ML20199F572
| ML20199F572 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 01/23/1998 |
| From: | Grant G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Wadley M NORTHERN STATES POWER CO. |
| References | |
| 50-282-97-19, 50-306-97-19, NUDOCS 9802040045 | |
| Download: ML20199F572 (30) | |
See also: IR 05000282/1997019
Text
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Jr.nuary 23, 1998'
'
, . Mr. M. Wadley, Vice President
2
Nuclear Generation
Nor1 hem States Power Company
s 414 Nicollet Mall .
Minneapolis, MN 55401---
SUBJECT: ~ INADVERTENT RELEASE OF A DRAFT COPY OF PRAIRIE ISLAND L
INSPECTION REPORT NO. 50 282/97019(DRS); 50-306/97019(DRS)
l Dear'Mr. Wadleyi
- _ On December 9,1997, the NRC Resident Inspector Office at Prairie Island inadvertently provided - .
l your staff with an e-mail copy of Inspection Report No. 50-282/97019(DRS); 50-306/97019(DRS)
L before it had been approved for release. Although the draft version of the inspection report
provided to you at that time and the approved version that was issued on December 23,1997,
.were very similar, it is NRC policy not to release draft copies of NRC inspection reports to
licensees or the public. I am enclosing a copy of this draft report as a means for placing it in the
.
. Public Document Roor::(PDR).
-In order te assure that similar errors do not recur, I have lostructed the residents at Prairie Island
. and other Region 111 reactor sites to discontinue the practice of providing electronic versions of
= inspection reports to licensee personnel. . In the future, should your staff desire an expedited
- copy of an approved and issued inspection report, please call the regiori directly, and we will
, ~ either fax or express mail a copy of the report to you.
4
In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter and its
y enclosure will be placed in the NRC PDR.
We will gladly discuss any questions you have concoming this issue.
l
Sincerely,
..
L -/s/Geoffrey E. Grant ,
Geoffrey E. Grant, Director. '
_.
Division of Reactor Projects /
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Docket No. 50-282
F+ -
Docket No; 50-306
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Enclosure: Draft version of Inspection Report
, No. 50-282/97019(DRS); 50-306/97019(DRS) ,
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See Attached Distribution ~
- DOCUMENT NAME: G:\poin\ poi 97019.pdr
_ To receive a espy of this document, Indicate in the boa *C" = Copy without ottachment/enclosuse "E" = Copy with attachment / enclosure
- N* = No copy
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- cc w/ encl: Plant Manager, Prairie ' Island
' Stats Liaison Officer, State
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State Usison Officer, State >
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Tribal Council '
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Distribution: .
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PUBLIC lF.-01 w/enci , - Prairie Island PM,- NRR w/enci .,
SRI Prairie Island w/enci Rill Enf, Coordinator w/enci -
A. B, Beach w/enci Deputy RA w/ encl
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Mr. M. Wadley -
- - Vice President, Nuclear Generation -
i- Northem States Power Company
- 414 Nicollet Mall
Minneapolis, MN 55401 -
SUBJECT:- - NRC INSPECTION REPORTS 50-282/97019(DRS); 50 306/97019(DRS);
LICENSED OPERATOR REQUALIFICATION TRAINING PROGRAM
,
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EVALUATION AND NOTICE OF VIOLATION
p Dear Mr. Wadley:
L
On October 3' 1997, the NRC completed an on site inspection at your Prairie Island Nuclear
,
Generating Plant. The purpose of this inspection was to review the licensed operator
- - requalification program._ This initial inspection identified some issues which resulted in additional
.
inspection. The inspection findings were discussed with your staff on October 3 and December
2,1997, and during a teleconference on October 23,1997. The inspection findings were also
! discussed during a management meeting on November 25,1997. The enclosed report presents
the results.of the inspection.
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The inspectors concluded the licensed reactor operator and senior reactor operator
requalification programs were generally implemented in accordance with 10 CFR Part 55
,
requirements. Examination material prepared by your training stan, and operator evaluations
pedormed by your training staff, were satisfactory. '
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However, based on the results of this inspection, the NRC has determined that two violations of
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NRC requirements occurred. The first violations involved a procedure inadequacy that
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circumvented proper performance of the Reactor Trip and Anticipated Transient Without Scram-
4
emergency operating procedures. This violation is of concem because your staff failed to
understand the proper entry conditions for a fundamental plant operating procedure.- It is clear
. _
=
that your staff still did not understand these entry conditions until after ine November 25,1997
4- management meeting. The second violation involved a problem with your procedure change
4 management program and is of concem because appropriate management oversight of such
[ changes is an important element of the process.
~ These violations are cited in the enclosed Notice of Violation (Notice). The circumstances
, pertaining to the violations are described in detailin the enclosed report. Please note that you
'.
.are required to respond to these violations and should follow the instructions specified in the
-
enclosed Notice when preparing your response. The NRC will use your response, in part, to
determine whether further enforcement action is necessary to ensure compliance with regulatory
. requirements.
, In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter, its
. enclosures, and your response will be placed in the NRC Public Document Room (PDR).
l
) M. Wadley 2
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We will gladly discuss any questions you have conceming this inspection.
