IR 05000155/1990008
| ML20043B476 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 05/23/1990 |
| From: | Hasse R, Nejfelt G, Phillips M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20043B472 | List: |
| References | |
| 50-155-90-08, 50-155-90-8, NUDOCS 9005300078 | |
| Download: ML20043B476 (17) | |
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U.S. NUCLEAR REGULATORY COMMISSION i
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REGI9N III
~ Report No.- 50-155/90008 Docket.No. 50-155.
License No. DPR-6
' Licen see: Consumers Power Company 1945 West Parnall Road Jackson, MI 49201 k'
~ Facility Name:
Big Rock Point Plait Inspection Conducted At: Charlevoix, MI 49720 Inspection Conducted: ' April 17-26, 1990
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Inspectors:
C -13i D R.'Hasse, RIII
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Date Team Leader
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0523-90'
CNejfelpIII Date G. Bryan, Consultant (COMEX)
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. Wilford C nsultant-(SAIC)
Approved By:
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M. P. Phillips, Chief
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Operational Programs Section
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Inspection Summary I
Inspection on April 17 - 26, 1990 (Report No. 50-155/90008(DRS)).
Areas-Inspected: Special announced safety inspection to verify that the Big i
Rock Point Emergency operating procedures (EOPs) were technically correct and-l usable. 'This inspection was conducted in accordance with TI 2515/92 (SIMS
No.-HF-4.1).
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Results: No violations were identified.
The licensee showed strenoth in the f
approach to developing their symptom based E0Ps. Weaknesses were identified in the engineering assessments in the E0P basis document and the E0P verification effort.
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9005300078 900524 PDR ADOCK 05000155 Q
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REPORT DETAILS 1.
Persons Contacted'
Consumers Power Company (CPCO)
W. Beckman, Plant Manager L. Monshor, QA Superintendent G. Boss, Acting Operations Superintendent L. Darrah, Shift Supervisor
.D. Lacroix, Training Administrator R. Alexander Technical Engineering G. Pecitjean,' Supervisory Engineer P. Donnelly, Nuclear Assurance Administrator USNRC E. Plettner, SRI G. Wright, Chief, Operations Branch, R111 All personnel listed above attended the exit interview conducted on April 26, 1990. Other licensee personnel were contacted / interviewed during the inspection.
2.
Emergency Operating Procedures a.
Background Emergency Operating Procedures (E0Ps) have undergone significant changes due to the 1979 accident at the Three Mile Island (TMI)
facility. The post-TM1 procedures are symptom-oriented rather than event-oriented.
Symptem oriented E0Ps provide the operator guidance on how to verify the adequacy of critical safety functions and how to restore and maintain these functions when'they are degraded.
Symptom-oriented E0Ps are written in a manner that the operator need not diagnose an event to maintain the plant in a safe shutdown condition for all accidents that are within the scope of the E0Ps.
The purpose of this inspection was to verify that the Big Rock Point (BRP)E0Psaretechnicallycorrect;preparedinaccorIancewiththe writer's guide; that their specified actions can be accomplished using existing equipment controls, and instrumentation; and that the available procedures have the usability necessary to provide the operator with an effective operating tool.
b.
Inspection Methodology Symptom based E0Ps are derived from a set of plant specific Technical Guidelines (PSTG). The PSTG provides the fundamental
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' stepwise process for mitigating accidents. 'The PSTG'is supported by a Technical basis document which provides the engineering
- f rationale for the approach used. Most licensees use the generic technical guidelines (GTG) and basis document prepared by the
respective owners group as the' basis for their PSTG.
It is then necessary to document only the engineering rationale for deviations
=from the GTG. The licensee concluded that this approach was not appropriate.for BRP because of numerous differences between BRP and the reference plant. The licensee generated their own Technical guidelines and basis document based on the BWR Owners Group GTG (Rev. 2), the BRP plant specific PRA,'and existing emergency-procedures.
