ML20198A444
| ML20198A444 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 12/23/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20198A404 | List: |
| References | |
| 50-282-97-19, 50-306-97-19, NUDOCS 9801050372 | |
| Download: ML20198A444 (22) | |
See also: IR 05000282/1997019
Text
_.
. _ . .
_.
_
. _ _ .
_ . . _ .
..
. . _ . _
_ _ - _ . ._
l
.-
.
I
U.S. NUCLEAR REGULATORY COMMISSION
,
REGION 111
,
!
i
Docket Nos:
50-282: 50-306
License Nos:
i
Repod Nos:
50-282/97019(DRS); 50-306/97019(DRS)
,
Licensee:
Northern States Power Company
.
Facility:
Prairie Island Nuclear Generating Plant
Location:
1717 Wakonade Dr. East
Welch, MN - 55080
!
Dates:
September 29 - October 3, ar:d 21 23, December 2,1997
Inspectors:
H,- Peterson, Reactor Engineer, Lead Inspector
R. Bailey, Reactor Engineer
~;
S. Ray, Senior Resident inspector, Prairie Island
.
P. Krohn, Resident inspector. Prairie Island
i
- Approved by:
M. Leach, Chief, Operator Liceasing Branch
Division of Reactor Safety
P
!
l
AN! o
G
.
..
.
-
.
-
-.
.
. . . . -
.
.
.
--
-
-_.
_ _ _ _
_ _
_.
__ _-
. __
._
_ _ . . . . _ -
_ _ .
._
.
,
EXECUTIVE SUMMARY
Prairie Island Nuclear Generating Plant
NRC inspection Reports 50-282/97019; 50 306/97019
This inspection report contains the findings and conclusions from the inspection of the licensed
reactor operator (RO) and senior reactor operator (SRO) requalification training programs. The
inspection included a review of training administrative procedures and operating exvnination
material, observation and evaluation of operator performance end licensee evaluators during a
requalification operating examination; an assessment of simulator fidelity; an evaluation of
,
!
program controls to assure a systems approach to training; and a review of requalification
training records, in addition, the inspectors observed a period of control room operations. The
inspectors used the guidance in inspection procedures (IPs) 71001 and 71707.
'
The licensed operator requalification programs were implemented in accordance with 10 CFR Part 55 requirements.
All portions of the annual requalification examination were judged to be effective tools
for determining operator weaknesses (Sect;ons 05.2,05.3).
Control Room operators demonstrated an appropriate level of. attentiveness to the
e
operating panels and were knowledgeable of plant conditions (Section 01.1).
Licensee controls to revise the licensed operator requalification training program were
satisfactory (Section 05.4).
.
The licensee's remediation program contained adequate measures to ensure individual
and crew performance weaknesses were addressed prior to resumption of licensed
duties (Section 05.5).
However, weaknesses were identified with regard to the following:
Communications were at times informal and did not always meet management
expectations for three way communications (Section 01.1).
There was a lack of formal controls to restrict personnel occess to vital control areas
within the control room (Section 01.1).
A
e
The licensee implemented an inadequate procedure which circumvented required EOP
steps. This was considered a vlotation of 10 CFR Part 50, Appendix B, Criterion V,
" Instructions, Procedures, and Drawings." The licensee's interpretation of what
constituted an entry condition for the reactor trip emergency operating procedure was
incorrect (Section 03.1).
The licensee was implementing SWis as the underlying procedure in lieu of approved
and OC reviewed procedures. This was considered a violation of Technical
i
!.
Specification 6.5, * Plant Operating Procedures," (Section O3.1).
l
,
l-
2
i
l
-
.-, -.
- .
- - . .
-.- . - ,.- - -.
- ---.
_ _ _ _ . _ _ _ _ _ . -
-_ ... _ _. . _ _ . _ _ . . _ . . _ _ _ . _ _ _ . _ . _ . _ . _ . . _ . _ . _ _ _ - -
.
,
o
The licensee's use of the * dual-role * SRO/STA could potentially impair crew
j
performance (Section 04.1).
i
e
The licensee continued to demonstrate difficulties in procedure use (Sections 05.1,
l
05.4,05.7).
i
The licensee's instruction for Fire Brigade personnel on respirator fit qualification was
e
clear, but no such guidance or instruction was in place for all other licensed operators
(Section 08.1).-
l
i
!
I
!
i
I
.
?
. I
i
I
6
h
i
F
$
t
.
I
h
I
.e
..
t
!
p
I
l
3
.
t
t
I
!
.
- .
. . .
.-s,
.
.m..-.
. - - - - ~ , - - ~ ,
, - . . , ~ - - , , _ , ~ .
, . - . -
. -,. -l,4--,_,m
-_, _.,,. ....,_, -
,,.-m.?
_ - _ - __________ _ _
_ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - -
_ _ _ _____ _ _ _ _ _ _ _
_ _ _ _ _
_ _ _ - - _ - _
.
!
.
.-
Reports Details
{
.
l. Operations
t
01
Conduct of Operations
01.1 Control Room Observations
a.
Insoection Scone (71707)
The inspectors observed routine control room activities during full power operations,
performed a dual unit panel walk-down, reviewed control room logs, and questioned
operators about plant and equipment status,
i
s
b.
Observations and Findings
in general, the control room operators conducted themselves in a professional manner
and were silentive to their respective panelindications. Howe ser, access control to the
r
'
main controi room area and verbal communication practices we.1 not always consistent
with management expectations and guidelines. Conversations co.'tained informal and
incomplete three way communication phrases. Access to the main control room area by
non-licensed individuals occurred repeatedly without challenge. On two occasions, one
individeal approached the control area of one unit without challenge, and another
entered the back panel area without challenge.
!
