ML18151A515
| ML18151A515 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 05/09/1990 |
| From: | Peebles T, Schin R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18151A516 | List: |
| References | |
| 50-280-90-09, 50-280-90-9, 50-281-90-09, 50-281-90-9, NUDOCS 9005220376 | |
| Download: ML18151A515 (93) | |
See also: IR 05000280/1990009
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTA STREET, N.W.
ATLANTA, GEORGIA 30323
Report Nos.:
50-280/90-09 and 50-281/90-09
Licensee:
Virginia Electric and Power Company*
Glen Allen, VA
23060
Docket Nos.:
50-280 and 50-281
Facility Name:
Surry 1 and 2
License Nos.: DPR-32 and DPR-37
Inspection Conducted:
April 2 - 12, 1990
Inspector: ~
-f-
. _ *-
R. Schi n, Team Leader
Approved
Scope:
NRC learn Members:
J. Arildsen
NRC Contractors:
L. Mellen
M. Shannon
J. York
Haney,
Meeker, C_
~rt. T .. A. Peebles., Chief
Operations Brarich
Division *of Reactor Saf
SUMMARY
s-r-7o
Date Signed
Date Signed
This was
a
special
announced Emergency Operating Procedure (EOP) team
inspection.
Its purpose was to verify that the Surry EOPs were technically
accurate, and that their specified actions coulc! be accomplished using
existing equipment, controls and instrumentation.
The inspection evaluated
the adequacy of the licensee's EOPs [includi.ng Abnormal Procedures (APs) and
Fire Contingency Actions (FCAs)], conformance of these procedures to the
Westinghduse
Owners'
Group
Emergency
Response
Guidelines
(ERGs),
and
conformance to the approved writer's guide.
The inspection included a
comparison of the EOPs to the ERGs, a technical adequacy review of the
procedures, control room and in-plant walkthroughs, simulator eva*luation of
selected procedures, a review of on-going control of these: procedures, and
inte.rviews of operators who use the procedures .
\\
90051:5
0!50002:3()
FDC
2
Results:
The overall assessment concluded that the EOPs adequately covered the broad
range of accidents and equipment failures necessary for safe shutdown of the
plant, and were capable of safely shutting down the plant and placing it in
a stable condition.
The team identified as licensee strengths:
use of the
same writer
1 s guide for EOPs and APs, paragraph 3; and simulator support for
EOP inspection, paragraph 5.
The team identified weaknesses in procedures,
inc.luding.:
important operator actions missing
from
EOPs,
paragraph 3;
incorrect or incomplete directions to operators, paragraph 3; and inaccurate
or inconsistent use of key words and symbols, paragraph 3.
Violations or
deviations were not identified in this report .
REPORT DETAILS
1~
Persons Contacted
Licensee _Employees
- 0.
- W.
- R.
T.
- M.
A.
- D.
M.
L.
D.
- E.
G_.
- B.
- E.
- 0.
H.
R.
I.
- M.
w.
T.
- P.
- J.
J.
w.
- R.
G. s.
R.
- E.
D.
A.
E.
N.
A.
R.
R.
Beith, Procedures Group Human Factors Specialist
Benthall II, Licensing Supervisor
Blount II, Procedures Grau~ Supervisor
Bowden, Control Room Operator
Bowling, Assistant Station Manager, North Anna
Brown, Nuclear Training Supervisor
Christian, Assistant Station Manager, O&M -
French, Reactor Operator
Gardner, Senior Nuclear Training Instructor
Glover, Reactor Operator
Grechek, Assistant Station Manager, NS&L
Griffin, Control Room Operator
Gros~, Shift Operations Supervisdr
Harrell, Vice President, Nuclear Operations
Hart, Quality Assurance Supervisor
Johnson, Coritrol Room Operator
Johnson, Assistant Shift Supervisor
Jones III, Control Room Operator
Kansler, Station Manag~r
Kreheley, Senior Training Instructor
Kunkle, R~actor Operator
Linn, Contractor, Volian Enterprises Inc.*
Logan, Senior Staff Engineer
McCarthy, Operations Superintendent
Moore, Senior Nuclear Training Instructor
Mushenheim, Procedures Group Lead Writer
Prescott, Assistant Shift Supervisor
Ross, Reactor Operatpr
Scherer, Reactor Operator
Smith, Quality Assurance Manager
Souza, Assistant Shift Supervisor
Swander, Procedures Group
Turko, Testing Supervisor
Turner, Reactor Operator
Wheeler, Jr., Shift Supervisor
Yzzi, Assistant Shift Supervisor
Zoldork, Reactor Operator
Other licensee employees contacted included engineers, technicians,
operators, and office personnel .