Sincerely,
John A. Grobe, Director
Divialon of Reactor Safety
Docket Nos. 50-282; 50 306
License Nos. OPR-42; DPR-60
Enclosures: 1. Notice of Violation
2. Inspection Reports 50-282/97019(DRS); 50-306/97019(DRS)
cc w/encis: Plant Manager, Prairie Island
State Liaison Officer, State
of Minnesota
State Liaison Officer, State
of Wisconsin
Tribal Council, Prairie Island
Dakota Community
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I Mr. M. Wadley .
..
VKe President, Nuclear Generation ,
! ~ Northem States Power Company
. 414 Nicollet Mall-
- Minneapolis, MN 55401 <
i SUBJECT: NRC INSPECTION REPORTS 50-282/9701558.3); 50-306/97019(DRS);
- LICENSED OPERATOR REQUALIFICATION TRAINING PROGRAM *
,
.. EVALUATION AND NOTICE OF VIOLATION
[ Dear Mr. Wadley;
i .
. , On October 3,1997, the NRC completed an on site inspection at your Prairie Island Nuclear
! Generating Plant.:The purpose of this inspection was to review the licensed operator
!
. requalification program. This initial inspection identified some issues which resulted in additional
- : inspection. The inspection findings were discussed with your staff on October 3 and December -
~ 2,1997, and during a teleconference on October 23,1997. The inspection findings were also
discussed during c management meeting on November 25,1997. The enclosed report presents
- the results of the inspection. ,
The inspectors concluded the licensed reactor operator and senior reactor operator
p roqualification programs were generally implemented in accordance with 10 CFR Part 55
,
requirements. Examination material prepared by your training staff, and operator evaluations
- performed by your training staff, were satisfactory.-
Hawever, based on the results of this inspution, the NRC has determined that tivo violations of
'
NRC requirements occurred. The first violations involved a procedure inadequacy that
circumvented proper performance of the Reactor Trip and Anticipated Transient Without Scram
- emergency operating procedures.cThis violation is of concem because your staff failed to
F
understand the proper entry conditions for a fundamental plant operating procedure, it is clear -
that your staff still did not understand these entry conditions until after the November 25,1997
i - management meeting. The second violation involved a problem with your procedure change
i management program and is of concem because appropriate management oversight of such .
,
changes is an important element of the process.
l These violations are cited in the enclosed Notice of Violation (Notice). The circumstances
pertaining to the violations are described in detail in the enclosed report. Please note that you
are required to respond to these violations and should follow the instructions specified in the
, enclosed Notice when preparing your response. The NRC will use your response, in part, to
determine whether further enforcement action is necessary to ensure compliance with regulatory
requirements.-
i'
. In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter, its .
.
enclosures, and your response will be placed in the NRC Public Document Room (PDR).
p
L M. Wadley 2
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We will gladly discuss any questions you have concoming this inspection.
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- Sincerely,
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John A. Grobe, Director
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g.j Division of Reactor Safety- _
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Docket Nos,50-282; 50 306
- Enclosures:- 1, Notice of Violation
2, inspection Reports 50 282/97019(DRS); 50-306/97019(DRS)
cc w/encis: Plant Manager, Prairie Island
- State Liaison Officer, State
of Minnesota ,
State Liaison Officer, State
- . of Wisconsin .
! Tribal Council, Prairie Island
-
Dakota Community i
- . Ql.fl0I5lE9.0 - . . .
1- Docket File w/encls Rlli PRR w/encls Rlli Enf. Coordinator w/encls -
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PUBLIC IE-42 w/encls SRI, Prairie Island w'!,ncis TSS w/encls !
, LPM, NRR w/encis . J. L Caldwell, Rill w/encls DOCDESK w/encls _!
- DRP w/encls A. B. Beach, Rlll w/encls CAA1 w/encls t
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DOCUMENT NAME: G:DRS\PRA97017.DRS
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OFFICE. Rlli l Rlli l Rlil l Rlli l
i~ NAME PPeterson:]p MLeach JMcCormick-B *
JGrobe
f DATE 12/ /97 12/ /97 12/ 197 12/ /97
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OFFICIAL RECORD COPY
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NOTICE OF VIOLATION
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Northem States Power Company ' Docket Nos. 50 282: 50-306
Prairie Island Nuclear Generating Plant License Nos. DRP-42; DPR-
'60
During an NRC inspection conducted from September 29 th. ugh October 3,1997, three
violations of NRC requirements were identified. In accordance w;th the " General Statement of -
Policy and Procedure for NRC Enforcement Actions," NUREG 1600, the violations are listed
.below:
1. Code of Federal Regulation Title 10 Part 50, Appendix B, Criterion V, " Instructions,
Procedures, and Drawings," states, in part, that activities affecting quality be prescribed
'
by documented instructions and procedures of a type appropriate to the circumstances .
and be accomplished in accordance with these instructions or procedures.
