OPERATIONS (71707,71710, and 42001).......
- 1 --
..........
2.1 Operational Safety Verification
.............,,...........
2.2 Radiological Controls............
2.
..................
2.3 Engineered Safety Features System Walkdown.................
3.0 RISK-SIGNIFICANT OPERATOR ACTIONS DURING EMERGFNCIES....
3,1 Inspection Purpose and Scope.......
...................
3.2 Station Blackout (S BO)...............................
a 3.3 Review of Supporting Activities..........................
-
3.4 Summary
..
.......
........
...............
4.0 M AINTliNANCE..........
................... -13
.
.
....
5.0 SAFETY ASSESSMENT AND QU ALITY VERIFICATION (71707,90712, and 90713)
.............................................
5.1 Plant Operations Review Committee
......................
5.2 Review of Written Reports.,.
.,.,...
................
5.3 10 CFR 21 Report: Seismic Qualification of Foxboro N-2AO-V2H Module..
.....
....
...........................
6.0 EXIT MEETINGS-17
,,...................................
Note: The NRC inspection manual procedure or temporary instruction that was used as inspection guidance is listed for each applicable report section.
,
!
l
!
i
!
L I..
'
ii
_.
l.. -
-
r-
--
-
g
-
y
,,,,,,. -,.
.
.
.
.
.
.
..
.
- -, - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
.
.
DETAll,S 1.0 SUSD1AltY OF FACILITY ACTIVITIES The unit operated at full power throughout the inspection period. -No significant maintenance or operational problems occurred. Connecticut Yankee Atomic Power Company (CYAPCo)
plamiing for the upcoming refueling maintenance outage (May 1993) continued during the period.
On November 17, 1992, the NRC Region I Division of Reactor Projects Branch Chief for Haddam Neck toured the facility and interviewed licensee employees and managers. Plant personnel displayed a high degree of pride and " ownership" in their work. Overall, facility
.-
conditions were found to be satisfactory.
2.0 Pl, ANT OPI'.itATIO'cally at the emergency diesels as necessary: stripping unnecessary AC loads from the
.'
emergency huses and disable the !!CCS pumps to preclude uncontrolled restarting when AC power is restored; stripping unnecessary DC loads to preserve the station batteries; isolating the DWST from the condenser hotwell and establishing the make-up now path to replenish the DWST inventory; isolating RCP seal cooling in anticipation of their failure to preclude further seal degradation upon restoration of AC power; and, verifying containment isolation has occurred in anticipation of the loss ItCS integrity.
Once stable itCS conditions are verified in natural circulation cooling, the operators will use the AFW system and steam generator atmosphere dump valves to begin a controlled '
cooldown of the itCS. The strategy used in liCA 0.0 will not bring the plant to cold shutdown immediately, but rather to an RCS pressure of 260 psig, which corresponds to a temperature of about 400 degrees Fahrenheit (F), where RCP seal integrity is expected to be i
preserved until AC power is restored.
3.2.2 Rhk Significant Operator Actions
Anachment I identifies the manual actions that have a medium or high value in the PRA l
ranking of risk. The inspector considered the eight speciDe actions in IICA 0.0 determined to be the most risk signincant. These actior.s are veri 0 cation of auxiliary feedwater Dow;-
'
'
restoration of electrical power; reactor coolant seal cooling restorstion; local start of the emergency diesel generators; isolation of the condenser hotwell fiom the demineralized water storage tank (DWST); shedding of non-essential direct current (DC) loads; preservation of -
containment integrity; and monitoring of auxiliary feedwater (AFW) How. Of these actions, the follenving four actions were the most time critical to assure continued reactor safety and were reviewed in detail.
'
MiDn 111515 Ti.me.to occur
,
Wactor Coolant Fump Ilearing Temperatures 30 minutes from Initiation isoiation
- F.xpected to rise above.
of Sl30 170 F i
!
i
- ,
yy y"
V---PW*'y.'
Ts
'TT--ty--
g n-S we v -
'ye9.-a-ys-g,.g
.-gyug-
.,,. + - -.g 4p gp fy,
,9
,g,f,yv 9, y.q,.s
- ewww.,
,,,v
,ge--.9.,y.,.
,3-r-gr u m-T-tey=.--g
--t'
,e-g i
'
y i- :he w er*--TsPeW-?-v"-:'7
....
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _
,
.
DC load Shed itemoval of 20 amperes to With no load prolong life expectancy of shed batteries station batteries expected to last fiu hours, with load shed of 20A 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> No Al W 1 low /
Core Uncovery 1.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> Cooldow n (Assume itCP seals leakage at 50 gpm)
Decay lleat inventory Decay and Sensible Heat ite011 within 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> liased on either the simulator runs or the in-plant walkdowns (discussed below), the inspectors verified that each of the above time critical tasks within liCA 0.0 were accomplished within the required time.