The inspection consisted of'a detailed review of the PSTG and basis document and the E0Ps. These documents were also compared with the current revision (Revision.4) of-the GTG'to determine if the GTG contained any mitigative strategy. applicable to BRP:that was not addressed in the BRP E0Ps.
Since BRP lacked on-site simulation capability, the usability of the E0Ps and operator knowledge were determined by table top walkthroughs with.
an operating crew utilizing four scenarios prepared by the inspection team..A human' factors review, plant walkdowns, and staff interviews
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were also conducted. A detailed listing of these activities is
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c.
Inspection Results The inspectors concluded that the BRP E0Ps were adequate to mitigate
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accidents within their scope and could be implemented by the plant-
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staff.
Some concerns were-identified and are discussed below.
Detailed comments are given in Appendix B.
i (1) Desktop Review I
The concerns identified during this phase of the inspection are discussed below by document type:
(a) Basis Document The basis document was considered to be generally adequate; however, the need for some improvements was identified.
-The engineering basis or rationale was not always adequately presented for action or setpoints.
For example, the low vessel leve1~setpoint for entry into E0P-1, " Primary System Control" was selected as - 17" in the steam drum.
The stated basis for this selection was that this level was below the level reached during normal plant transients, that core l
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cooling may be jeopardized below this: level if
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additional makeup' was not providedL and it. was-N
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coincident with RDS timer initiation which would be'
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rationale was presented to verify that1the:setpointt readily recognized by the operator. -No> engineering-m J
was sufficiently conservative to? assure timely and
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adequate level recovery and control..-Another setpoint lacking-documented engineering rationale was the-
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- vessel highL pressure setpoint;for-entry:-into E0P-1.
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. With the_ exception discussed below, the inspectors;
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had no immediatel concerns with:the setpoints being'useda o
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In step RC/P-1 of-E0P-1, the operator was instructed
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to~ wait until PCS pressure exceeded 1800 psig_before
. actuating all four trains of the-RDS.; The pr. essure 7~
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vessel design pressure 'was 1715 psia and the emergency.
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condenser tube side design: pressure was 1700: psia.'
The basis or need for exceeding these: design pressures before initiating full.RDS was not clear.
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The basis for the. diagnostic.l.oop contained in the;
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second and third decision' steps _under RC/Q in E0P-1--
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was notipresented in the basis-document, w
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'The basis-for the valve manipulation logic-in-
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' step CN/P-3 in~E0P-2 was not presented in the basis-document. -
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The-basis document contained n'o discussion of:how adverse environmental-conditions in containment impacted instrument channel accuracy or how any -
impact was accomodated.inLthe setpoints chosen. >
Discussions with' licensee personnel indicated this had-been~done but not documented in.the basis-
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document. This_ effort should'be referenced.or
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discussed in the baris document.
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i The resolution of-these issues will be tracked as an Openitem-(50-155/90008-01).
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(b) PSTG and E0Ps
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The PSTG provided the technical outline on which the E0Ps were based. The E0Ps were the procedures actually used by-the operators.
At BRP the E0Ps were presented in flowchart format.
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The inspectors identified no technical concerns with these documents-except as discussed above for the basis document.
Two concerns were identified during the table top scenarios discussed in paragraph 2.c.(3). The inspectors also:had some concerns with E0P presentation and consistency with the Writers Guide (See Paragraph 2d) and. inconsistencies between documents.
(See Appendix B for examples). These observations indicated a-need for a more thorough verification effort by the~1icensee.
(t) Plant and Control Room Walkdowns The ins'ectors walked down the E0Ps and supporting procedures p
in the control room and plant with personnel who would normally-perform these activities.
Only one item of significance was-identified.
The containment temperature monitor was located on a back' panel in the control room. The maximum temperature that could be recorded was 120 F.
Containment temperature was one of the controlling' parameters used in~the execution of'EOP-2. The s
licensee stated that the instrument location problem had been i
identified-during the DCRDR-and was scheduled to be moved-to a front panel during the next refueling outage. The rather limited range of the instrument was under review for possible extension at'that time.