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Contrary to the above, on July 24,1996, the licensee implemented procedure
P SWI O-10, " Operations Manual Usage," Revision 28, Section 6.11.9, which contained
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without a scram) condition not to enter emergency operating procedures (EOP) unless a
i valid protection system set point was reached and the reactor trip breakers were not open
i and cannot be opened manually. This instruction was in direct conflict and circumvented
, the requiremats of EOPs 1(2)E-0, " Reactor Trip or Safety injection," Revision 17,
l Section A and Step 1, and 1(2)FR S.1, " Response to Nuclear Power Generation /ATWS,"
Revision 8, Step 2.
This is a Severity Level IV violation (Supplement 1).
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2. Technical Specification 6.5 required that detailed written procedures, including the
applicable checkoff lists and instructions, covering areas listed be prepared and followed. ;
The specification further required that the procedures and chan'ges thereto be reviewed
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by the Cperations Committee (OC). Areas listed under Plant Operations included the
following:-- (1) integrated and system procedures for normal startup, operation and
shutdown of the reactor and all systems and components involving nuclear safety of the 1'
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 leaks and abnormal reactivity changes and including folloiv-up
actions requirert after plant protective system actions have initiated; (4) implementing
procedures of the fire protection program.
Contrary to the above, the inspectors identified that as of October 9,1997, the licensee
- had established a series of Operating Department Section Work Instructions (SWis),
l < covering areas listed in Technical Specification 6.5, which were not reviewed by the OC.
The Section Work Instructions included the following:
SWI O-1, " Work Rules and Philosophy for Operation of Nuclear Plants," Revision
9, dated July 17,1997, Sections 6.2, 8.6, and 6.7
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- Notice of Violation 2
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- SWI O-10, " Operations Manual Usage," Revision 29, dated March 24,1997,
j Sections 6.4 and 6.7.5.
L SWl'O-41, " Duties and Responsibilities of Fuel Handling Personnel," Revision 4,
dated July 17,1997, Sections 6.1,7 and 6.2.
This is'c Severity Level IV violation (Supplement I).
r
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Pursuant to the provisions of 10 CFR 2.201, Northem States Power Company is hereby required
to submit a written statemert or explanation to the U.S. Nuclear Regulatory Commission, ATTN:
Document Control Desk, Washington, D.C.,20555_with a copy to the Regional Administrator,
Region lil, and a copy to the NRC Resident inspector at the facility that is the subject of this
- Notice, with5 30 days of the date of the letter transmitting this Notice of Violation (Notice). This
E reply should be clearly marked as a " Reply to a Notice of Violation" and should include for each
violation: (1) the reason for the violation, or, if contested, the basis for disputing the violation, (2)
l the corrective steps that have been taken and the results achieved, (3) the corrective steps that
will be taken to avoid further violations, and (4) the date when full compliance will be achieved.
. Your response may reference or include previous docketed correspondence, if the -
- - correspondence adequately addresses the required response. If an adequate reply is not
received within the time specified in this Notice, an order or a Demand for Information may be
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Issued as to why the license should not be modified, suspended, or revoked, or why such other
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action as may be proper should not be taken. - Where good cause is shown, consideration will be
- ;given to extending the response time.
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Because your response will be placed in the NRC Pubic Document Room (PDR), to the extent
!- possible,t should not include any personal privacy, proprietary, or safeguards information so that
,. it can be placed in the PDR without redactica; if personal privacy or proprietary information is
L 'necessary to provide an acceptable response, then please provide a bracketed copy of your
, - response that identifies the information that should be protected and a redacted copy of your- ,
-
response that deletes such information if you request withholding of such material, you mg.31
specifically identify the portions of your response that you seek to have withheld and provide in
t detail the bases for your claim of withholding (e.g., explain why the disclosure of information will
create an unwarranted invasion of personal privacy or provide the information required by 10
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CFR 2.790(b) to support a request for withholding confidential commercial or financial
7. Information); if safeguards information is necessary to provide an acceptable response, please
- - provide the level of protection' described in 10 CFR 73.21.
.
DSted at Lisle, Illinois,
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this day of December 1997
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U.S. NUCLEAR REGULATORY COMMISSION
REGION lli
, Docket Nos: 50 282;50-306
Report Nos: 50-282/97019(DRS); 50-306/97019(DRS)
,
Licensee: Northem States Power Company
Facility: Prairie Island Nuclear Generating Plant
i Location: 1717 Wakonade Dr. East
,
Welch, MN 55089
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Dates: September 29 - October 03, and October 21-23,1997
Inspector: H. Peterson, Reactor Engineer, Lead Inspector .
R. Bailey, Reactor Engineer
S. Ray, Senior Resident inspector, Prairie Island
P. Krohn, Resident inspector, Prairie Island
Approved by: M. Leach, Chief, Operator Licensing Branch
Division of Reactor Safety
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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 examination
material; observation and evaluation of operator performance and 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 (IP) 71001 and 71707.
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The licensed operator requalification programs were implemented in accordance with 10 CFR
- Part 55 requirements.
e All portions of the annual requalification examination were judged to be effective tools for
determining operator weaknesses (Sections 05.2,05.3 ).
,
e Control Room operators demonstrated an appropriate level of attentiveness to th9
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).
e 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:
o 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 access to vital control areas
within the control room (Section 01.1).
g e The licensee implemented an inadequate procedure which circumvented required EOP
steps. This was considered a violation 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 O3.1).