3.2.3 in-Phint Walkdowns and Simulator Itun itesults The inspectors walked down the following !!CA 0.0 actions, which include the high risk actions, in the company of a CYAPCo training representative and a qualified auxiliary operator:
- local start of the DC airside seal oil pump
- remote start-up of the auxiliary feedwater pump
- isolation of reactor coolant pump seals
- hwal start of the emergency diesel generators
- isolation of the demineralized water storage tank from the condenser hotwell
- remote control of auxiliary feedwater flow
- isolation of a ruptured steam generator
- direct current load shed
- make-up to the demineralized water storage tank (i.e., gravity drain, or dicscl Ore pump)
- manual steam generator blowdown, and cooldown
- alignment of containment spray from the fire main system.
,he inspectors verified that all these actions could be completed within the time needed to maintain eritical safety functions and to preserve stable itCS conditions in natural circulation.
For a station blackout scenario with no other faulted conditions, the total time required to complete the manual actions was 37 minutes on the simulator, as compared to 49 minutes in the plant. The 12 minute difference in time is not considered significant due to the mamer in which the plant times were measured; i.e., all actions started from the AO room. The inspectors also noted that under actual casualty conditions, time improvements in completing the actions in the plant will be realized as some tasks would be performed concurrently and the operato.s would nm, not walk, through the plant.
_-___
_ _ _ _ - - _ _ _ _ - - _ _ - _ _ _ _ _ - _ -
_
_ _ _ _ _ _. _ -
-
_ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _
'
,
0 -
Aber investigation of the critical tasks, inspector reviewed the shift compliment.
Administrative control procedure (ACP) 1.0 23, * Operations Department Shift Staffing itequirements," requires a reactor operator (110) or a senior reactor operator (SitO) at the controls at all times and a second SitO be present within the control mom. The minimum
,
shift compliment sircified by Technical Specification 6.2.2, requires two SitOs, two reactor operators, two Aos, and one shift technical advisor. Inspector review of the current shift
schedules noted that all the necessary SitOs and 1(Os, and between three and four AOs were on the rouers. At least two AOs on shift are fully qualified but the others could be in a training status to achieve AO qualification Taking into account the minimum shift i
compliment, the inspector concluded that for a Silo without any other malfunctions or
'
external manning requirements (i.e., fires) sufficient manpower ests to accomplish critical operational tasks. The review concluded that the Imbine building AO has a majority of critical tasks to perform as compared to the auxiliary building AO.
,
The inspectors noted that an Silo affech the ability to accomplish some liOP tasks. For
,
example, the site evacuation alarm and paging system were deenergized and the 'll' AFW-Imbine and pump must be monitored hically due to the lack of control room indications.
,'
Also, the feedwater bypass valves fail wide open on a low of air and must be manually throitled and a majority of the containment isolation valve position indications de-energire.
The inspector determined that the instructions in liCA' 0.0 adequately compensate for the plant system responses. Actions to verify containment isolation are addressed further below.
The inspector questioned the licensee if an alternate methmi exists for in-plant communications when the station annunciator alarm and site paging are inoperable, At the close of the inspection, the licensee was implementing a revision to liPIP 1.515, " Station livacuation," for Silo conditions.
'ihe inspectors identified minor human engineering deficiencies during the in plant walkdown.
'
The following deficiencies were identified when operating the AFW system remotely: (1) a copy of abnormal operating procedure (AOP) 3.2-51, " local Manual Operation of the Auxiliary Feedwater System," was available in the auxiliary building, but a copy would also be useful in the turbine building for the actions performed near the demineralized water area; i
(2) although emergency lighting was gmxt on either end of the terry turbine room where most actions are performed, emergency lights were missing in the mid-section of the room; and, (3) the *ll' auxiliary feedwater discharge pressure gauge could be re-oriented to facilitate
,
system operation, For the hical start of the emergency diesel generator, step 2.1.a. and
'
2.1.b. have the operator perform the action at the diesel auxiliary board in the main control board, yet the actions are done within the emergency diese; generator room. The inspector plesented these deficiencies to the licensee and actions to address them were in progress at
.
the close of the inspection period, 3.2.4 Containment isolation Verification Step 19 of EOP liCA 0.0 requires the operators to verify that the containment isolation
,
,
occurred successfully by checking for the proper alignment of valves listed in a table in the l
-
,_
_ _.-,
. _,_., _ _,_ _, _ _ _ _. _, -,,_... ~. -.,__., _ _.,, _..
-, _. _ _ _ _.. m,, _ _.
-
_- -
_._-
_. _ _ _ _ _ _ _ _ _ _ _. _
._ _.- _._ _ _
.