Considering the fact that the environmental. qualification temperature for instrumentation inside containment was 235 F, increasing the containment temperature monitoring range to include this t.emperature appears well justified.
Resolution of this issue will be-tracked as an Open Item (50-155/90008-02).
Other observations made during the walkdowns are presented in Appendix B.
(3) Tabletop Exercises Four scenarios exercising various Emergency Operating Procedures (EOPs) were discussed in detail with an operating crew consisting-of a licensed Senior Reactor Operator (SRO), two Reactor Operators (R0s), and a Station Technical Adviser (STA). The.
scenarios used and corresponding E0Ps utilized are summarized below:
Scenario Primary E0P Utilized Anticipatory Transient E0P-01, E0P-02, and E0P-4, without a scram (ATWS)
Contingency-4.
with small Loss of Coolant Accident (LOCA)
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Containment Pressure E0P-02 and E0P-03,
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and Temperature Contingency-1.
Excursion
Containment Level E0P-01 and E0P-03, Control
_ Contingency-3.
The crew used the E0Ps and EPIPs (Emergency Plan Implementing Procedures) well. The satisfactory performance of the E0Ps was
consistent with the three crews evaluated-for the Big Rock _
Point NRC Requalification Examination on' November ~1, 6, and 11,
- 1989, atsthe General _ Electric Dresden Simulator (See NRC
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Examination Report No. 89-02). Two E0P procedural items of-
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concern were. identified with the ATWS scenario:
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l E0P-02, " CONTAINMENT CONTROL," Revision 6, Sheet 2, stated that with a "feedwater/ steam flow mismatch and-turbine load decreasing then manually scram [the] reactor and initiate emergency-condenser." Given the situation for the ATWS scenario (i.e., no liquid poison 1 system injection available due to loss of' nitrogen pressure with main condenser available) the initiation of the: emergency condenser only. aggravated the problem posed by causing the reactor power output to-increase for the additional heat'
load.
The step referred to in EOP-02 should_be evaluated to determine-if it'would;be appropriate to insert a verification of the manual scram success before initiating the emergency condenser.
E0P-01, " PRIMARY SYSTEM CONTROL," Revision 4, Sheet-1, provided several situations (e.g., reactor pressure greater than 1,360 psig and rising,-etc.) to " inject boron and isolate cleanup system."
For the situation given in the scenario, with reactor pressure being maintained below 1,360 psig and steam drum level above -17 inches, no direction was provided in E0P-4, Contingency 4', Revision 3, Sheet 4, to use procedure.EIP-2, " ALTERNATE BORON INJECTION."
Alternate boron injection would be used per the E0P only
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if control of reactor pressure or steam drum level was Icst.
The option to consider the use of alternate boron injection should be done immediately after the primary method to inject boron fails in E0P-01.
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Resolution of these items will be tracked as an open item (50-155/90008-03).
Also, it would be useful to consider developing a method to:
accurately _ identify each step in an E0P flow diagram (e.g., step number, grid number, etc.).
-(4) Verification and Validation of the E0P3
- c A review of the Big Rock Point system of on going evaluation
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and revision of E0Ps was conducted to assess whether the licensee's' current system could ensure high quality E0Ps over
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time. The system was evaluated'on,the oasis of a number cf elements, including but not-limited to:
a.
The completeness of a method for ensuring that changes in plant design, technical-specifications, technical guidelines, the writers guide, referenced plant procedures,-and the control room would be promptly reflected in the E0Ps; b.
The' completeness of a method for revising the E0Ps to-reflect findings from operational _ experience and use, training experience, simulator exercises, and control room and in plant walkdowns; c.
The timeliness of revisions to the E0Ps when incorrect or incomplete information has been identified; d.
the adequacy of the system for determining necessary training.. validation, and verification, when procedures ate changed or revised; e.
the adequacy of basis documents, including technical quidelines and writer's guide; f.
the adequacy of verification and validation; and g.
the effectiveness of a system of soliciting and utilizing feedback from procedure users and other cognizant personnel.