\\
2
NRC Personnel
- W. Holland, Senior Resident Inspector
- C. Julian, Engineering Branch Chief, DRS
- P. Kellogg, Operational Programs Section Chief, DRS
- Attended exit interview on April 12, 199.0
- Procedures reviewed during. this inspection are listed in Appendix A.
A list of abbreviations used in this report is contained in Appendix E.
2.
EDP comparison with ERGs and Regulatory Guide 1.33
The tea.m compared the index of Surry EOPs with the index of NRC approved
ERGs and found that the licensee had an EDP corresponding to each ERG with
the exception of ES-0.4, Natural circulation cooldown with steam void in
vessel without RVLIS.
The licensee had an operabl~ RVLI5.and had an EDP
ES~0.3, Natural circulation cooldown with steam void in vessel with RVLIS
(Appendix B, ES-0.3, comment h).
The team compared the 1 i censee I s index of EOPs and APs with the NRC
Regulatory Gui de 1. 33 1 i st of procedures for addressing emergencies and
other significant events.
This comparison determined that the licensee
had deve 1 oped sufficient procedures to cover the broad spectrum of
accidents and equipment failures.
.
.
The teal!) compared the Surry EOPs to the ERGs and found that they followed *
.the accident mitigatirin strategy and action sequence of the ERGs, except
as properly justified in the Surry EDP SOD or as noted in Appendix B to
this report.
In addition, the team evaluated EOP step deviation~ from the
ERGs incident to . the EDP wa l kthroughs and found them to be properly
justified and documented, except as noted in Appendix B.
The licensee's
EDP procedure entry and transition conditions closely followed the ERGs.
The licensee had made many changes to individual EOP steps from the way
they were written in the ERGs; to adapt them to the Surry uni ts, imp rove
the human factors, and comply with the writer's guide.
In addition, the
licensee's EOP steps contained some technical improvements over the ERG
steps.
In most cases, these EDP step deviations from the ERG were
adequately justified in the EOP
SOD.
The. licensee's plant specific
technical guidelines (PSTG) included the EDP
SOD, the EDP Setpoint
Document, and the Westinghouse Owners* Group ERGs.
The team identified a
few cases where the resulting EOPs deviated from the ERGs without adequate
justification in the EDP SOD.
These EDP deficiencies are described in
Appendix 8; where they are identified by
11 PSTG DEV 11 *
The team concluded
that none of these EOP deficiencies precluded the ability of the EOPs to
safely shut down the units.
There were no violations or deviations noted in this area.
3
3.
Independent technical adequacy review of EOPs
The team reviewed the licensee 1*s EOPs and selected samples of the APs and
FCAs, as listed in Appendix A,
for technical adequacy.
During this
review, the NRC identified numerous examples of technical or human factors
deficiencies as described below and in Appendix B.
The team found that the 1 i censees I
procedures did not fully address
important operator actions that were identified by an engineering study
on loss of intake canal.
(The elevated intake canal was the licensee's
For example, procedures did not address required
times for performing certain operator actions to preserve intake canal
inventory (Appendix B, AP 12.01, Steps 1 and 10a).
Procedures also did not
address local operation of condenser waterbox SW valves to preserve intake
canal inventory (Appendix B, ECA 0.0, Step 7b).
Manual valves required for
quickly cross connecting charging between units to protect RCP seals were
not clearly labeled or readily accessible (Appendix B, ECA 0.0, Step 10).
Also, during a loss of all _AC po~er, the operators had no ?ffective means
of communication with personnel outside the control room.
This lack of
communication severely hampered the ability of op~rators to implement the
In response to these concerns, the 1 i censee * stated that they would
immediately revise AP-12.01, Loss of intake canal, with the revision to be
issued within one week.
Affected steps in that procedure are indicated in
Appendix 8.
The licensee also stated that they would initiate operator
use of sound powered phones for certain emergency conditions within one
month, pending availability of required equipment.
In addition, the licensee stated that they would make a design change to
the operations radio repeaters, to provide emergency power from batteries.
These three corrective actions are identified as IFis 280,281/90-09-01,
02., and 03.
Licensee design calculations for intake canal level showed that, in an
accident situation, the condenser waterbox valves must be closed within
70 seconds to preserve s~fficient water in the intake canal to handle a
LOCA.