.
procedure. If a containment isolation valve (CIV) did not isolate properly or if valve position cannot be determined from the control room, the contingency action for Step 19 is for the auxiliary operator to vesiG successful containment isolation locally. Position indication for
twenty eight of the 47 CIVs listed in the table is provided on the main control panels. The remaining 19 valves have local indication only, and 7 of these CIVs are located in an auxiliary building pipe trench. CYAl'Co has identified this design as a human engineering deficiency, and, is considering a modification to put position indication (PI) for the CIVs on a hical panel in the Pall. This topic is being evaluated under the ISAP.
$
During the Silo scenerio, the AO was dispatched at 22 minutes into the event to verify the position of the 19 CIVs kically. lie was further instructed to not address the valves located e
in pipe trenches. The priority given to the task was to complete the verification to the extent possible and as time is available. The inspector estimated during the field walkdown that the l
AO would complete this step in 15 minutes, for those valves not located in pipe trenches.
The inspector noted that the control room operators completed the verifications for three of i
the 28 valves with control board indications within 5 minutes; however, further verification of the remaining 25 valves was deferred pending the restoration of AC power since the remote indication for the valves was unavailable as a result of the Silo event, The inspector
reviewed Step 19 in detail to verify the acceptability of the licensee's approach for completing this action.
Most CIVs listed in the table are on small diameter vent, drain, and sample lines in several plant systems. The remaining valves are either in closed systems or are in other systems
'
(e.g. feedwater, main steam) which are not expected to contain iadioactivity during a Silo -
i event. The inspector verified through a review of the associated flow drawings that each CIV ~
will fail to the closed position for a containment isolation function. The design was also verified to the extent practical by observing the valves during a plant tour. Thus, the containment isolation function on loss of air or AC power is assured as provided by the design of the control circuits for the affected valve. The strategy in ECA 0.0 is to verify that
,
the expected automatic action has indeed occurred.
,
Step 19 verifies containment isolation at this stage in the response to a station blackout in anticipation of a loss of reactor coolant pump seal integrity, ECA 0.0 later has the operator re-verify the need for containment isolation based on actual conditions (elevated pressure or radiation levels). The inspector reviewed the issue of RCP seal integrity without normal cooling and noted the estimates for the time to seal failure varied widely, ranging from 30 minutes in worst case conservative PRA analyses, to several hours in best estimate analyses and mock-up tests (reference: NU letter 1113681 dated December 5,1990). Actual experience at lladdam Neck has shown the RCP seals did not degrade following a full 30 -
l minutes without cooling. The strategy of ECA 0.0 is to reduce RCS temperature to a value that will preserve seal integrity for the long term to help cope with a SBO. Thus, the actions to assure containment integrity at Step 19 would be well in advance of the actual loss of-integrity and the consequent release of RCS activity to the containment atmosphere.
.
.
.
..
a.. u
._ ____...____ _ _,_
.-.-
_ - - -.
_ - - - - - - -. - - -. - - - -. -
_
!
'
i
1
1
In sununary, the inspector concluded the strategy used by the licensee in ECO 0.0 was proper and would result in assurance of containment isolation in a time frame commensurate with the j
-
need for the safety function. The decision to avoid sending the auxiliary operators into
-
possibly hazardous areas under conditions of inadequate lighting (i.e. pipe trenches), and to defer the completion of those verification actions until after AC power is restored was i
appropriate. The inspector noted that the licensee's implementation of Step 19 was in concert
!
'
with ACP 1.2-6.15. " Emergency Response Procedure Users Guide " which states that... "a required task need not be fully completed before proceeding to the next instruction; it is
,
suf0cient to begin a task and to have assurance that it is progressing satisfactorily " Finally, i
-
the licensee has identi6ed weaknesses in the ability to verify the position of some CIVs and
'
has plans to address this issue. Ilased on the above, the inspector had no further comment regarding the implementation of Step 19 in ECA 0.0.
>
3.3 Reslew of Supporting Activities
,
3.3.1 Emergency Plan Actions
in view of the minimum shift compliment, the inspector evaluated the licensee's procedure and process for relotting a SBO and the response of the emergency organization to support operations personnel. The emergency classincations for a Silo are a function of how long the condition exists. As required within emergency planning implementing procedure (EPIP)
.
1.5-1 "limergency Assessment," an AlliRT classification would be required if the event t
lasted greater than five minutes, a SITil AREA classi6 cation if greater than Ofteen minutes, and a GEN 11RAL EMERGENCY if the event lasted greater than two hours. The shift supervisor coordinated well with the on shift DSEO to provide a timely classincation of the
Silo emergency conditions in accordance with the emergency plan implementing procedures.
in the November 10 simulator exercise, the SS declared a Site Area Emergency 14 minutes i
after the occurrence of the SBO, as required by the EALs.