- The licensee was implementing SWis as the underlying procedure in lieu of approved and ;
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OC reviewed procedures. This was considered a violation of Technical Specification 6.5,
" Plant Operating Procedures, " (Section O3.1).
- The licensee's use of the " dual-role" SRO/STA could potentially impair crew performance
(Section 04.1).
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DRAEI 1
- The licensee continued to demonstrate difficulties in procedure use (Sections 05.1,05.4,
05,7).
- . The licensee's instruction for Fire Brigade personnel on respirator fit qualification was
clear, but no such guidance or instruction was in place for all other licansed operators
(Section 08.1),
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Reports Details
1. Operations
01 Conduct of Operations
-. 01.1 Control Room Observationg
af Inspection Scope (71707) -
The inspectors observed routine control room activities during full power operaticns,
pwformed a dual unit panel walk down, reviewed control room logs, and questioned
operators about plant and equipment status,
b. Observations and l'indinas i
in general, the control room operators conducted themselves in a professional manner
and acre attentive to their respective panelindications. However, access control to the
main control room area and verbal communication practices were not always consistent
with management expectations and guidelines. Conversation contained informal and ;
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incomplete three-way communication phrases. Access to the main contro! room area by
non-licensed individuals occurred repeatedly without challenge. On two occasions, one
individual 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
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. Howevor, the inspectors were concemed that operator performance in
the area of verbal communications was not always consistent with management
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expectation and guidelines, and that a lack of formal controls to restrict personnel access
l to vital control areas existed
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03 Operations Procedures and Documentation
03.1 Section Work Instructions Vice Emeroency Operatina Procedures
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s. Inspection Scope (71707. 71001)
The inspectors observed the licensee's evaluation of three operating groups during the
- requalification examination simulator scenarios. Based on these observations, the
inspectors identified inappropriate use of sechon work instructions (SWI) in lieu of
properly following an Emergency Operating Procedure (EOP). Also, the inspectors
identified an inadequate interpretation of an abnormal operating procedure (AOP) step.
The inspectors reviewed licensee procedures and instructions,10 CFR Part 50, EOP
.
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DRMI
eme,gency response guidelines, and technical specifications to ascertain the safety and
regulatory impact of SWis,
b. Observations and Findinas
On Svptember 30,1997, during one simulator scenario, a seal failure on a reactor coolant
pump required entry into abnormal operating procedure 103 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 failed 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 condition was identified. The reactor
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 resctor trip. This was not identified
by the crew for about 40 seconds.
The inspectors questioned the licensee on this incorrect practice in responding 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 instruction
was Revision 28 dated July 24,1996). Procedure SWI O-10, Section 6,11.9,
" Anticipated Transient Withcut Scram (ATWS)," contained instructions that when a
conservative decision was made to initiate a rcactor trip, based on deteriorating plant
conditions, but before a reactor trip setpoint was reached, then the reactur should be
tripped manually using one of the two trip switches. If this was unsuccessful, then the 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 manually, 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 transient on the plant, and that it was desirable to avoid this event
unless absolutely necessary. The licensee, therefore, was interpreting that an ATWS
condition did not exist, even if a manual attempt to trip the reactor failed using
the trip switches, unless an automatic trip safety signai was present and if the trip
, breakers could not be opened manually.
The Code of Federal Regulations Title 10, Part 50, Appendix B, Criterion V," Instructions,
Procedures, and Drawings," stN., in part, that cctivities affecting quality be prescribed
by documented instructic" m ocedures of a type appropriate to the circumstances
and be accomplished in acc .. m w th these instructions or procedures.
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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 prior to entering eny
EOPs and not to consider the event cs an ATWS condition, was inadequate and directly
conflicted with approved EOPs. The EOP 1(2)E-0, " Reactor Trip or Safety injectbn, "
Revision 17, Section A and Step 1, required entry into the procedure following a manual
or automatic actuation of a reactor trip and entry inic FR-S.1 on failure of a reactor trip,
respectively. The EOP 1(2) FR-S.1, " Response to Nuclear Power Generation /ATWS," a#
Revision 8, Step 2 required an immediate turbine trip for any ATW6 condition. Tho action
to manually open reactor trip breakers was a follow up action in Step 5 of FR-S.1.
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, Criterion 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 and followed.
The specification further required that the procedures and changes thereto be reviewed i
by the Operations Committee (OC). Areas listed under Plant Operations includerd the
following: (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
ala ms, primary system leaks and abnormal reactivity changes and including follow-up
actions required after plant protective system actions have initiated; (4) implementing
,- procedures of the fire protection program.
The inspectors identified that (1) SWis were not part of the list cf procedures described in
the administrative section of the technical specification, and theref .re were not routinely
reviewed by the 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 handling, 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 that the fire
brigade and fire brigade support personnel be clean-shaven at the start of the
shift. This instruction was provided to ensure personnel were ab!s to wear
breathing apparatus. The instruction was not contained in any OC reviewed
procedure and was considered an implementing procedure of the fire protection
program which required OC review.