A plant vulnerability was noted in that a loss of all AC power
to a unit for 30 minutes could cause a LOCA due to RCP seal failures
(Appendix B, ECA 0.0, Step 10) and without AC power to either unit, which
is beyond the design basis, the operators could not close the condenser
waterbox valves in time to prevent the intake canal from draining well
- below the required level (Appendix 8, ECA 0.0, Steps 7a RNO and 7b).
Once
the canal level was down, it could not readily be regained.
The three
diesel powered emergency service water pumps could be started within two
hours, and then could pump water into the canal at a rate that could be
sufficient for both units to remain at hot standby, but less than the flow
rate of cooling water needed for a unit during a LOCA.
With a low intake
canal level, the eight large circulating water pumps that normally
supplied the large flo~ rate of water into the intake canal for both units
could not be started ev~n if their offsite power was restored ..
4
These pumps required a vacuum assist in their discharge p1p1ng, and with
the end of the discharge piping above the canal level, the required vacuum
could not be obtained.
In essence, once the canal level was lost, it
would be very difficult to be regained during a LOCA.
The licensee is
conducting an IPE.
The plant vulnerability to this beyond the design
basis event will be addressed in that analysis.
This is identified as
IFI 280,281/90-09-04.
Technical deficiencies identified by the team included incorrect direction
for operators.
One EOP step failed to recognize that the instrument used
to verify RCP s~al return temperature would have been isolated from the
flow path during an SI condition.
One EDP step incorrectly directed the
operator to transition past an important action (Appendix B, ECA-1.1,
step 22b RNO).
Another type of technical deficiency was the existence of
some incorrect setpoints in EOPs.
For example, one EOP included a natural
ci rcul at ion cool down rate of 100 *degrees per hour, when the setpoi nt
document number was 50 degrees per hour (Appendix B, ES-0.2, step 3a).
In
addition, the team observed some mathematical errors in the EOP setpoint
document.
There was an apparent weakness in the VP review and receipt
inspection of the setpoint document, which had been prepared by contractors.
The team concluded that none of the technical deficiencies precluded the
ability of the operators to use the EOPs to safely shut down th~ units .
The team found the technical adequacy of the APs and FCAs to be less than
that of the EOPs.
They were less complete, provided 1 ess guidance and
detail to the operators, and contained many more technical errors.
For
example, one FCA step directed operators to initiate the cold overpressure
protection system with the PORVs disabled (Appendix B, FCA-1.01, step 54).
Another FCA step directed operators to establish conditions for starting
an RCP when they could not do so (Appendix B, FCA-1.01, step 33).
Entry
conditions for one AP depended on an alarm whose setpoint had not been
changed to reflect current plant conditions.
(Appendix B, AP-24, entry
condition 2).
The APs and FCAs also had less well defined entry conditions.
In addition, these procedures were not written to the standards of the EDP
WG.
However, the new EOP dual column WG applied also to APs, and the
licensee was upgrading the APs to meet the standards of this WG.
Having
the EOPs and APs written to the same standards will aid the operators by
removing some potential confusion when using EOPs and APs together.
In response to the NRC identified technical deficiency concerns, the
licensee stated that they had changed the alarm setpoints of the steam
generator blowdown radiation monitors to be correct prior _to the exit
interview.
The licensee also stated that they would promptly revise 18
steps in 15 procedures.
These procedure revi.si ans were to be issued
within about 12 weeks, with Rev. 2 to the EOPs.
In addition, the licensee
stated that one procedure step would be revised. before the next refueling
outage.
These procedures and steps are identified in Appendix B.
4.
5
The licensee also stated that the remaining technica*l and human factors
items in Appendix B would be addressed with Rev. 3 to the EOPs as
appropriate.
These three scheduled corrective actions are identified as
IFis 280,281/90-09-05, 06, and 07.
The team found the EOPs to be unnecessarily complicated by many instances
of inaccurate and inconsistent use of key words and symbols.
Examples
included the words verify and should, transition terms, bullets, and
asterisks.
In addition, the use of cautions or notes often conflicted
with WG direction.
The inconsistent application of the writer 1 s guide
made the EOPs not amenable to verbatim operator compliance.
In general
the operators during interviews and simulator scenarios were able to
properly use the EOPs in spite of the writing deficiencies.
Operators did
not follow the key words and symbols when they appeared to conflict with
a procedure 1 s intent; however, these conflicts did lead operators to
interpret the intent of some steps differently.