-
The emergency organization is required to respond and be activated within one hour of an ALERT classification. The on-site director of the site emergency organization (DSEO)
recently added to the emergency program provides assistance to the shift supervisor in
implementing the emergency plan and classifications. EPIP 1.5 27, " Manager of On Site
'
Resources " details the responsibility to ensure on-call people have responded, and will attempt to call in additional personnel. Attachment 12.1 of EPIP 1,5-27 recommends calling in a senior control operator and a control operator, The licensee expects this call in to take.
approximately one hour to accomplish after an ALERT classification, CYAPCo also stated that other available paths exists for the shift senervisor (SS) to request additional personnel, The SS could either call the operators in directly (reference: EPIP 1.5-26, " Manager of Controls Room Operations," step 6.1.k) or the SS could request through the DSEO to the
- Manager of On-Site Resources to have an individual respond,- The inspector concluded that sufficient means exist to supplement onsite personnel, but the additional resources would be the most benefit during the plant recovery from the event.
l l
i
r ~.,,.-
-__.m.,
..-.-.,,-_.-,.__.,_..--.-,A...,__1..~
~,-7,-s,--
,.__m.,..%
_ _ _ _
_-..... _
--
_ _ _ _ _. _
_._
b i
!
-
I
>
.
I
'
'
3.3.2 Validation of Emergency Operating Procedure liases The inspector validated steps within ECA 0.0 using either the auxiliary feedwater design basis document, or the step deviation document. One item noted by the inspector wu that a special test (ST) 11.7-17, "DWST Emergency Fill From PWS1 and RPWST." was l
performed in October 1989, to determine if the portable gasoline pump was capable of
.
'
transferring adequate quantities of PWST/RPWST inventory to the DWST during a Silo.
'
The flowrate acceptance criteria was a minimum of 50 gallons per minute (gpm). The actual
!
,
measured flowrate was between 200 to 215 gpm. The inspector noted that ST 11.717 did not include supporting engineermg calculations for the required DWST inventory and maximum AFW Gowrates during a SHO. The engineering calculations (completed in March 1989) were 99-517-896 GE, " Station Blackout Calculation for NUMARC 87 00," and REB-
89-7. "CY Maximum AFW 00ws During a Station Blackout." The inspector independently calculated average AFW Hows from NUSCo calculation 99 517-896-RE and determined that a make-up rate of NX) 215 gpm at the point specified in ECA 0.0 for Olling the DWST was-
acceptable. The inspector also independently validated other steps including the acceptability
of the diesel fire pump as an alternative DWST make up source; the basis for maintaining
!
reactor coolant system pressure at 260 psig (i.e., for RCP seal integrity considerations); the e
espected diversion of How from the DWST to the condenser hotwell; the basis for isolation of cooling to the RCP seals; and the battery load profile and the basis for shedding DC loads.
a The inspector determined adequate operational and engineering basis existed for the steps evaluated.
3.3.3 Auxiliary Operator Training The inspectors evaluated both the initial and continued emergency operating procedure
)
training program for AO's. The initial training program is described in Nuclear Training l
Manual (NTM) 3.005, " Initial Training Program implementing Procedure." After the initial
introductory training, the candidate completes both turbine building and auxiliary building
,
qualification cards. Specific on-the-job (OJT) cards (OP-NLIT-EOP T30000 and-T30001) for liOPs are performed during the turbine and auxiliary AO quali0 cation. Once completed, the AO candidate is provided comprehensive classroom training on a majority of EOPs including reactor trip (E-0), Loss of Reactor / Secondary Coolant (E-1), faulted steam generator (E-2),
steam generator tube rupture (E-3), response to loss of secondary heat sink (FR H,1),
response to anticipated transient without scram (FR S.1), station blackout (ECA 0 -0), loss of control air (EOP 3.1-34), Loss of semi vital power (EOP 3.1-46), I oss of DC bus (EOP 3.1-49) and EOP 3.1-50, (Loss of MCC-5). Once the candidate has completed this training session, he/or she is considered fully qualified by the licensee.
The continued training for AOs is generally consistent with the control room operator requalification training. The program includes a classroom phase, simulator briefine, training,-
simulator observations, and on-shift training. The classroom training consists of system refresher training, recent modifications, plant specific events, and familiarity with the EOPs.
The simulator briefings are based on EOP actions in the simulator, and cover the basis of the
, _ _ _
._,_ _ _.. _ _ _. _ _
.
..
.
.
.
_. -
. - -
-
.-
_--...---.
--.----.._-
-
.
>
.
'
,
!
actions, and consequence of failures. The simulator observations are to familiarire the AO of i
control room activitics, and to improve inter-crew communications. The on-shift training are various OJT cards to be completed once every two years. IIxamples of the OJT cards are
local operation of the AFW pumps and a response to a S130.