'
e SWI O-1, " Work Rules and Philosophy for Operation of Nc. lear Plants," Revision
9, Section 6.6, contained specific instructions for monitoring plant parameters
after a reactivity manipulation. The instructions included a list of instrume'tation
to observe. The instructions were similar to but much more detailed than
instructions in OC reviewed procedures for plant operations. In fact, one
1perating 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 was considered a procedure for normal operation
of the reactor which required OC review.
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e 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 instructicns included a list of instrumentation to use and time limits for
inadvertent operations above 100 pe, cent of rated thermal power. The
instructions were similar to but much more detailed than instructions in OC
reviewed procedures for plant operations. The SWI was considered a procedure
- for normal operation of ine reactor which required OC review,
o S$NI 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 operators were expected to initiate isolation of the
moisture sep.arator - 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 operater 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,
o SWI O-10, " Operations Manual Usage," Revision 29, Section 6,7.5, contained
instructions requiring that four specific checklistr for establishing containment
integrity be ccmpleted twice whenever they were performed to satisfy a
procedural rnquiremer.t. That instruction, to perform the checklists twice, was nct
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 which required OC review,
e SWI O-41, " Duties and Responsit"lities 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 th9 spent fuel handling
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.
e SWI O-41, " Duties and Responsibilities cf 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 fue'
handling procedures reviewed by the OC. The SW1 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 violation of Technical
Specification 6.5, " Plant Operating Procedures, " (VIO 50-282/306-97019-02(DRS)).
c. Conclusions
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
7
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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 rt. viewed by the Operations Committee.
04 Operator Knowledge and Performance
04.1 Sbift Manaaer As Shift Technical Advisor
a. inspection Scope (71707. 71001)
The inspectors reviewed the I;censee and industry documents associated with licensed
shift personnel duties and responsibilities, sh!!! organization, onsite emergency
organization, and operating 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 Tatte S.1-1. * Minimum Shift Crew Compositwn, " Revision
105
e Section Work Instru ; tion (SWI), SWI O-2, " Shift Organization, Operation &
Tumover, " Rev;sion 36
- Emergency Plan implementing Procedure, F3-1,"Onsite Emergency Organization,
" Revision 14
- NUREG-1275, " Operating Experience Feedback Report - Human Performance in
Operating Events, " Volume 8, December 1992
z
e information Notice (IN) 93-81," implementation of Engineering Expertise on Shift,"
October 12,1993
,
b. _ Observations and Findinos
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-rcie" STA was allowed per an October 28,
1985, Federal Register notice 50 FR 2. 621, "NRC Policy Statement on Engineering
Expertise on Shift."
Within th3 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 situr4ons.
Also, in accordance with emergency plan implementing procedure FS-1, 'ne SM has the
'
initial responsibility to assume the duties of the Emergency Director (ED) during an
emergency event. The Emergency Director's respons;bilities were very significant and
included the following: (1) coordinate response of the piant ons;te emergency
organization, (2) emergency classification and notification of offsite authorities, (3)
authorize offsite Protective Action Recommendaticns, (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
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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 during abnormal and accident conditions and
recommend action. The STAS were to have a bachelor's degree in er.gineering or
equivalent to render engineering technical advise during an accident. Furthermore, the
STAS for We9tinghouse 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 g3-81 and NUREG-1275) pertaining to operating
experience concoming multiple-role SRO/ STAS, the inspectors identified that problems
havo occurred at other facilities which resulted in overburdening the SRO/STA while
fulfilling duties involving EOP reading, event ciassification, fire protection concems, and
implementation of the emergency p!an; The licensee stated that the SMs role as
Emergency Director was supported by the SRO from the unai-L.ied unit. This individual
- prepared the necessary paperwork for the SMs review and approval. The viability of this -
'd
operating structure will be reviewed during a future inspection (IFl 50 282/306 g7019-
03(DRS)).
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c. .Qonclusioq
The inspectors concluded that the licensee's use of the ' dual-role" STA could potentially
impair crew per*ormance and this will be reviewed during a future inspectien.
05 _ Operator Training and Qualification
05.1 Operatina Historv .
a. Insoection Scoes (71001)
Th9 inspectors reviewed the following to assess the licensed operator requalification
training program's effectiveness regarding operator performance:
e SALP Report Nos. 50 282/306-96001,
e Resident inspector observations and reports covering the time frame of 1996 to
present,
e. Licensee event reports covering the time frame of 1996 to present. 1
e Initial license examination Report Nos. 50-282/306-97306(OL), J
e- - Licensed operator requalification training Report Nos. 50-282/306-95013 (DRP).
b, Observations and Findinas
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 Geensee's action itsms centered around
procedure use and cc npliance, and overall proceours development. The inspectors,
however, identified continued problems concoming proper use of procedures. In
particular, the licensee's over reliance 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 problem. The licensee had recently begun
corrective actions for these problems but these were not sufficiently implemented to allow
for an objectivo evaluation.
<-
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05.2 Reaualification Examinations
1
05.2.1 Examination Material ?
a. Inspection ScoJLe (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
questions, dynamic scenarios, and job performance measures (JPM) with previously
administered examinations.
b. Observations and Findinos
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
fok 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 this 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 for
successful completion. However, program deficiencies were noted during the
oerformance of JPMs: (1) more than one JPM had incomplete cues which required the
j evaluator to improvise; (2) one JPM contained an inaccurate performance standard that
required evaluatorjudgement; and (3) one JPM contained performance tasks that ware
not consistent with the procedure in use which required evaluatorjudgement. The
inspectors were concemed that the review and validation process had not identified these
deficiencies even with two levels of technical review being performed.