The control room noise level was minimal and no annunciator alarms were left
I
unattended or in a prolonged alarm state. Upon questioning by the inspectors, the
operators demonstrated satisfactory knowledge of plant conditions and equipment
status,
c.
Conclusions
The inspectors concluded that, in general, an appropriate level of awareness existed in
the control room. However, the 'nspectors were concemed that operator performance in
the area of verbal communications was not always consistent with management
expectation and guidelines, and that a lack of formal controls to restrict personnel
access to vital control areas existed.
I
03
Operations Procedures and Documentation
.
03.1 Section Work Instructions Vice Emergenev Ooeratina Procedures
a.
Insogelion Scone (71707. 71001)
The inspectors observed the licensee's evaluation of three operating groups during the
requalification examination simulator scenarios. Based on these observations, the
4
.I
- -
. . - . -
. . .
~ . .
.
.
Inspectors identified inappropriate use of section work instructions (SWI) in lieu of
properly following an Emergency Opeiating Procedure (EOP). Also, the inspectors
identified an inadequate interpretation of an abnormal operating procedure (AOP) step.
The inspectors reviewed licenses procedures and instructions,10 CFR Part 50, EOP
emergency response guidelines, and technical specifications to ascertalri the safety and
regulatory impact of SWis.
b.
Observations and Findings
On September 30,1997, during one simulator scenario, a seal failure on a reactor
coolant pump required entry into abnormal operating procedure 1C3 AOP3, * Failure of a
Reactor Coolant Pump Seal, * Revision 3. Based on the pump sealleakoff of greater
than 6.0 gpm and increasing seal outlet temperature, the crew correctly determined to
perform step 2.4.5.C of procedure AOP3. The procedure required tripping the reactor
within 5 minutes, then initiate EOP 1E-0, " Reactor Trip or Safety injection." The crew
appropriately attempted to initiate a reactor trip using the manual reactor trip switches;
however, the reactor fai'ed to trip and the plant was in an Anticipated Transient Without
Scram (ATWS) condition. The crew, however, determined not to enter E-0 or perform
the immediate actions of 1FR S.1," Response to Nuclear Power Generation /ATWS,"
which included manual trip of the main turbine instead, the crew proceeded to allow
the reactor to continue to operate until the reactor trip breakers were manually opened.
,
Additionally, the crew assumed that they had 5 minutes to implement opening the trip
breakers. The crew incorrectly interpreted the step," trip the reactor within 5 minutes,"
as having 5 minutes to enter E 0, even after an ATWS conditfon was identified. The
reacto" trip breakers were eventually simulated open, the reactor tripped (control rods
inserted), and the crew entered E-0. However, the crew did not appropriately enter E-0
and perform the required actions of FR S.1 at the time of the ATWS condition. Of note,
was the fact that during the period the crew were trying to open the trip breakers the
reactor coolant pump tripped which generated an automatic reactor trip. This was not
identified by the crew for about 40 seconds.
The inspectors questioned the licensee on this incorrect practico in recponding to an
ATWS condition and were informed that SWI O 10," Operations Manual Usage,'
Revision 29, dictated the crews actions (original revision that incorporated the incorrect
,
inst uction was Revision 28 dated July 24,1996). Procedure SWI O-10, Section 6.11.9,
' Anticipated Transient Without Scram (ATWS)," contained instructions that when a
conservative decision was made to initiate a reactor trip, based on deteriorating plant
conditions, but before a reactor trip setpoint was reached, then the reactor should be
tripped manually using one of the two trip switches. If this was unsuccessful, then the
I
trip breakers should be opened locally in the rod drive room. Section 6.11.9 went on to
say that if a valid protection system setpoint was reached and the reactor trip breakers
were not open and cannot be opened manual!y, then FR S.1 shall be entered from E-0.
The procedure justified this course of action by stating that the turbine trip step of FR-
S.1 imposed a significant t,'ansient on the plant, and that it was desirable to avoid this
!
event unless absolute!y necessary. The licensee, therefore, was interpreting that an
!
ATWS condition did not exist, even if a manual attempt to trip the reactor failed using
l
l
l
!
5
_
_
.
the trip switches, ur,less an automatic trip safety signal was present and if the trip
breakers could not be opened manually.
The Code of Federal Regulations Title 10, Part 50, Appendi:: B, Crite. ion V,
- Instructions, Procedures, and Drawings," states, in part, that actMiles affec'ing qustity
be prescribed by documented instructions and procedures ci a ype appropriate to the
circumstances and be accomplished in accordance with thM instructions or
procedures.
The inspectors identified that the actions, dictated by procedure SWI O 10, to open
reactor trip breakers to correct a malfunction of the trip switches pitor to entering any
EOPs and not to consider the event as an ATWS condition, was inadequate and directly
conflicted with approved EOPs. The EOP 1(2)R-0," Reactor Tnp of Safey injection,"
Revision 17, Sectb.n A and Step 1, required entry into the procede.b following a manual
or automatic actuation of a reactor trip and entry into FR-S,1 on failure of a reactor trip,
respectively. The EOP 1(2) FR S.1, * Response to Nuclear Power Gercratic1/ATV/S,"
Revision 8, Step 2 required an immediate turbine trip for any A1WG conaition. The
action to manually open reactor trip breakers was a follow up action in Stap 5 of f RC
The inspectors determined that the licensee implemented an inadequate procedure,
SWI O 10, which circumvented required EOP steps, and this was a violation of 10 CFR Part 50, Appendix B, Criterlon V," Instructions, Procedures, and Drawings," (VIO 50-
282/306-97019-01(DRS)).