,
'
The inspector reviewed both the initihl trair.ing qualification cards, and on going training
,
program objectives with the licensee's training staff, liased on the proficiency of the AO during inspector walkdowns of Si30 actions, and general understanding of the program, the inspector concluded that appropriate emphasis on liOPs exists within the program. The AO
,
training program appears to be a notab!c strength.
'
3.3.4 Operator Performance Although the primary focus of this inspection was to validate the emergency procedures, several observations were made regarding operator performance. The inspectors determined the operating crews performed very well during the three simulator drills observed on
November 10. The operators demonstrated a good familiarity with plant controls and
,
procedures used to combat tbc events. The inspector noted good command and control of the board operators by the SCO and good practices in implementing the EOPs, such as the
" repeat back" of commands. The shift supervisor provided good support to the crew in the STA and I)SEO function, as well as oversight of crew activities. The inspector also noted the AOs performed well during the system walkdowns in the plant and demonstrated a very good knowledge of plant, procedures, and equipment.
3.4 Summary The inspector concluded that sufficient manpower exists to accomplish critical tasks during a S110. Ilased on either the in-plant walkdowns or the simulator scenarios, time critical steps in procedure !!CA 0.0 were accomplished within the required time. All functions can be i
completed within the time needed assuming only 2 AOs are present on shift, which is the
.
minimum shift complement required by the Technical Specification 6.2.2.a. The CYAPCo shift roster includes 3 or 4 AOs per shift and three AOs are usually on shift.-_While the third
.^
AO on any shift might not be fully qualified, the extra AO represents a valuable resource to a
'
the SS/SSCO for completing in plant actions.
For the non-faulted S130 scenario, most of the manual actions need to be completed by the secondary side AO. The primary side AO could assist the secondary AO as needed. The insight realized by this inspection was that actions by the SS/SCO to prioritize and coordinate the AO activities will be more critical under faulted SI30 conditions, such as in scenarios
. where the plant operator must locally start and manually control the AFW system, fires, etc.
~
'
.
The inspectors identified minor human engineering deficiencies during in-plant walkdowns, weakness in the llPIPs for alternative site communications during a Si!O, and a improper basis for a special test acceptance criteria. None of the identified deficiencies were
,
A
'.
.
.
.
.
____.a__,____
. _ _, _ _ _ _ _
, -
-
_
_ _ _ _ _ _ _ _ _ _ _ _ - _
. _..
-
.
.
determined signincant in the mitigation of a postulated Silo. The inspector noted the licensee took prompt actions to address the concerns. The inspector concluded the AO training program ; sced appropriate emphasis on the emergency operating procedures.
4.0 M AINTl;N ANCE (fi2703)
The inspectors observed various corrective and preventive maintenance activities for compliance with procedures, plant technical speci0 cations, and applicable codes and standards. The inspectors also verified appropriate use of safety tags, equipment alignment and use of jumpers, radiological and fire prevention controls, personnel qualifications, and post-maintenance testing. Portions of activities that were reviewed included:
_
CY-92-12002 llorie Acid Filter Replacement
The inspector evaluated the boric acid filter design features, specific corrective maintenance, the maintenance history, and past CYAPCo corrective actions. The boric acid Glter provides
,
particulate filtration of the boric aciu pump discharge prior to the volume control tank or the charging pump. The filter consists of eleven elements each rated for twenty micron filtration.
The filter media is wound polypropylene within a stainless steel core. A two inch manual bypass allows for Elter replacement during operation. The primary purpose for the Olter is to remove particulate within the boric acid system before it enters the reactor coolant system.
The safety function of the boric acid system is emergency borating of the reactor coolant system following an uncontrolled positive reactivity addition. Technical specifications - 3.1.2.2. defines operability of the system. The filters are within the flow stream of the horic acid system.
The need for corrective maintenance under authorized work order (AWO) CY 9212002 was identined during replacement of the Alter head gasket on December 10. CYAPCo identified
-
degradation of the Olter elements. The degradation consisted of unraveling and partial loss of the wound polypropylene. CYAPCo initiated plant information report (PIR)92-198 to document degradation of the Olter element. The maintenance personnel replaced the Olters and the short term corrective actions will be to replace the filters on a three month basis.
The inspector reviewed past maintenance activities and found that the Olters had been replaced due to degradation of the elements four times since March 1988. The last replacement was in December,1990. CYAPCo corrective action in June 1990 included changing the filter material from rayon to polypropylene under plant modification PDCE 90 098. The basis for the material replacement was a recommendation from the filter vendor (LARCo) to CY A PCo engineering. The basis for the vendor recommendation was a postulated bacterill attack of the rayon material. The new material (polypropylene) has degraded twice since replacement. The inspector concluded that past licensee corrective actions did not prevent filter degradation. At the end of the inspection period, the licensee
,
review to identify a root cause for filter degradation was in progress as part of the PIR r
... _ _ _ _ _.