The written questions were operationally oriented and contained an appropriate level of
difficulty. A majority of the open reference questions were of higher cognitive knowledge
level. The static examination questior s 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 significantly modified from the previous
examination cycle.
c. Conclusions
The inspectors concluded that the requalification examination material contained the
necessary quantitative and qualitative attributes to provide an effective evaluation of
operator skills. However, simulator scenarios contained a disproportionate number of
ATWS scenarios and some JPMs did not provide appropriate cues r d performance
standards.
05.2.2 incorporation of Current industry Events
a. Jrtsoection Scope ppt
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' The inspectors reviewed the licensee's program to assess and incorporate current !
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industry events applicable to the facility into training and testing. Particular attention was i
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, placed on recent industry concems or. 'ho capability of timely performing the emergency
L operating procedure for a steam generator tube rupture (SGTR) based on the Prairie
n Island Updated Safety Analysis Report (USAR), Section 14 Revision 13, time criterion. .
L . The inspectors also reviewed Emergency Operating Procedure 1E-3, " Unit 1, Steam ,
Generator Tube Rupture," Rev.13.
l
b. Observations and Findinas-
f- -
, The inspectors identified that the licensee, on July 2g,1997, initiated a non-conformance
report which rioted that isolation of SGTR by operators may exceed the 30 minutes USAR
- - assumption. The non-confontance 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. The licensee conducted timed simulator
( SGTR training during the week of July 7,1997. Four different operating groups, were
L
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given evaluation scenarios that included a SGTR with loss of offsite power. The four
crews performed the SGTR mitigating actions to terminate the primary to secondary leak
- by finally terminating safety injection in 34, 37,40, and 36 minutes, respectively.
!
Following the evaluation, the licensee concluded that the scenarios were not pro-
l evaluated to correspond to licensing basis or design basis requirements, but also
'
concluded that it was a rough approximation of possible response times. The licensee
[ noted that in all four scenarios, the steam generator (SG) narrow range levels remained '
. belovt g5%, but the times exceeded the 30 minutes assumption in the USAR. The
l . licensee determined that the test was not valid, such that no design basis was exceeded.
L Additionally, the licensee determined that the USAR 30 minutes time limit was a
l conservative time estimate, that as long a radioactive release was minimized the
- . requirements of the USAR were met.
l' The licensee informed the inspectors that discussions were held with Westinghouse to
, resolve the issue of the 30 minutes time limit, and that continued assessment was being
p made to develop the appropriate simulation scenario to test the USAR time limit.
!
j- Following the on site inspection the licensee informed the inspectors that additional
'~
testing with a rupture size of about 600 gpm had shown that four of five cc:ws could meet
- the 30 minutes specified in the USAR. However, during the performance of these
L scenarios the crew had performed the reactor coolant system (RCS) cooldown and
'
statement, "If SG overfill is an immediate concem, THEN cooldown and depressurization
steps may be performed concurrently with shift supervisor approval." The procedure
does not include steps for performing these activities concurrently even though the
?
background document states "Although concurrent RCS cooldown and depressurization
l 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) termination. Furthermore,
j concurrent cooldown and depressurization would also require more precise pressure
L 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 ccncurrent RCS cocidown and depressurization." The
, failure to provide specific instructions in EOP 1E-3 and the adequacy of the procedure to
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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 -
The inspectors concluded that the licensee was relying on concurrent actions to
cooldown and depressurize the RCS in order to meet the time limits specified in the
USAR. The inspectors were concemed over the adequacy of the procedural instructions
- to perform thebe tasks concurrently and this will be evaluated further.
05.3 Reaualification Examination Administration Practices
.. a. -Inspection Scope (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 thirteen 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 critiques,
b. Observations and Findinas -
The licensee's evaluation team Identified no unsailsfactory crew performance. However,
two operators were identified as having demonstrated poor performance during JPMs and
- required follow up training. The evaluators appropriately documented the operators'
_ performana as unsatisfactory on 1 of the 5 JPYs administered (See Section 05.5 for a
discussion of the remediation process).
The evaluators performed the examination administration in a professional manner and
property documented operator performance deficiencies. No evaluator miscuing or
prompting was identified.
Appropriate security measures were taken throughout the examination process.
Individual operators were sequestered and separated into test groups durir g each portion
of the examination process. No exam compromise was identified.