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The Big Rock Point Nuclear Power Plant program for ongoing maintenance of E0P's was controllad by and documented in Volume 1, Procedure 1.1, " Procedures Program." This document described the general requirements and methods for review, revision, and approval of all Big Rock Point Plant procedures including E0Ps.
Volume 1,.
Procedure 1.1 was reviewed and found to be adequate to control the E0P revision process, with the exception of the following weaknesses:
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Vol. 1 Procedure 1.1 did not require that E0P changes / revisions be sub.iect to a Verification and Validation (V&V) program. - E0P changes / revisions should be verified and validated in accordance
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with pre-established criteria that specify the extent of the d
required > verification and validation. The licensee had agreed to specify. requirements _for Verification and Validation of E0P
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revisions in an update to' Procedure 1.1.
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The; original Verification and Validation effort for the E0Ps
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did not include EIPs or any procedures referenced from the -
E0Ps. The V&V program should extend to the total E0P-network, that is_all procedures referenced directly from the EOPs-The
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licensee.had agreed to review and evaluate the extension of the V&V program to the total E0P network.
The original V&V effort did not include-in plant walkdowns.
V&V on E0P revisions should include in plant walkdowns as one j
of the validation-methodologies. The licensee had agreed to
- include this requirement in the revised Procedure 1.1.
d.
Human-Factors Review j
k The E0Ps were reviewed for consistency with guidance provided in Big
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Rock Point writers guide (Volume 1, Procedure 1.1.1, " Procedure'-
Writer's Requirements and Guideline") and accepted human factors j
principals as described in NUREGs 0899 and 1358. The review l
identified a number of areas where human factors related t
improvements could be made; however, none were determined to pose i
significant safety concerns..These concerns pertained to deviations
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between writers guide and'the E0Ps, or with_ inconsistencies within
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the writer's guide itself.
Specific examples of these concerns are j-detailed in Appendix B, j
(1) General E0P/ Writer's Guide Comments-l
Cautions - Cautions within the E0P flowchart were highlighted I
through the use of a hexagonal symbol. The writer's guide on i
pg 51 item c. stated that cautions would be in bold print and
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highlighted.
The use of appropriate symbology to distinguish cautions from command steps was satisfactory.
The licensee
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agreed to revise the writer's guide. ~ Also the writer's guide
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stated on page 53 item c that more than-one caution could be i
contained within a caution symbol.
This was contrary to
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NUREG 0899 guidance and good human factors practice. The l
licensee concurred and agreed to change the writer's guide
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to reflect only one caution per caution symbol.
Branching - The human factors review of the E0Ps revealed a number of inconsistencies in branching between E0Ps as well as b
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- tootherprocedur's'(EIN, ONPs and. SOPS). The f011owing-
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concerns were identified:-
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' Exit l arrow symbology was not always utilized to
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transfer'to another procedure or step..
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-Inconsistent ~ use of GO TO: terminology.
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.-Lack of clarity regarding exiting and concurrent performance of a referenced procedure,
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E The licensee had. agreed to_ review and evaluate these concerns and to develop a consistent referencing methodology. This.would'
include the development of a distinct method of differentiating
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between exiting and concurrent performance of referenced-j[
procedures.
Flowpath.--The writer's guide defined ~a dashed line as a conditional or alternate flowpath.- The dashed line was incorrectly. used in some-instances, -particularly. in. exit
.i transfers where no-alternative existed. The-licensee had
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agreed to review.and evaluate this concern to ensure
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consistency with the ; writer's guide.
Readability - The current size of the text in the flowcharts-
was well below that recommended by standard human engineering
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guidelines.. Letter height of the current.E0P flowcharts was approximately.0433 inch, while a minimum acceptableLper
. standard human factors-guidelines would be approximately.
.0785 : inch.:.In the interviews with the operators small type size and difficulty in reading were cited by several operators..