_
_
_.
_
_ _ _ _
,
.
follow up actions. The periodic replaecment schedule noted above was based the lack a known failure mechanism.
The inspector reviewed the chemical effects of the filter media on the reactor coolant system (RUS). llased on discussions with CYAPCo chemistry personnel, the instrctor noted that the chemical bonds of the polymonomere break down due to radiolysis. The polymonomere breakdown would be evident during chemistry monitoring with an increase in conductivity and organic spikes on the ion chromatograph. No impact on RCS components is espected.
Based on the review of RCS chemistry results during the present cycle of operation, no significant changes in either of these parameter 3 have been identified by the licensee.
Ihe inspector reviewed ti e licensee's control of boric acid system operability during filter conective maintenance. The system has been maintained operable based on maintaining the manual tiller bypass valve open. Inspector review of technical specification 3.1.2.2., and system design response found this action to be acceptable. Ilowever, system reliability is lessened due to the need for recurrent corrective maintenance on the litter, and the prob! cms measuring filter performance (report 50 213/92-20). Routine NRC inspections of plant operations include a review of the availability of boric acid systems and licensee actions to assme operability of the flow path.
CY-9211485, 'll' Purge Fan Discharge Damper
On November 18. the inspector observed portions of the corrective maintenance on the primary ausiliary building ventilation system. The inspector determined that the maintenance personnel were knowledgeable of the task, the equipment tag-out was acceptable, and operability decisions were appropriate.
5.0 SAFETY ASSESSMENT AND QUAL.ITY VEltlFICATION (71707,90712, und 90713)
5.1 Plant Operations itetieu Committee The inspectors attended several Plant Operations Review Committee (PORC) meetings and verified member attendance satislied Technical Specification 6.5 requirements. The meeting
-
agendas included procedural changes, proposed changes to the Technical Specifications, Plant Design Change Records, and minutes from previous meetings. PORC meetings were
"
characterized by frank discussions and questioning of the proposed changes. In particular, consideration was gisen to assure clarity and consistency among procedures, items for which adequate review time was not available were postponed to allow committce members time for further resiew and comment. Dissenting opinions were encouraged and resolved to the satisfaction of the committee prior to approval. The committee closely moni ored and evaluated plant performance and conducted a thorough self-assessment of plani rtivities and programs. The inspectors concluded PORC was effectiv.. -. _ _ _ _ _ _
_ __ _ _ _ _ __.__ _ _ _ _ m _ _ _ _ _
j
!
,
i
.
1 5.2 Reglen of Written Reports
-
'
The inspectors reviewed periodic and Licensee Event Reports (1,liRs) for clarity, validity, accuracy of the root cause and safety signincance description, and adequacy of corrective
!
action. The inspectors determined whether further information was required. The inspectors also verified that the reporting requirements of 10 CFR 50.73 and Technical Specification 6.9 had been met. The following relmrts were reviewed:
LER 92-16-01, Blocked Air intakes Render lloth Emergency Diesel Generators Inoperable The licensee submitted this supplement of LER 92-16 to update the root cause and corrective actions. The licensee conducted additional testing of the Emergency Diesel Generator (EDG)
cubi le ventilation flow. The inspector reviewed the results of the four EDG ventilation flow tests conducted by Thomas Young Associates (QA vendor) on July 1, July 7. July 21, and
'
August 25. 1992. The results showed that with both EDGs running, the bird screen installed, the motor operated intake damper installed, and the north and south EDG niom access door shut, the ventilation system was not capable of supplying enough air flow to the EDG room.
'
The heat load from the EDGs would result in excessive EDG temperatures, which could cause liDG failure and prevent the diesels from achieving required accident loading. The testing also verified that with both south access doors open, the ventilation system is capable of providing enough air flow to maintain the EDG room temperatmes within the de'.ign limit of 140 degrees Fahrenheit (F).
The inspector reviewed the planned modifications for the EDG air intake portion of the ventilation system described under project assignment No.92-043, " Addition of Lou-vers / Dampers to the Emergency Diesel Generator Cubicle " and discussed the plans with the project engineer. The EDG air intake hood will have additional louvers installed to create additional net free area, and existing motor operated intake louvers will be replaced with less restrictive back draft dampers which are fully open when air is flowing, and are designed to go shut when there is no air now for fire protection considerations.
The licensee expects to complete the modifications by the spring of 1993. If the new modifications do not yield the required design air flows, the licensee will utilize a computer model of the BDG room which takes into account the EDG heat loads and actual ventilation flow rates, and calculates the maximum temperature the EDG cubicle will reach with both EDGs running, if this calculated temperature is above 140 degrees 17 (the limit specified in the technical speci6 cations), the licensee plans to conduct further testing to accurately determine the EDG heat loads, and use these refined values in the computer model. The inspector did not identify any areas of concern.