No new simulator fidelity issues were identified during the exam observation (See
= Enclosure 2, Simulation Facility Report). .
c. .Qonclusions
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 Proaram Feedback System
a. Inspe'ction Scope (71001)
,-
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- The inspectors reviewed the following documen;s to assess the licensee's training -
f' _ program feedback system effectiveness:
-. e Quality Assurance Audit Report, AG 1996-O-1, for Plant Operations Training
e Generation Quality Services Status Report, Second Quarter 1997 (a data analysis
and trending report)
e Quality Assurance Procedure 1 QAP 2.8, Revision 7 (requirements for audits)
e Program Group Summary _
-t
e Self-Assessment Operations Training (a self-assessment on the conduct of-
- classroom training and individ'Jalized instruction and trainee evaluation of -
Operations Training)
e Training Procedure 1.11, ' Training Effectiveness Self-Evaluation," Revision 1 ' i
- dated September 20,19g6 =
o- Se'4-Assessment Operations Training (a self-assessment on the analysis design
and development area of Operations Training)
e . Administrative Work Instruction (AWI)- 5AWI 3.15.2, "E nployee Observation
Reporting," Revision 6 -
e CHAMPS lssues Module, Revision 1 dated September 1997 (a new program for
assessment and tracking of intomal and extemal issues / problems)
b. Observationt and Findinas
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 performed periodic audits of the Operations and Training programs.
Subsequently, the Programs Group gathered, evaluated, and assigned priorities for the
- results of all the self-evaluations. including those conducted by the Nuciear Quality
Assurance group. One continuing theme identified through the Nuclear Quality
Assurance group's audit was procedure weaknesses, including improper procedure
usage and control.
The licensee's se'f-assessment program appeamd 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 ~ ' the examination bank when procedure
changes or modifications were implementeo by the plant. During this inspection, the
licei.;ee was updating its job performance measures (JPM) examination bank.'
c. Conclusions .
The inspectors determined that the feedback process was satisfactorily implemented.
05.6 RemedialTrainina Proaram
'
a. Inspection 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 performance.
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This review included an interview with seleMed personnel involved with the remedial
training process.-
b. Observations and Finc8inas
During previous evaluations, the licensee had identified a number of unsatisfactory
performances on both the written and JPM portions of the examination process. The
inspectors determined that selector' remedial training plans had incorporated a -
comprehensive retraining and evaluation process, and were consistent with the licensee's
assessment of operator's poor perforn,ance. The licensee acknowledged that the poor
written examination perfermance had been attributed to a recent revision in exam
_ question difficulty which made each one more operationally discriminating.
i 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 properiy -
identified and correded 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 Operator License Conditions
a .- - Inspection Scope (71001)
The inspectors reviewed the licensee's medical and operator qualification programs and
selected records 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.
b. Observations and Findinas 4
r
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 violation of regulatory requirements was identified.
On January 24,1997, the licensee implementcd a new procedure, SWI O-43, that gave
guidance for maintaining operator licenses 'a an active status. This issue was originally
identified on October 24,1995, by the %nsee's Quality Services group duilng 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
limnss in an active status. However, on August 28,1996, the licensee became aware of
a ther licensee performing the similar practice and found that WCC duty was not
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acceptable for credit towards mahtaining active license status. Subsequently, the issue
, was identified to the NRC in inFpection Report 50-282/306-96008, Section 05.1. This
item was later closed in Inspwtica Report 50-282/306-97002, Section 08.3, based on a
letter submitted by the licens te to the NRC stating that they had discontinued the practica
of crediting duty in the WCC 1s meeting the criterir, 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 cf 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 EquipmeiJ. and
Reactor Operator.
c. Conclusions
The 'nspectors concluded that the operator's license conditions were in conformance with
program guidance and regult. tory requirements stated in 10 CFR Part 55.53 and 10 CFR
Pcrt 55.21.
.
05.7 Follow-uo of Previously identified Weaknesses
a. Inspection Scope
The inspectors reviewed the identified weaknesses from the last Licensed Operator
Requalification Training (LORT) program inspection (NRC Inspection Nport 50-282/306-
95013) to ascertain the licensee's actions to resolve any weaknesses.
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b. Observations and Findinas
There were four weaknesses identified in the last LORT inspection report. One
weakness was associated with no additional training for individuals or crews that
demonstrated significant weaknerses, but where the overall performance was evaluated
as satisfactory. The licensee initiated " follow up" training for those individuals and crews,
Two other waaknesses were associated with operator performance in procedure
implementation and communications. The licensee was aware of these concems and
wat actively pursuing the issues; however, procedure problems continued to be a
concem. Also, communications 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 Respirator Fit Prooram
a. IrLpection Scope (71707)
The inspectors reviewed the licensee's plant safety procedure F5, Appendix B, " Control
Room Evacuation (Fire)," Revision 17, for operator actions required during an evacuation
'
of the control room / relay room area. Expected operator actions 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 Mift 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 that some
of the licensed senior reactor operators (SRO) had facial hair (beard) of such length that
a proper mask fit would not be poss;ble when donning and using the SCBA.
The inspectors were concerned 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 fur control room operators except for those (RO licensed only) assigned to the
Fire Brigade. Also, the inspectors identified and informed the licensee of a concem that
an on shift control room operator (RO !! censed) 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 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.
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Additionally, the inspectors questioned the availability of special corrective lens for
- respirator masks and were informed that access to speciallens existed. The inspectors
verified that special corrective lens did exist and were available for individual use when
l 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. Wnaaement Meetinas
X1 Exit Meetina Summary
The inspoctors presented the inspection r3sults to members of licensee management on October
3, and December 2,1997, and during a teleconference on October 23,1997. The licensee
p acknowledged the findings presented.