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The licensee agreed to evaluate this concern with'the operations department since it was their-decision to go with the current
"E" size flowcharts. 1The-inspection' team recommended the use
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of "G" size flowcharts-as a minimum'
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i Placekeeping - Placekeeping methodology was not addressed in the writer's guide.
Interviews revealed various placekeeping methodologies were used by the. control room staff. Use~of
marker / grease pencil and a stick on tape were some of the
methods cited. -The licensee agreed to document an' appropriate-
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placekeeping methodology in the writer's guide.
Logic Terms - Use of logic terms was inconsistent with writer's r
guide guidance. The inspection team found numerous examples of underlining of non-logic terms, confusing use of conditional
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statement.
Specific examples.of these findings can be found in
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concerns to ensure consistency with the writer's guide.,
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e 0ther. human factors' comments contained in the SER for thes
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< Procedures. Generation: Package should also.be evaluated for j
g program changes.
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E I(2)1OperatorInterviews.
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Operators-(R0s and'SR0s) were interviewed to determine their-I
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executing the procedures as part of:the' control' room _ team.7
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input:in revising'the E0Ps, and overall satisfaction with>the-
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technical accuracy and useability of the procedures. -Interviews o'
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roles in supporting the implementation'of-the E0Ps.:
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In: general,: licensed operators demonstrated-a good understanding N
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a accuracy and usability of the procedures with the exception =
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'that.several of_the. operators specifically cited the small text'
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size in the E0P flowcharts as~being suboptimum. This finding-
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. supported the human factors review of the E0Ps which also-
P concluded that text size in the flowcharts may be a problem.
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Interviews with Auxiliary operators were also conducted.
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- Non-licensed training did not specificall,y addressLauxiliary j
? operator _ actions in support of the'.EOPs. This weakness'was
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'also identified;in the review of the' training program andiwas 9^
beingaddressed.(see: paragraph 3). The-inspection team considered the inclusion of auxiliary = operators in the
simulator training as a positive training approach. All A0's N
. interviewed thought this was one of their most valuable
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training experiences.
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Training and Qualification Effectiveness-L M
Only one performance based training concern was identified during this-
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inspection. During interviews several SR0s Oave' differing interpretations 1 _
'of Technical Specification-11.3.1.4F.
One SR0 stated he would not declare the core spray system inoperable if both fire pumps were inoperable.
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Other SR0s would declare it inoperable and enter the LCO. At BRP, the only source of water for the core spray system was via~ the fire pumps j
(except in the recirculation mode). Some of the confusion may result, in-j
- part, from the confusing wording of TS 11.3.1.4.E, which addresses fire l
Lpump operability. Resolution of this issue from the training standpoint J-and potential Technical Specification revision will be tracked as an open j
'x item-(50-155/90008-04).
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=The inspectors also reviewed the. training programs relative to EOPs for i
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licensed and non-licensed operators.
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(a) _' E0P Training of-Licensed Operators:
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The licensee training department has developed, since the NRC
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E Requalification Examination in November 1989, several job-h performance measures (JPMs) to specifically address tasks unique for l
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b the E0Ps (See NUREG-1021, ES-601, Revision 5; for discussion of L'
JPM). The following E0P JPMs were reviewed for technical content
W and found to be acceptable:
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JPM u.
Revision i
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Number-Task
040501
Perform actions associated with L
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conditions requiring liquid poison
injections.
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040601
Ability to ope. ate a reactor-
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depressurization system (RDS)
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Opening main steam isolation valve
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050501
Initiate containment sprays.-
,g 050502
Place post-incident system into long
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term cooling.
10T01
Perform actions to shutd w the reactor with a fire in the Control Room in accordance with EMP-3.10
" FIRE."
.300102
Connecting a shutdown pump using EMP-3,10. " FIRE. "
100103'
Connecting a reactor cooling water pump to the emergency diesel generator through the Alternate Shutdown System (ASD).
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1000104
Perform line up of emergency diesel J'
generator to ASD.