I.liR 92-21, Missed Hourly Fire Watch Patrol The inspector reviewed the LER for timeliness, completeness, accuracy of the event description, the validity of the root causes, and the effectiveness of the corrective actions.
,
_ _, _ _
. _ _,_
__
_
-
.
.
This is the second time in two months that a fire watch did not complete the required rounds in the EDG EG-2A and EG-211 rooms due to personnel error. Although two similar LElts in a short time do not constitute a trend, this reflects poor attention to detail. The Operations Manager implemented highly structured controls to prevent the recurrence of missed fire watch patrols. The inspectors obser ed the new Orc watch controls on several different occasions, and concluded the licensce's corrective actions were satisfactory.
Iladdam Neck Monthly Operating Report 92-08 Iladdam Neck Monthly Operating Report 92-09 The i: spectors determined the reports reviewed were acceptahle based on required information, content, and quality.
5.3 10 CFR 21 Repo 1: Seismic Quntilleation of Foxboro N.2AO-V2il Module On November 30, CYAPCo documented PIR 92-193 to identify a notice from the vendor (Foxboro) concerning seismic qualification deficiency of SPEC _200 N-2AO-V2H modules.
Specincally, during simulation of a seismic event, two capacitors broke off the module card.
The capacitor failure could result _in electrical shorts to other components within the card or
components in adjacent modules within the nest. Also, the output voltage ripple could be i
approximately ten percent (five volts for a 0-60 volt range).
The inspector reviewed licensee actions in response to the 10 CFR 21 report. CYAPCo identified three applications using the Foxboro modules. The applications were auxiliary feedwater (AFW) How indications, rod control system, and main steam dump pressure and temperature indication. The safety-grade application requiring seismic qualification was the -
AFW flow indication. The specific modules are for signal condition and electrical isolation, and were installed under plant modification 87-860 in 1987. CYAPCo evaluated the physical
!
location of the modules within the cabinet and determined that failure would not effect -
adjacent modules either due to no safety function, or the SPEC 200 enclosure design.
Additionally, based on the amount of output voltage ripple, no noticeable change in AFW flow indication would be expected during the postulated scismic event.
CYAPCo maintained the instrument channels operable based on the above description, CYAPCo corrective action included purchase and installation of the strain relief adhesive kit recommended by the vendor, inspector review of the licensee's operability basis concludes that it was appropriate, Corrective actions were based on the vendor recommendation.
5.4 NU Performance Enhancement Program During a facility tour on November 17, 1992, several CYAPCO managers and supervisors were interviewed as part of an effort to assess NU management's efforts in communicating i
'
L I
- - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _.. _ _ _ _ _
_ _ _ _
__
_ _ _ _.
_ _ _ _
,
the pur}w.e, contents and results thus far of the Nll Performance Enhancement Program (Pl!P).
All employees interviewed were familiar to some degree with PliP and its contents. There was good support for the program at the manager level; program support was less cetain at the supervisor level. limployees were familiar with the PEP through either presentation provided by the Senior Vice President, by direct participation in plan, by reading newsletter or other written conununications, and/or by reading the plan manual.
Some positive impressions were shared with the NRC. The PliP was viewed as beneficial by increasing needed staffing at the site, particularly in the engineering and niaintenance areas.
The PliP was viewed as having the potential to help improve performance. Some concerns were also expressed, including how the PEP would impact the availability of supervisors and managers needed to run the plant, and whether the plant staff would be adequately prepared for the next refueling outage. A concern was expressed that the PliP was not necessary.
Overall. it appears that CYAPCo could do more to communicate with the plant staff about the PEP, including the status of its implementation, how the program is expected to benefit them, and to re-emphasize the PliP is an " enhancement" to already good performance at lladdam Neck. The above observations were presented to CYAPCo station management in a meeting on November 18. 1992. The implementation of the PEP will be considered further in future NRC reviews of performance at NU facilities.
6.0 ENIT NIEETINGS During this inspection, periodic meetings were held with station management to discuss inspection observations and findings. At the close of the inspection period, an exit meeting was held on December 18, 1992, to summarize the conclusions of the inmection. No written material was given to the licensee and no proprietary information celated to this inspection was identified.
Entrance and exit meetings were also held for the following inspections conducted by Region I based inspectors during this report period.
Inspection Reporting Areas
!M wtNo.
.!WL Inspegar hLsprcted 50 213/92-22 November 16 -
1. Cheung Engineering Open 20. 1992 Item Followup 50-213/92-24 November 30 -
P. Patniak Engineering and l>ecember 4,1992 Technical Support
_.
.-
-
-
____-__- - - ______ -
- - -
- _
...