Technical Specification 6.5 required that detailed written procedures, including the
applicable checkoff lists and instructions, covering areas listed be prepared arJ
followed. The specification further required that the procedures and changes thereto be
reviewed by the Operations Committee (OC). Areas listed under Plant Operations
included the fo' lowing: (1) integrated and system procedures for normal startup,
operation and shutdown of the reactor and all systems and components involving
nuclear safety of the facility; (2) fuel handling operations; (3) actions to be taken to
correct specific and foreseen potential or actual malfunction of systems or components
including responses to alarms, primary system Icaks and abnormal reactivity changes
end including foilow-up actions required after plant protective system actions have
initiated;(4) implementing procedures of the fire protection program.
The inspectors identified that 0, SWis were not part of the list of procedures described
in the administrative section of the technical specification, and therefore were not
routinely reviewed by th9 OC; and (2) the licensee was using SWis to implement actions
outside of OC reviewed procedures. The examples included procedures associated
with fire brigade duties, fuel hanoling, and normal and off-normal operations:
SWI O-1, " Work Rules and Philosophy for Operation of Nuclear Plants," Revision
9, dated July 17,1997, Section 6.2, contained the requirement ihat the iPo
brigade and fire brigaae support personnel be clean-shaven at the start of the
shift. This instruction was prcvided to ensure personnel were able to wear
bre1 thing apparatus. The instruction was not contained in any OC reviewed
6
.
procedure and was considered an implementing procedure of the fire protection
program which required OC review.
SWI O-1, " Work Rules and Philosopny for Operation of Nuclear Plants," Revision
9, Section 6.6, contained specific instructions for monitoring plant parameters
after a reactivity manipulation. The instructions includert a list of instrumentation
to observe. The instructions were similar to but much more detailed than
instructions in OC reviewed procedures for plant operations. In fact, one
operating procedure, C12.5, * Boron Concentration Control," Revision 6, Section
3.2, referred the operator to SWI O 1 for instructions for monitoring the effects of
boration and dilution. The SWI w&s considered a procedure for normal operation
of the reactor which required OC review.
SWI O 1," Work Rules and Philosophy for Operation of Nuclear Plants,' Revision
9, Section 6.7, contained specific instructions and limitations for maintaining
rated thermal power within the limits of Technical Specifications and NRC
guidance. The instructiuns included a list of instrumentation to use and time
limits for inadvertent operations above 100 percent of rated thermal power. The
instructions were similar to but much more detailed than instruction; in OC
reviewed procedures for plant operations. The SWI was considered a procedure
for normal operation of the reactor which required OC review.
SWI O 10, * Operations Manual Usage," Revision 29, dated March 24,1997,
Section 6.4, contained instructions that upon hearing the announcement of a
reactor trip, turbine building cperators were expected to initiate isolation of the
moisture separator - reheaters, by memory, per Attachment J of the emergency
operations procedures (EOPs), without awaiting further direction from the
Control Room. That instruction, to perform the activity without further direction,
was not contained in any OC approved procedure and EOP 1(2)E-0, * Reactor
Trip or Safety injection," Revision 17, Step 8, to notify the turbine building
operator to perform Attachment J, was not considered an immediate action to be
memorized. The SWI was considered an instruction for follow-up actions
required after plant protective systems actions have initiated which required OC
review.
SWI O-10, " Operations Manual Usage," Revision 29, Section 6.7.5, contained
instructions requiring that four specific checklists for establishing containment
integrity be completed twice whenever they were performed to satisfy a
procedural requirement. That instruction, to perform the checklists twice, was
not contained in any OC reviewed procedure including the checklists
themselves. The SWI was considered a procedure for normal operation of a
system involving nuclear safety of the facility 'vhich required OC review.
SWI O-41, " Duties and Responsibilities of Fuel Handling Personnel," Revision 4,
dated July 17,1997, Section 6.1,7, contained a requirement for a fuel
accountability engineer to concur with fuel moves before any assembly was
placed into a new location in the spent fuel pool or before the spent fuel handling
7
l
__
__
__
_ . _
_ _ _ _ .
_ .
.
_ _ . _ _
. _ _ - _
. _ _ _ _ _ .
.
.
.
,
tool was lowered on to an assembly in the spent fuel pool. That requirement was
not included in the fuel handling procedures reviewed by the OC. The SWI was
considered a procedure for fuel handling operations which required an OC
review.
SWI O-41, * Duties and Responsibilities of Fuel Handling Personnel," Revision 4,
Section 6.2, included several requirements for communications, verifications,
and permissions for fuel handling operations that were not included in the fuel
handling procedures reviewed by the OC. The SWI was considered a procedure
for fuel handling operations which required an OC review.
Failure to perform an OC review for the above SWis was a vlotation of Technical Specification 6.5, * Plant Operating Procedures," (VIO 50-282/306-97019-02(DRS)).
c.
Cot.alusions
The inspectors identified two violations of NRC requirements. The first involved the
licensee's interpretation of what was an ATWS condition and how to mitigate such a
condition was incorrect. This demonstrated a fundamental misunderstanding with
,
respect to procedure adherence. The second involved implementing operational
procedure steps through a means which was not reviewed by the Operations
Committee.
04
Operator Knowledge and Performance
04.1 Shift Manager As Shift Technical Advisor
'
a.
Insoection Scone (71707. 71001)
The inspectors reviewed the licensee and industry documents associated with licensed
shift personnel duties and responsibilities, shift organization, onsite emergency
organization, and operatin0 experience and events. The inspectors reviewed the
following documents to assess operator roles and responsibilities:
Technical Specification Administrative Section 6.1, ' Organization," Revision 105
.