Human factors comments
are addressed in Appendix Band writer 1 s guide comments are addressed
in Appendi,x C.
The licensee stated that they would address Appendix C
items* with Rev. 3 to the EOPs.
Appendix C items are identified as
IFI 280,281/90-09-08.
The team found the licensee 1 s EOP development, revision, and maintenance
procedure to be well organized and comprehensive. It included a verifi-
cation of EOPs by a station engineer and a human factors specialist as
well as a licensed operator and a procedure writer.
In addition, it
required in plant walkthroughs to determine if local actions could be
conducted as required.
These were in addition to the simulator
validation.
The. licensee
1 s writer
1 s guide for dual-column procedures was generally
well organized and complete, but was lacking in some important specifics.
Deficiencies identified with the writer 1 s guide are described in Appendix C.
The team found the EDP step deviation documents to be typically well
organized and cross referenced.
The EDP set point document was also well
organized; however, it had some inconsistencies in the quality of the
documentation.
In general, the information required to verify the set
points was referenced and was available.
There were no violations or deviations noted in this area.
Review of the EDPs by control room and plant walk-throughs
The NRC conducted control room and in plant walkthroughs of the EOP, AP,
and FCA procedures listed in Appendix A.
Transfers and branching were
checked and found to be proper, except as noted in the Appendix B.
The
team found the relationship between EOP. procedure nomenclature and
equipment labeling to be generally consistent and capable of being easily
understood without confusion, except as noted in Appendix 0.
This was
also the case with APs and FCAs; although more instances of inconsistent
procedure nomenclatur\\ and equipment labeling were found.
The team
noted an improving trend in equipment labeling and pl ant housekeeping.
Appendix D items are identified as IF! 280,281/90-09-09.
6.
Excepi;. as noted in Appendix B, the team found that i ndi ca tors, contro 1 s,
annunciators, and other equipment required by the EOPs, APs, and FCAs were
a~ailable.
The team noted during a limited number of walkthroughs that
some operator~ had difficulty in locating all controls and equipment used
in thete procedures~
The team also noted that ARPs have not received as
much licensee attention as other procedures -
some had not been revised
since 1975 or 1976.
In the control room, one complete set of controlled
EOPs, APs, and FCAs for each unit was located on the apprdpriate side of
the dual unit control room - these procedures were current and readily
available to the operators.
While the results of *the walkthroughs were generally satisfactory, .
many discrepancies in the areas of technical accuracy, writer 1s guide
adherence, and human factors were noted.
.Techni ca 1 and human factors
discrepancies are noted in Appendix B, writers guide discrepancies* in
Appendix C, and nomenclature discrepancies in Appendix D.
There were no violations ~r deviations noted in this area.
5_
Simulator observation
The team. observed two crews I performance* on the Surry Power Station
simulator. The following six scenarios were performed over a two-day
period:
CSF status tree red.path during transfer to cold leg recirculation
Loss of all AC (pre.ceded by fire. and loss of offs'ite p'ower)
Large break LOCA from RHR conditions
Main control room evacuation
Steam generator tube rupture and LOCA
The operators performed the EOP exercise scenarios in r.eal time on the
simulator with the exception of the last half .of the main control room
evacuation scenario.
When the crew evacuated the main control room in
that scenario, the simulator was frozen, and a walkthrough discussion of
the crew* s follow-up actions was conducted. The team noted no significant
deficiencies with accident mitigation strategy or recovery action *in any
scenario. The crews were able to successfully incorporate their training
and experience with the procedures to mitigate the accidents. . The
procedures did not unnecessarily dup 1 i cate operator actions, and the
procedures did not cause the operators to physically interfere with each
other while performing the EOPs.
Ih g~neral, the procedures ~nsured the
operators addressed all the required prerequisite actions for their
.transitions from one EOP to another.
After one scenario, the operators stated that there was no pre-planned
procedure in p 1 ace for a fire in the switchyard.
They exhibited a need
and expressed a desire for such guidance.
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7
EOP place kee~ing aids; which included a location for a check mark by each
high level action step in an EDP to indicate completion of the* step, were
not consistently used by the operators.
The operators stated that they
had not been trained o~ the use of these place keeping aids.
Prior to
these EDP *exercise scenarios, the training staff had directed that the
check marks not be u~ed in simulator training. Operators also indicated a
need for ~lace keeping aids to be used to mark pages when transitioning
amo~g a number of procedures.
Sound powered phone jacks were not available in the simulator, but wete
in the main
control
room.