~ _ _..
-
-
. --.-
.. _.
..
.
I i
<
.
ATTACilMENT I ECA 0.0, " STATION IILACKOUT" SIMUI,ATOR TIME I,INE
.. - -
ll\\lf(MIN)
l'CA 0.0 S11'P PR4 RANK 11:NC110Ns'CO%IMI.N1S o
Kil'P O
. Sito T(0)is 9:10 m.au.
.
1.24 S11:P 1 VI:RilY REAC10R & lt'RillNE1 RIP
SIEP 3 MI'Dil:M
%TRilY RCS 15 IM)l All:1)
-.
2.25 Nil P 4 IllGil VI:Hil11.D AlW START AND ll AVI: i1 A)W
,
Sil P $
lilGli 1RY 10 Hl3 TORE PWR FRO %1 CON 1ROI, ROO\\l.
l'All.l:D 32%
Sil P o SilllT SL'PYR lilSPA1Cill:S AO 101:IMI ROO%1
'
h.75 S11'P 6 Ml.lilCM C RO PI ACl31 CCS In 1 RIP.PL't.l AIUT
Sil P N tilGil AO Sr.NT 10 ISOIATI: RCP SEAI4 2 74 Sil P 9.a tilGil Mst TRIP VAI.VI:S Cl.OSI:D 9 73 5 IEP 9.d Mi 111.0% DOWN M ant'Al,1S01, VAI,Vl3 Cl OSED
Si e l' e RCA de4ta.T 15 GOOlbNATL RAI, CIRCCIATION SAT to S11:P 10.a ll1011 CI AISE dol % El.1, M4KIIP VAI,VI: STOP 1)WST DRAIN 11.5 N11 P 12 MI:Dit:\\1 VI RllT SG 1L'IlE INTL'GRITY
Kil:P c AO ACllONS PRIOR 11I/.ED PACI:IS SIA) WING lsO% N
SI EP 13.h lilGli SG l.EVi1.S< 6% - IlOI.D COOL.DOWN t
14 24 Sil P o Sill: ARfA 1;MFRGENCY DlIIAl4ED 14.25 S11:P o MI:D!t!M RCP SEAL.S ARI: ISOI ATF.D (VN41 IS CI AKED)
13.5 S1EP o Al W ll,0W IA) cal, CONT (01, @ 640 GPM 19.25 STl:P 16 SG l.EVI:1.5 SAT SG#1 @ 44 AND INCREASING 2u S1EP16 Ki ART ATMOS IDCMP val,VI: CONTROL. IN MANUAL, 21.5 STI:P 17 REAC10R SUllCRITICAl, Dt' RING Cl.1)N NAT CIRC IS SAT
,
I i
STI:P IN VERIIT SI REl,AYS R13ET l OR COOI.DOWN 22.5 5 Il'P 19 tilGil DISPAlfil AO IUR ClV I NCI.D VAI.V13 IN TRENClll3
- 24 25 KlVP 16.h START DI: PRESS TO 260 PStG I
STEP 14.c MEDit:M l>C llOSI'S AI.lGNI.D
S'l1:P 19 CONTAINMI:NT ISOI, VI:Hilli:ll llY CRO -
l
-
-
.
!
STEP 20 tilGli CTMT PRESS < 45#1 RCS 32E SCllCOOI.ED
SIEP 22 Cili,.i i OR AC REKIOHl:D Pl ANT IS STAllt.E
-.
.
, - - _ _ _ _ _ _ _..
,. _, ~,. _ - _ _. _. - - _ -. _.
..,,,
.. -.,. ~.. - -
. ~.
~ --
=
_~ -..-.-.- - - - -
-. - - - -
- - - - - -...
-.
--
.
'
a
,
r t
,
e i
listi-: (MIN)
I'CA 0.0 S11.l'
l'HA HANK itNC110WCOMMl_NTN O
Ni l'I' O SlH) 1(OI is 9:10a.m.
l
1 24 N'l l l' I VI:RilY Hl:ACIOR A 1L dill %I:1Rll'
CN1G 22.A.4 MON 110R RCN. l'ACI: IN NI AlW: RCN Wil.t. WAIT Jn sil:l* o
'A' l'.ly; NT AH1s. Al.l Am t li A Ig; 10 RI:1t'RN t
N'l l'.l* 2 4 111'S 14)%l'.RI:ll A111'M 1;lWi Ki ART t
l'. Nil S10l* NClWARIO
.n
..
,
.
t
)
,..
..~.r,-.
_...,_.;,.._m-,,_,..,.y.
,_.y.m..,,...y.,_,._,,
7.,
. _,.,.,
,,m,.wm,,___.,,_m,m_._
,__.,
,m..,,.._.O,....__,,,m u.m,_._,.--. - r %
-