Technical Specification Table 6.1 1, " Minimum Shift Crew Composition,"
'
Revision 105
Section Work Instruction (SWI), SWI O-2, "SNft Organization, Operation &
Turnover," Revision 36
Emer0ency Plan implementing Procedure, F3-1, *Onsite Fmergency
Organization," Revision 14
NUREG-1275, " Operating Experience Feedback Report - Human Performance in
Operating Events," Volume 8, December 1992
Information Notice (IN) 93-81," Implementation of Engineering Expertise on
Shift," October 12,1993
8
_
-
.
_
_
_
_
-_
_
___.
\\
.
b.
Observations and Findings
The inspectors identified that the Shift Technical Advisor (STA) was also the Shift
Manager (SM), a licensed senior reactor operator (SRO) who was the senior person on
shift. On May 4,1993, a change was added to the Technical Specification Table 6.1 1,
' Minimum Shift Crew Composition, * which noted that the SM performs the functions of
the STA. The licensee's action to use a " dual-role' STA was allowed per an October 28,
1985, Federal Register notice 50 FR 43621, *NRC Policy Statement on Engineering
Expertise on Shift."
Within the licensee's organization, the SM, in accordance with SWI O 2, was
responsible for supervising activities affecting operation of the plant as a whole and has
the ultimate authority and responsibility during routine, abnormal, and emergency
situations.
Also, in accordance with emergency plan implementing procedure F31, the SM has the
initial responsibility to assume the duties of the Emergency Director (ED) during an
emergency event. The Emergency Director's responsibilities were very significant and
included the following: (1) coordinate response of the plant onsite emergency
organization, (2) emergency classification and notification of offsite authorities, (3)
authorize offsite Protective Action Recommendations, (4) direct the activation of all
onsite emergency response centers, (5) direct plant evacuations and personnel
accountability, (6) authorize radiation exposure in excess of normal limits, and (7)
ensure onsite and offsite radiological monitoring initiated.
The inspectors were informed that the SM position was the only position on shift that
received the STA training. In reference to NUREG 1275, the function of an STA was to
objectively evaluate the plant condition durir,g abnormal and accident condtlons and
recommend action. The STAS were to have a bachelor's degree in engineering or
equivalent to render engineering technical advise during an accident. Furthermore, the
STAS for Westinghouse facilities perform the safety function as independent eyes during
an accident and review plant status per the emergency procedure functional status
trees.
After reviewing NRC documents (IN 93-81 and NUREG 1275) pertaining to operating
experience concerning multiple-role SRO/ STAS, the inspectors identified that problems
have occurred at other facilitics which rcsulted in overburdening the SRO/STA while
fulfilling duties involving EOP reading, event classification, fire protection concerns, and
implementation of the emergency plan. The licensee stated that the SMs role as
Emergency Director was supported by the SRO from the unaffected unit. This individual
prepared the necessary paperwork for the SMs review and approval. The viability of
this operating structure will be reviewed during a future inspection (IFl 50-282/306-
97019-03(DRS)),
l
l
l
9
l
l
l
.
_
._
.
.
.
c.
ConrJusjon
The inspectors concluded that the licensee's use of the * dual-role" STA could potentially
impair crew performance and this will be reviewed during a future inspection.
05
Operator .'talning and Qualification
05.1 Operatina History
a.
trispection Scope (71001)
The inspectors reviewed the following to assess the licensed operator requalification
training program's effectiveness regarding operator performance:
SALP Report Nos. 50-282/300 96001.
Resident inspector observations and reports covering the time frame of 1996 to
present.
Licensee event reports covering the time frame of 1996 to present.
Initiallicense examination Report Nos. 50-282/306 97306(OL).
Licensed operator requalification training Report Nus. 50-282/306-95013 (DRP).
b.
Observauons and Findinos
The inspectors noted that poor operator performance as documented in the above
reports was attributable in part to incorrect use of procedures or inadequate procedures.
The inspectors noted that the licensee was continuing to take actions to improve
operator performance pertaining to procedure usage. The licensee's action items
'9ntered around procedure use and compliance, cnd overall procedure development.
The inspectors, however, Identified centinued problems concerning proper use of
procedures. In particular, the licensee's over relience on SWis resulted in circumventing
approved emergency operating procedures (see Section 03.1 for details).
c.
Conclusions
The inspectors concluded that the licensed operator requalification program had not
been effective in the past in reenforcing proper procedural usage. Procedures and use
of procedures continued to be a recurring p',oblem. The licensee had recently begun
corrective actions for these problems but these were not sufficiently implemented to
allow for an objective evaluation.
10
-- -
..
- -
- . _ _ . - _ _ . _ ._
_
_ _ _ . _ _ _ _
_ _ _ _ _ _
.
.
05.2 Reaualification Examinations
05.2.1 Examination Material
,
a,
insoection Scoon (71001)
The inspectors reviewed the written and operating examination material with Appendix A
checklists in Inspection Procedure 71001. This review included a comparison of written
i
questions, dynamic scenarios, and job performance measur es (JPM) with previously
administered examinations.
-
b.
Observations and Findings
The dynamic simulator scenarios were comprehensive and provided sufficient
quantitative attributes to evaluate the crew and individual members on safety significant
tasks and competencies. Also, the scenario objectives incorporated PRA significant
events in the examination process. However,3 of the 4 scenarios contained a related
task objective to have the operator diagnose and perform corrective actions for an
'
ATWS event during a failure of the manual, automatic, or both trip protective functions.
The repeated coverage of th9 task was not consistent with the licensee's requalification
training plan, in that, this item was a very low percentage of the total training conducted
this requalification training cycle.