In addition to the above tasks, several other JPMs were technically reviewed and walked down, in preparation for the NRC November 1989 Requalification Examination (e.g., manually starting diesel generator, c
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startup of the Shutdown Cooling System, control of steam drum level remotely in manually, etc.).
These JPMs involved expected license operator actions both inside and outside the Control Room;
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-(b) E0P Training for Auxiliary Operators:
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No performance based training. concerns were identified for-non-licensed operators; however, no formal documented E0P training
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was identified.
The doeurrentation for the job certification for
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i several auxiliary operators (AO) was compared with tasks required to
be performed outside the Control Room to implement the Emergency
- l Operating Procedures (EOPs). Alt nugh the A0 training and
certification (Adm$nistration Procedure (AP) 1.7.2, "0PERATOR TRAINING," Revision 1, pp. 27-69, pp. 205-206) covered system normal
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lineups, the A0 training did not delineate specific E0P tasks to be done outside the Control Room.
For example, the following specific E0P tasks were not Jocumented in the training program material to directly verify that an A0 was capable of the following E0P tasks:
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lining up fire water to hotwell using MO-7073 and MO-7074 (EOP-01, Revision 4, Step RC/L-2),
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utilization of alternate boron injection (EOP-4, Revision 3, implementation of EIP-2, Revision 1).
^ throttling ECCS flow u:ing VFP-29 and VFP-30 (ECT 4, Revision 3 Step C4-2.1)
The addition of E0P related OPMs would provide assurance of continued A0 capability to perform these tasks.
4.
Quality Verification Effectiveness The licensee had performed a 10 CFR 50.59 evaluation of the E0Ps to assure they-fell within the design basis of the plant as described in the UFSAR or had been approved by an appropriate NkC SER (either generic or plant specific). This was considered a strength.
Only one surveillance of E0P or related procedures had been performed since.the E0PS had been irrplemented.
This surveillance entailed a detailed walkdown of the Alternate Boron Injection Procedure (EIP-2).
Several problems were identified and corrected. A walkdc vn of this procedure by the inspectors identified no additional problems.
As discussed in Paragraoh 2.c.(2), additional quality oversicht was
required in the E0P verification process.
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~Open Items r-i
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Open Items art, matti!rs which have been discussed with the licensee which
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will be reviewed further by the inspectors cr which involve some actions
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on the'part of the NRC'or licensee.or both. Open items disclosed during.
this inspection are described in paragraphs l2.c.(1),(a), 2.c.(2),
E 2.c(3),and.3.-
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Exit Interview-l
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The inspectors. met with licensee representatives (denoted in
Paragraph 1) on April 26, 1990'
The inspectors summarized the purpose,.
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scope, and. findings of:the inspection and the likely informational content of_the inspection. report.
The licensee acknowledged thist
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information and did not identify any information as proprietary.,
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Appendi d.
- D_escription of Inspection Activities
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'The following E0Ps and related procedures were reviewed during the desktop i
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reviews-and walked down as described in Paragraph 2.b.
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v EOP-1 Primary System Control-
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. E0P-2.
. Containment Control
E0P-3 Radioactivereleasecontrol; Contingency 1-Levelrektoration;.
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' Contingency 2-Emergency Depressurization; Contingengy 3 - steam
cooling, E0P-4 Contingency 4 - ATWS
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EIP-2=
. Alternate Boron Injection
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EIP-3 Bypassing containment High Pressure Isolation ' Scram
- EIP-4
- Venting:the-PCS with one Train of RDS
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Appendix B t
Detailed Comments E0P-1
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RC/L-1 directed the operator to " confirm initiation of" and then listed 5 items. The words "as applicable" should be added to this statement due to the fact all items in the list may not be applicable.
RC/L-2 directed the operator to open MO-7073'and MO-7074 Direction to'
open VPI-33 should be added.
- RC/L-2, bo11et concerning fire water:
Manual valve FR 33 must also be
. opened; it 15 in series with 7073 and 7074.
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RC/P - The second caution in the flowpath directed the operator to
" Observe Reactor Pressure." This is an operator action and should not be contained within a caution symbol.