Operators stated that such means of
communications may be useful in a loss of all AC power.
Operators further
stated that they had not been trained on sound powered phone communications.
The controlled keys in the. simulator differed from those in the main
control room in location and labeling.
Although prebriefed on this
matter, one crew .exhibited confusion in obtaining certain EDP required
keys.
In the performance of the evacuation of the main control room scenario,
the NRC team verified that steps 33 and 54 in procedure FCA-1.01, Limiting
- MCR fire,. failed to provide adequate -guidance to the operator.
Step 33
required operators to establish conditions for starting an
RCP from
outside the main control room:.
The operators indicated .that this could*
not be accomplithed in accordance ~ith existing procedures~
They further
stated that, without the ability to establish proper conditions, they
would not start an RCP.
Step 54 required operators to place the overpressure mitigation system
(OMS) in service.
Due to the complexity of FCA-1.01, the operators* did
not realize that the PORVs had been isolated in step 10a.
The operators
indicated that by the time. that st~p 54 was performed it would probably
be performed by the relief shift and that it was unlikely they would have
~ecogniied that the PORVs were blocked.
Furthermore, the operators stated
th.at all indications they would receive. would support that the OMS was i-n
service, when in fact it wa~ not.
The two crew.s exhibited significant concern for the time required for the
initiation of step 18c in procedure E-1, Loss of r'eactor. .or secondary.
coolant.
Operators stated that the actions in -that step might better be
accomplished through the use of an attachment.
One crew stated that they were unclear in their understanding of step~ 1n
FR-C.l, Response t6 inadequate core cooling. In particular, checking the
availability of seal leakoff flow to be greater than 0.3 GPM appeared
inconsistent with the RNO action to establish the support condition.
The NRC team noted that the simulator maintained subcooling for approxi-
mately 15 minut~s without PRZR level during a simulated LOCA/SGTR.
8
The Surry Training Department's simulator support personnel were a
valuable assistance to the team in every aspect of the simulator portion
of the inspection. The joint NRC-Surry thorough scenario development and
validation efforts resulted in the simulation of well defined events which
afforded significant support to the EOP inspection process.
There were no violations or deviations noted in this area.
6.
Management control of EOPs and interfacing documents (APs, FCAs, ARPs,
OPs, etc.)
Guidelines for the preparation of emergency operating
procedures, paragraph 6.2.3, recommends licensees establish a program for
the on-going evaluation of EOPs.
Also, paragraph 6.2.4, updating EOPs,
recommends that when changes occur that will affect the EOPs, the changes
should be reviewed to ensure consistency with the technical guidelines
and the writer's guide.
The team reviewed the licensee I s procedure VPAP-506,
EOP deve 1 opment,
revision, and maintenance, dated October 24, 1989, for comp 1 i ance with
these recommendati ans.
This procedure addressed the EOP maintenance
program, which incorporated the fo 11 owing attributes:
- Writer's Guide
- V&V Guidelines
- EOP Revision Process
- Maintenance of Related Documents (basis, setpoints, etc.)
- Plant Modifications
- WOG ERG Revisions
- Feedback from Training, Operating Experience, etc
- Training on Revisions
- Safety-Evaluations/Engineering Evaluations
- Human Factors Involvement
- Management Involvement
The licensee had been using an informal process for recommended changes to
EOPs, but had a procedure revision in progress that included an EOP
specific change recommendation form.
This form required a formal feedback
to the originator. The recommended change form would be processed using
station procedure VPAP-502, Procedure Process Control, dated February 5,
1990.
The licensee was revising APs to conform with the guidelines in
procedure VPAP-50.5,
Writers guide for . dua 1-co 1 umn procedures, dated
March 12, 1990.
- The licensee had performed an EOP assessment during February 5 through 23,
1990. This assessment was performed by the Corporate Nuclear Safety (CNS)
group and identified approximately 300 items for review.
Each of these
items were to be addressed by the Surry Nuclear Station and changes were
to be made to the EOPs where appropriate.
The licensee stated that none
of the identified itB{!ls required any immediate action.
The licensee's
EOP assessment was no't issued prior to the end of this NRC
inspection .. However, the executive summary was discussed with the team
leader during the inspection.
7 .
9
Follo~ing the inspection, the NRC was provided with a copy of the licensee 1 s
assessment.
A ~eview of this document indicated the scope and depth of
the licensee 1s review was slightly different from the NRC 1 s inspection in
that it placed more emphasis on the verification and validation aspects
of EOP preparation.