The JPMs contained clearly stated critical steps and termination criteria required foi
successful completion. However, program deficiencies were noted during the
performance of JPMs: (1) more than one JPM had incomplete cues which required the
'
evaluator to improvise; (2) one JPM contained an inaccurate performance standard that
required evaluator judgement; and (3) one JPM contained performance tasks that were
not consistent with the procedure in use which required evaluatorjudgement. The
inspectors were concerned that the review and validation process had not identified
these deficiencies even with two levels of technical revie v being performed,
The written questions were operationally oriented and contained an appropriate level of
difficulty A majority of the open reference questions were cf higher cognitive
,
'
knowledge level, The static examination questions made good use of the simulator as a
,
reference. Less than ten percent of the questions were repeated from week to week,
and one hundred percent of the questions were new or significant;y modified from the
i
previous examination cycle.
c.
Conclusions
,
The inspectors concluded that the requalification examintf!on materiel contained the
,
necessaly quantitative and qualitative attributes to provide an effective evaluation of
cperator skills. However, simulatot scenarios contained a disproportionate number of
i
ATWS scanarios and some JPMs did not provide appropriate cues and performance
L
standards.
!
i
-
11
l^
P
- . . . .
- , -
-
c-
. ,, r
4
-a
<-
~
v
_
-.r
.- -.
.
.
OS 2.2 inconomtion of Current Industrv Events
a.
Insoection Scogt
The inspectors reviewed the licens6e's program to assess and ir corporate current
industry events applicable to the facility into training and testing. ParticAr attention
was placed on recent industry concerns on the capability of timely perfo ming the
emergency operating procedure for a steam generator tube rupture (SG 'R) bat M on
the Prairie Island Updated Safety Analysis Report (USAR), Section 14, R wision 13,
time criterion. The inspectors also reviewed Emergency Operating Procedure 1E-3,
" Unit 1, Stehm Generator Tebe Rupture," Rev.13.
b.
Observations and Findinos
The inspectors identified that the licensee, on July 29,1997, initiated a non-
conformance report which noted that isolation of SGTR by ooerators may exceed the 30
minutes USAR assumption. The non-conformance report was in response to an NRC
daily event report that described another facility's problem in meeting the USAR time
requirements for terminating SGTR flow during simulator training. *ihs licenseo
conducted timed simulator SGTR training during the week of July 7,1997. Four
different operating groups were given evaluation scenerlos that included a SGTR with
loss of offsite power, The four crews ps;fomed tht, SGTR mitigating actions to
terminate the primary to secondary leak be inally terminating safety injection in 34,37,
40, and 36 minutes, respectively.
.
Following the evaluation, the licensee concluded that the scenarios were not pre-
evaluated to correspond to licensing basis or design basis requirements, but also
concluded that it was a rough apprrimation of possible response t;mes. The licensee
noted that in all four scenarios, the steam generator (SG) narrow rsnge levels rema!ned
helow 95%, but the times exceeded the 30 minutes assumption in the USAR. The
ilcensee determined that the test was not valid, such that no design basis was
exceeded, Additior, ally, the licensee detertrJned that the USAR 30 minutes time limit
was a conservative time estimate, that as long as radioactive release was minimized the
requirements of the USAR were met.
The licensee informed the inspectors that discussions were held with Westinghouk 'o
resolve the issue of the 30 minutes time limit, and that continued assessment was bbing
made to develop the appropriate simulation scenario tc test the USAR time 'imit.
Following the on site inspectica the licensee informed the inspectors that additional
testing with a rupture elze of about 600 gpm had shown that four of five crews could
meat the 30 minutes specified in the USAR. However, during the perforn.ance of these
scenarios the cre'# had performed the reactor coolant system (RCS) couldown and
depressurization steps concurrently. EOP 1E-3 contained the following caution
statement. *1f SG overfillis an imm3diate concem, THEN cooldcwn and
depressuritation eleps may be performed concunently with shift supervisor ap,3roval."
The procedure does not include steps for performing tnese activities concurremly ev6n
12
.
-
.
.
+
.
__
_
_
..
_
_ _
_
_ _ _ . _
_.
.
.
,
i
' though the bcckground document states "Although concurrent RCS cooldown and
depressurization may reduce the amount of leakage into the secondary initially, it
increases the demands on the operator and may lead to a delay in SI (Safety injection)
i
termination. Furthermore, concunent cooidown and depressurization would also require
!
more precise pressure control to maintain RCS subcooling. Such control may require
cycling of a PRZR PORV (pressurizer power operated relief valve)if normal spray is
unavailable, Careful consideration must be given to concurrent RCS cooldown and
!
1epressurization." The failure to provide specific htructions in EOP 1E 3 and the
i
adequacy of the procedure to meet the USAR time limits is an unresolved item pending
further review of the adequacy of this procedure against 10 CFR Appendix B, Criterion V
(URI 50 282/306-97019-04(DRS)).
c.
Conclusions
i
- The inspectors concluded that the licensee was relylrig on concurrent actions to
cooldown and depr ssurize the RC3 in order to meet the time limits specified in the
'
USAR. The inspectors were concerned over the adequacy of the procedural
instrucilons to perform these tasks concurrently and this w41 be evaluated furthei
05.3 Recualification Examination Administration Practices
a.
Insoecticn Scone (71001-
The inspectors observed the licensee's evaluators during one operating crew's and one
staff crew's performance during dynamic simulator and JPMs. The two crews consisted
of thirteea operators which was divided into three groups. Each group was required to
perform two dynamic scenarios and a set of five JPMs. The inspectors also attended
the crew evaluation ciitiques.
b.
Observations and Findings
The licensee's evaluation team identified no unsatisfactory crew performance.
However, two operators were identified as having demonstrated Mr performance
during JPMs and required follow up training. The evaluators appropriately documented
the opeinters' performance as unsatisfactory on 1 of the 5 JPMs administered (See
Saction 05.5 for a discussion of the remediation process).