- RC/P-1: - The. guideline required initiation of the ersrgency condenser if PCS pressure is >1435 or if any SRVs are cycling.
The E0P did not consider SRV cycling.
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RC/P-1:
The term " gross fuel failure" was undefined.
- RC/P-3 - This step stated.,... THEN DEPRESSURIZE PCS BY MAINTAINING w
<100 DEGREES F/HR C00LDOWN. BY is not a logic term and should not be underlined.
- RC/P-2 - This step c..etained the use of AND and OR in ways which could have two possible interpretations.
This statement should be rewritten to-
' eliminate confusion.
- RC/P-2:
Typo; the EOP, guideline and basis used 1385 psig; the flowchart.
on pg. 23 of'the basis used 1535 psig.
- RC/P-3 - The word " RODS" should be the singular " ROD".
- RCP - Several decision boxes were not worded as questions contrary to guidance contained in the writer's guide.
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RCP - Transfer to EIP-4 was made without.use of the exit arrow symbol.
Exit / Entry symbols should be used consistently.
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- RC/Q-1 - This step. contained an incorrect use of the term IF NOT as defined by the writers guide.
IF NOT should only be used following a
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ONP - 2.31 was entered upon a reactor scram.
It should be referenced at i
the beginning of this procedure.
- CN/P-3 - CONFIRM was underlined contrary to writers guide direction. Only logic terms should be underlined.
- CN/L-3 - SOP-8 was referenced uithout use of exit arrow symbology.
Exit / Entry symbols should be used consistently.
- CN/P-2 had two flowpaths exiting from this command step.
This was the i
only command step in the E0Ps that had two flow paths emanating from a command step.
The use of a decision symbol may be appropriate.
- CN/L-4 was mislabeled and should be CN/L-3.
Terminology contained in Exit / Entry arrows was not consistently used.
Some used GO T0, while others did not. Use of Exit / Entry arrows and terminology should be consistently applied.
- Step C1-6.2 was contained within a caution symbol. This is a command step.
- Step C3-1 began with " Confirm.
CONFIRM should not be
"
....
underlined.
- Transfer to EIP-4 should be via an EXIT symbol.
" CONTINUE TO ATTEMPT REACTOR SHUTDOWN AS TIME AND PLANT CONDITIONS PERMIT
THEN GO TO ONP2.9 CONCURRENTLY" was contained in a caution symbol.
Tnis caution contained an operator action.
Recommend eihrinating "THEN GO TO ONP2.9 CONCURRENTLY". Use of exit arrow will direct operator to ONP2.9.
Add words GO T0 to the exit arrow wording.
- _C4-4 directed the operator to return to step C4-1.0.
This should be accomplished via an exit arrow symbol.
- Terminology GO TO is missing in several exit arrow symbols.
EIP-2
Section 3.0 contained prerequisites for performance of this procedure.
Two auxiliary operators and assistance from maintenance were required to perform this procedure and should be specified in this section.
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Section 5.3.le required the A0 to call the control root and directed the contr71 room staff to close MO-7073. Words should be added that-specifically state this.
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Section b.3.2b - VW-8 should be added as item 4
Section 5.3.2d directed the operator to start the sludge pump. This would be accomplished by opening valve NWA2 and positioning the switch to ON. These specifics should be added to the procedure.
Control Room.
- Control Room panel C01: The feed pump start operator aid plaqua did not specify time (rated I start; ambient 2 starts").
Per unit time was not specified, ie, per hour, per day, per week?
Most control circuit indications (e.g. breaker or valve open and closed)-
were single bulb, single filament type. NUREG-0700 stated these should be-double bulb / double filament, The alarm panel tiles were OK and followed the NUREG-0700 guidance.
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-. C01 :
Poison system P1 376 and 377 labels had tape add-ons which read LO and Hl. These labels should be permanent.
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Col: No engineering unit labels on percentage scales for level'
indicators LI. 3300-3303, 3305 and 3325.
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