The inspectors asked the licensee whetner any materials held by the inspectors following the
- inspection could be considered proprietary No proprietary information was identified.
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PARYlAL LIST OF PERSONS CONTACTED
4
Licensee
K. Carlson, Audit Team Leader
B. Ellison, Shift Manager
M. Gardzinski, Simulator Instructor
D. Hening, Daily Operating Shift Manager
J. Hill, Quality Manager
J. Kempkes, Requalification Coordinator
M. Ladd, TraininD issues Manager
B. Mather, Shift Manager
T. T V herg, General Superintendent Plant Op3 rations
J. Octusen, Plant Manager
D. Smith, Snift !4 nager -
D. Westphal, Operations Training Superintendent
tiRG
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
Opened -
50-282/306 97019-01 VIO Inadequate procedure, SWI O ;d, which was not OC reviewed and
whk.h circumvented required EOP steps. Violation of 10 CFR Part
50, Appendix B, Criterion V," Instructions, Procedures, and
Drawings."
56 232/?O6-97019-02 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 rr SRO/STA
50 282/306-97019-04 URI Adequacy of r% IE-3 to perform concurrer.t cooldown and
depressurizwn e RCS and ability to meet the USAR time
limit of 30 minute 5 accomplish the SGTR EOPs.
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LIST OF DOCUMENTS REVIEWED
e Prairie Island Updated Safety Analysis Repor' ,
e Technical Specification Administrative Section 6.1, "Or, :nization," Revision 105
o Technical Specification Table 6.1 1, * Minimum Shift Crew Composition," Revision 105
o SWI O 1," Work Rules and Philosophy for Operation of Nuclear Plants," Revision 9
e SWI O 2, " Shift Organization, Operation & Tumover," Revision 36
e SWI O 10, * Operations Manual Usage, " Revision 29
e SWI O-36, * Plant Security," Revision 2
e SWI O-41, *Dutlas and Responsibilities of Fuel Handling Perso6nel," Revision 4
e SWI O-43, ' Operator Qualification Program," Revision 0
o E-0, " Reactor Trip or Safety injection," Revision 17
e E 3, " Steam Generator Tube Rupture," Revision 13
e FR S,1, " Response to Nuclear Power Generation /ATWS," Revision 8
e TOP-01, " Accident Analysis Topical DID," Revision 1
e F 5, Appendix B,' Control Room Evacuation (Fire)," Revision 17
e Emergency Plan implementing Procedure, F31,"Onsite Emergency Organization,"
Revision 14
e NUREG 1275," Operating Experience Feedback Report Human Performance in
Operating Events," Volume 8 December 1992
e Informatiot Notice (IN) 93 81, *lmplementation of Engineering Expertise on Shift,"
October 12,1993
e SALP Repor1 Nos. FO-282/306 96001.
- Resident inspector observations and reports covering the time frame of 1996 to present,
e Licensee event reports covering the time frame of 1996 to present.
e Initial license examinat!on Report Nos. 50-282/306 97306(OL).
- Licensed operator requalification training Report No. 50 282/306-95013 (DRP),
o Quality Assurance Audit Report, AG 1996-01, for Plant Operatloas Training
e Generation Quality Services Status Report, Second Quarter 1997 (a data analy11s and
trending report)
e Quality Assurance Procedure 1 QAP 2.8, Revision 7 (requirements for audits)
e Program Group Summary
e Self Assessment Operations Training (a self assessment on the conduct of classroom
training and individualized instruction and trainee evaluation of Operations Training)
e 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 Operabans Training)
e Administrative Work Instruction (AWI)- SAWI 3.15.2, " Employee Observation Reporting,"
Revision 6
e CHAMPS lssues Module, Revision 1 dated September 1997 (a new program for
assessment and tracking of intemal and extemalissues/ problems)
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20
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e,\
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3DN
usT oF ACRONYMS USED
AOP Abnormal Operating Procedure
ATWS Anticipated Transient Without Scram
AWI Administrative Work Instruction
CFR Code of Federal Regulations
DBD Design Bases Document
DRS - Division of Reactor Safety -
ED- Emergency Director
EOP Emergency Operating Procedure
opm Gallons per Minute
IP inspection Procedure
LORT Licensed Operator Requalification Training
NRC Nuclear Regulator Commission
NRR NRC Office of Nuclear Reactor Regulation
NSP Northen; States Power Company
OC Operations Committee .
PDR Public Document Room
RO Reactor Operator
SCBA Self Contained Breathing Apparatus
SGTR . Steam Generator Tube Rupture
SI Safety injection
SM Shift Manager
SRO- Senior Reactor Operator
STA Shift Technical Advisor
SWI Section Work Instruction
USAR Updated Safety Analysis Report
VIO Violation
WCC Work Control Center
21
5h
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l. .
Attachment 1
SIMULATION FACILITY REPORT
Facility Licensee: Prairie Island Units 1 and 2
.
Facility Licensee Dockets No: 50 282, 50-306
Operating Tests Administered: September 29,1997 October 3,1997
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):
lIEM DESCRIPTION
NONE OBSERVED
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m