The evaluators performed the examination administration in a prriessional manner and
properly documer. led operator performance deficiencies. No evaluator miscuing or
prompting was identified.
Appropriate security rneasures were taken throughout the examination process.
IndMdual operators were sequestered and separated into test groups during ea'.h
portion ">f the examination process. No cxam compromise was identified.
'
No new simulator fidelity issues were identified during the exam observation (See
Enclosure 2, Simulation Facility Report).
L
13
l
j .-
'
,
.
., .- ..
-
.-
.
.-
-_ .
--
.
c.
Conclusions
The inspectors concluded that the licensee was implementing the Licensed Operator
Requalification Training (LORT) program in accordance with program guidance and
regulatory requirements stated in 10 CFR Part 55.59.
05.4
Reaualification Trainina Proarem Feedback System
a.
Insoection Scooe (71001)
. % iraspectors reviewed the following documents to assess the licensee's training
program feedback system effectiveness:
Quality Assurance Audit Report, AG 1996-O-1, for Plant Operations Training
Generation Quality Services Status Report, Second Quarter 1997 (a data
analysis and trending report)
Quality Assurance Procedure 1 OAP 2.8, Revision 7 (requirements for audits)
Program Group Summary
Self-Assessment Operations Training (a self aswssment on the conduct of
classroom training and individualized instruction and trainee eveluation of
Operations Training)
Training Procedure 1.11. " Training Effectiveness Self-Evaluation," Revision 1
dated September 20,1996
Self Assessment Operations Training (a self-assessment on the analysis design
and development area of Operations Training)
Administrative Work Instruction (AWI) - 5AWI 3.15.2, " Employee Ot'servation
Roporting," Revision 0
CHAMPS lssues Module, Revision 1 dated September 1997 (a new program for
assessment and tracking of internal and external 'ssues/ problems)
b.
Observations and Findings
The licensee performed self-assessment activities by assessing identified individual
operator and crew weaknesses, operator training requests, and plant and industry
events. Additional self-assessment processes included Program Advisory Committee
meetings, course evaluations, instructor evaluations, classroom feedback, simulator
evaluations and critiques, and on-the-job training evaluations. Also, the licensee's
Nuclear Quality Assurance group perfomled periodic audits of the Operations t -
Training programs. Subsequently, the Programs Group gathered, evaluated, and
assl ned priorities for the results of all the self evaluations, including those conducted by
0
the Nuclear Quality Assurance group. One continuit.g theme identified through the
Nuclear Quality Assuranca group's audit was procedure weaknesses, including
improper procedure usage and control.
The licensee's self-assessment program appeared to be up to date, and flexible enough
to incorporate emerging training issues. In addition, the licensee had a satisfactory
tracking program to incorporate changes to the examination bank when procedure
14
.
. _
_
. .~ .
._
_ _ _ .
._. . _ _
-_
,
l
.
.
1
changes or modifications were implemented by the plant. During this inspection, the
licensee was updating its job porlormance measures (JPM) examination hank.
,
c.
Conclusions
.
t'
The inspectors determined that the feedback process was satisfactorily implemented.
05.5 Bemedial Trainina Proaram
a.
Insoection Scope (71001)
The inspectors reviewed the licensee's remedial training program and selected records
to assess corrective actions for identified weaknesses in operator and crew
l
performance. This review included an interview with selected personnel involved with
the remedial tralning process.
t
b.
Observations and Findinas
,.
During previous evaluations, the licensee had identified a number of unsatisfactory
performances on both the written and JPM portions of the examination prccess. The
inspectors determined that selected remedial training plans had incorporated =
comprehensive retraining and evaluation process, and were consistent with the
'
licensee's assessment of operator's por performance. The licensee acknowledged that
the poor written examination performance had been attributed in a recent revision in
exam question difficulty which made each one more operationally discriminating.
.
The inspectors noted that the licensee had developed remedial training plans for
{
individuals with demonstrated weaknesses and required successful completion of the
remedial training prior to resuming license duties. The remedial training program
properly identified and corrected licensed operator performance deficiencies.
c,
Conclusions
The inspectors concluded that the remediation program contained adequate measures
to ensure individual and crew performance weaknesses were addressed prior to
resumption of licensed duties.
05.6 Conformance with Ooerator License Conditions
a.
Insoection Scone (71001)
The inspectors reviewed the licensee's med: cal and operator qualification programs and
. selected ret,ords to assess licensed operator compliance with regulatory requirements.
- This review included a sampling (10 percent) of the available medical records. Also, the
- licensee's new procedure for maintaining active operator licenses SWI O-43, " Operator
'
Qualification Program," Revision 0 dated January 24,1997, was reviewed.
.
,
15
,
6
.
.,
m,
,-,
- .
-
.
-
,
w
, -
y,--,3e
.
-
. _ _
.
.
,
1
b.
Observations and Findings
The licensee maintained a copy of individual medical records at the facility. The
inspectors determined that the records contained appropriate documentation to validate
operator qualifications to perform license duties. No physical exam dates exceeded the
program allowed date and no vlotation of regulatory requirements was identified.
On January 24,1997, the licensee implemented a new procedure, SWI O-43, that gas e
guidancu for maintaining operator licenses in an active status. This issue was originally
identified on October 24,1995, by the licensee's Quality Services group during an audit
of the requirements of 10 CFR 55.53,' Conditions of Licenses." The licensee had
occasionally credited working in the work control center (WCC) as " actively performing
the functions of an operator or senior operator" for the purposes of maintaining operator
license in an active status. However, on August 28,1996, the licensee became aware
of another licensee performing the similar practice and found that WCC duty was not
acceptable for credit towards maintaining active license status. Subsequently, the issue
was identified to the NRC in Inspection Report 50-282/306-96008, Section O5.1. Thic
item was later closed in inspection Report 50-282/306-97002, Section 08.3, based on a
letur submitted by the licensee to the NRC stating that they had discontinued the
practice of crediting duty in the WCC as meeting the enteria for actively performing the
functions of an operator or senior operator.
During this inspection, the licensee's new procedure SWI O-43 was reviewed. The new
procedure dictated a strict requirement that for an operator to maintain active license
status, the operator must perform the functions of Control Room Duty Operator or
Watchstander for a minimum of five 12-hour shifts per calendar quarter, even during
outages. The control room positions were specifically identified as the Shift Manager,
Shift Supervisor, Lead Plant Equipment and Reactor Operator, and Plant Equipment
and Reactor Operator,
c.
Conclusions
The it spectors concluded that the operator's !icense conditions were in conformance
with program guidance and regulatory requirements stated in 10 CFR Part 55.53 and 10 CFR Part 55.21.
05.7 Follow uo of Previousiv identified Weaknesses
a.
Insoection Scoce
The inspectors reviewco the identified weaknesses from the last Licensed Operator
Requalification Training (LORT) program inspection (NRC Inspection Report 50-
282/306-95013) to ascertain the licensee's actions to resolve any weaknesses.
16
_ _ _ _
_--- _ -
-
_
,
.
b.
Observations and Findinos
I
-
There were four weaknesses identified in the last LORT inspection report. Ono
weakness was associated with no additional training for individuals or crews that
demonstrated significant weaknesses, but where the overall performance was evaluated
as satisfactory. The lic9nsee initiated " follow up" training for those individuals and
crews.
Two other weaknesses were associated with operator performance in procedure
implementation and communications. The licensee was aware of these concerns and
was actively pursuing the issues; however, procedure problems continued to be a
concern. Also, cominunications did not always meet management expectations.
c.
Conclusions
Although corrective actions were being implemented by the licensee to eliminate
previously identified weaknesses, the inspectors concluded that weaknesses continued
in procedure implementation and communications.
08
Miscellaneous Operations issues
08.1 Resoirator Fit Proaram
a.
Insoection Scoce (71707)
The inspectors reviewed the licensee's plant safety procedure FS, Appendix B,' Control
Room Evacuation (Fire)," Revision 17, for operator actions required during an
evacuation of the control room / relay room area. Expected operator actims were
compared with the licensee's training and qualification program to ensure operator
readiness to perform assigned duties.
b.
Observations and Findinas
The inspectors noted that one of the requirements for a Unit-1 or Unit-2 Shift Supervisor
(SRO licensed) was to pick up a self-contained breathing apparatus (SCBA) and
proceed to an assigned in plant location. During a plant tour, the inspectors observed
tilat some of the licensed senior reactor operators (SRO) had facial hair (beard) of such
length that a proper mask fit would not be possible when donning and using the SCBA.
The inspectors were concemed that any one of these SROs might not be able to
complete their required actions to place the units in a safe shutdown condition following
a fire. The licensee had not put into place any management guidelines to address a
facial hair policy for control room operators except for those (RO licensed only) assigned
to the Fire Brigade. Also, the inspectors identified and informed the licensee of a
concern that an on shift control room operator (RO licensed) with facial hair might not be
able to obtain a proper mask fit if required to don a SCBA to respond to a fire. While the
operator had not been assigned to Fire Brigade duties, the licensee management
17
_ _ _ _ _ _ _ _ _ _ _ - _ _
_- _
. _ _ - _
. _ _ __
_ _ _ _ _ _ . _
_
_
.
.
..
.
-
.
-
.
.
.
acknowledged that the occurrence did not meet current management guidelines. The
licensee implemented immediate corrective action to have the individual shave off all
facial hair ' The licensee management acknowledged the NRC concems and would
review these issues for future corrective action.
Additionally, the inspectors questioned the availability of special corrective lens for
respirator masks and were informed that access to special lens existed. The inspectors
verified that special corrective lens did exist and were available for individual use when
needed. No concems were identified by the inspectors.
c.
Conclusions
The inspectors concluded that, while the licensee's instruction for Fire Brigade
personnel on respirator fit qualification was clear, no such guidance or instruction was in
place for all other licensed operators.
V. Management Meetings
X1
Exit Meetina Summary
The inspectors presented the inspection results to members of licensee management on
October 3,1997, and during a teleconference on October 23,1997. The licensee
acknov4 edged the findings presented.
Durir g a management meeting held on November 25,1997, one of the significant findings of
this inspection concerning ATWS entry conditions was discussed. The disposition of one
.
remaining open inspection item was discussed during the regularly scheduled resident
inspector exit meeting on December 2,1997. No proprietary information was identified by the
licensee during the inspection period.
a
l
p
.
18
.
_
'
.-
.
PARTIAL LIST OF PERSONS CONTACTED
,
LiceDant
K. Carlson, Audit Team Leader
B. Ellison, Shift Manager
M. Gardzinski, Simulator Instructor
D. Herling, Daily Operating Shift Manager
J. Hill, Quality Manager -
J. Kempkes, Requalification Coordinator
M. Ledd, Training Issues Manager-
B. Mather, Shift Manager
T. Silverberg, General Superintendent Plant Operations
J. Sorensen, Plant Manager
O. Smith, Shift Manager
D. Westphal, Operations Training Superintendent
Nf1C
P. Krohn, Resident inspector
S. Ray, Senior Resident inspector
INSPECTION PROCEDURES USED
IP 71001," Licensed Operator Requalification Program Evaluation"
IP 71707, " Plant Operations"
ITEMS OPENED, CLOSED, AND DISCUSSED
Ooened
50-282/306-97019-01 VIO Inadequate procedure, SWI O-10, which was not OC reviewed
and which circumvented required EOP steps. Violation of 10 CFR Part 50, Appendix B, Criterion V," Instructions, Procedures, and
Drawings."
50-282/306-97019-02 VIO Licensee implemented SWis as the underlying procedure in lieu of
approved and OC reviewed procedures. Violation of Technical Specification 6.5, " Plant Operating Procedures. "
' 50-282/306-97019-03- IFl
Licensee's use of dual role SRO/STA
.
50-282/306-97019-04 URI Adequacy of procedure 1E-3 to perform concurrent cooldown and
l'
depressurization of the RCS and ability to meet the USAR time
limit of 30 minutes to accomplish the SGTR EOPs.
i-
19
i
. - -
.._ .
-
_
-
-
-
--
_
-
. --
1
]
. .
.
'
LIST OF DOCUMENTS REVIEWED
Prairie Island Updated Safety Analysis Report
- '
Technical Specification Administrative Section 6.1, " Organization," Revision 105
Technical Specification Table 6.1-1, " Minimum Shift Crew Composition," Revision 105
SWI O-1," Work Rules and Philosophy for Operation of Nuclear Plants," Revision 9
SWI O 2, " Shift Organization, Operation & Turnover," Revision 36
SWI O-10, " Operations Manual Usage, " Revision 29
SWI O-36, " Plant Security," Revision 2
SWI O-41, " Duties and Responsibilities of Fuel Handling Personnel," Revision 4
SWI O-43, * Operator Qualification Program," Revision 0
E-0, " Reactor Trip or Safety injection," Revision 17
E-3, " Steam Generator Tube Rupture," Revision 13
_
'
FR-S.1, " Response to Nuclear Power Generation /ATWS," Revision 8
TOP-01," Accident Analysis Topical DID," Revision 1
F-5, Appendix B, " Control Room Evacuation (Fire)," Revision 1?
Emergency Plan implementing Procedure, F3-1, "Onsite Fr.iergency Organization,"
Revision 14
NUREG-1275, " Operating Experience Feedback Report - Human Performance in
Operating Events," Volume 8, December 1992
- Information Notice (IN) 93-81, *lmplementation of Engineering Expertise on Shift,"
October 12,1993
SALP Report os. 50-282/306-96001,
Resident insp ctor observations and reports covering the time frame of 1996 to present.
o
Licensee event reports covering the time frame of 1996 to present.
Initial license examination Report Nos. 50-282/306-97306(OL).
Licensed operator requalification training Report No. 50-282/306-95013 (DRP).
Qualhy Assurance Audit Report, AG 1996-O 1, for Plant Operations Training
Generation Quality Services Status Report, Second Quarter 1997 (a data analysis and
trending report)
Quality Assurance Procedure 1 OAP 2.8, Revision 7 (requirements for audits)
Program Group Summary
Self-Assessment Operations Training (a self-assessment on the conduct of classroom
,
training and individualized instruction and trainee evaluation of Operations Training)
Training Procedure 1,11. " Training Effectiveness Self-Evaluation," Revision 1 dated
- '
September 20,1996
Self-Assessment Operations Training (a self-assessment on the analysis design and
development area of Operations Training)
Administrativo Work Instruction (AWI)- 5AWI 3.15.2, " Employee Observation
- Reporting," Revision 6
CHAMPS lssues Module, Revision 1 dated September 1997 (a new program for
assessment and tracking of internal and external issues / problems)
20
---_--_.-------------__-a-_a,,.u-,,,
_,,__,._,,
w-
w-
- - -
- - -
-
-
_ - -
-
. . -
.-
.
.
.
-.~_.. - -- -
- . .
.
- .
.
.
~.
A.'
v
i
- .p; w-
>
.
AOPI
Abnormal Operating Procedure
-
- Anticipated Transient Without Scram
. AWI
Administrative Work Instruction-
!
'
-CFR:
Code of Federal Regulations
Design Bases Document
DRS=
Division of Reactor Safety -
ED:
Emergency Director
EOP.
Emergency Operating Procedure
gpm
Gallons per Minute -
,
IP
inspection Procedure -
JPM.
LORT-
Licensed Operator Requalification Training
NRC
Nuclear Regulator Commission
--NRR
NRC Office of Nuclear Reactor Regulation
Northem States Power Company-
.OC
Operations Committee
7
. Public Document Room
Reactor Operator
Self Contained Breathing Apparatus
SGTR-
Steam Generator Tube Rupture
St.
Safety Injection
SM _
Shift Manager
Senior Reactor Operator
SWI
Section Work Instruction
Updated Safety Analysis Report
Violation
~
Work Control Center
21
- . . .-.
. -
. -
-
.
. . -
-.
,.
. ..
- _ . - - . . . - . - - . . - - -
- . - - . . -
..
. .
. ..-. -
.
I'
. .;
m
'
Attachment 1
-
.
' SIMULATION FACILITY REPORTI-
--
Facility Licensee:
Prairie island Units 1 and 2
- Facility Licensee Dockets Noi
50-282, 50-306
Operating Tests Administered:
September 29,1997 - October 3,1997
1
-.
This form is to be used only to report observations. These observations do not constitute audit -
~ or inspection findings and are not, without further verification and review, Indicative of
- noncompliance with 10 CFR 55.45(b). These observations do not affect NRC certification or- -
,
approval of the simulation facility other than to provide information that may be used in future
evaluations.: No licensee action is required in response to these observations.'
While conducting the simulator portion of the operating tests, the following items were observed
.
- (if none, so state):
IIEM
DESCRIPTION
4
' NONE OBSERVED
,
4
h
I
>
-
A
i
J
'
.m
a..)e
s.,.a
em,sa
,- - qs
e
i-