The NRC inspection was broader in that it 16oked at
In.the areas where the two audits overlapped many of the general concerns
identified were similar ~n nature, although specific individual items were
for the most part different. The licensee 1 s assessment was an aggres~ive
in depth examination of the EOP 1 s, supporting procedures and documenta-
tion .. Th-is-assessment and the NRC 1 s inspection both indicate that a
large- effort is necessary to upgrade and maintain the EOP 1 s and related
procedures at the facility.
The QA group performed an EOP audit on April 19, 1989, thrbugh May 31,
1989, before the new Revision lA EOPs were complete.
The licensee
expressed plans to audit the EOP procedures on an.annual basis.
There were no violations or deviations noted in this area.
EOP user interviews
The team conducted intervie~s with seven licensed operators.
Two of the
operators were on the training staff and one of the operators was ori the
engineering staff.
One of the opera.tors was a shift supervisor.
All operators held active
licenses except one who had just gone to inactive status. The interviews
were conducted to sample the operators* opinions on the quality, usability,
and adequacy of the EOPs., to collect information on the approach to
training, and to augment the identification of specific deficiencies in
the EOPs.
.
Generally the operators were confident that the EOPs would work and could
be used in an actual emergency.
Specific comments identified in the
- interviews* are discussed below.
One operator commented that the Rev ... lA EOPs were relatively .new and
the operators were not as familiar with them as they were with other
procedures.
He felt the operations staff had not been appropriately
- involved in the preparation of the procedures.
He believed more simulator
time on the EOPs once they were finalized would be helpful.
He felt the
EOPs were workable in their present form for experienced operators but
that a less experienced operatof would have a more difficult time with
some of the more complicated EOPs.
One operator noted that the new Rev. lA EOPs contained fewer notes and
cautions than the old EOPs and felt that some of the excluded information
had been useful. Another operator commented that in some places the EOPs
were very detailed and in other places did not contain enough detail. One
interviewee felt that there were isolated areas in EOPs where more back-
ground information was needed.
10
Operators stated that division* of responsibility in the control room
during an emergency was understood and, although not formally covered in
an administrative procedure, was well covered and practiced in training.
When asked if there were any circumstances where additional staffing would
be helpful for implementing the EOPs, operators indicated the following
instances:
Station blackout, Control room fire, Intake canal loss, and
Dual unit events.
When asked if they felt they had enough time to .execute
the. EOPs, one operator said yes for most EOPs - but that for SG tube
rupture, and preventing the pressurizer from going solid upon spurious SI,
it would be close.
One operator indicated he felt the EOPs were too wordy
thus increasing performance time. Another operator indicated IA steps as
an example where the rules of usage (i.e., reading) made performance of
the procedures unnecessarily slow. Another operator interviewed also felt
reading procedures word for word slowed the process and that more training
to enhance EOP familiarity would speed procedure execution.
When asked about place keeping (i.e., transitions, branching, parallel
performance of procedures, continuous actions), operators indicated that
some method such as physical place markers in addition to currently used
check offs and shared responsibility (i.e., procedure reader and procedure
performer) would be very helpful .
Six of the operators were asked a specific example question (ECA-1.1
Step 22b RNO) involving an incorrectly written transition step to assess
operator use of procedures in terms *of step performance.
All performed
the same steps in the same order and explained consistent rationale.
While. the procedure incorrectly transitioned past step 22d, all operators
stated that they would have performed step 22d because they knew that
it was needed.
Operators indicated that communications between personnel was generally
good in the control room and between the control roam and locations
outside the control roam.
Operators indicated closed loop communication
(complete answer back) was emphasized during training.
Operators indicated that communication equipment was considered adequate
for most situations.
A concern for communications was expressed for
station blackout events.
During SB, radios would be limited to line of
sight and other communication would be limited to sound powered phone and
word of mouth.
One operator indicated that the radi as di dn I t work
well in some plant areas under normal conditions.
An interviewee felt
that for 1 ass of all AC someone from every department to serve as runner
was needed and perhaps this should be the case for every EOP. *
Most interviewees believed mare training was needed for the EOPs.
This
included more time and more specialized training for EOPs.
Operators
felt NLOs should be included in EDP training and drills.
One operator
indicated there was an ambiguous time between training on a revision and
its implementation (i.e., operator was trained on new procedure but
required to operate on previous version).
Operators indicated there was
no training on PRA or on important operator actions identified by PRA.
11
Some specific technical and human factors defi ci enci es in the EOPs were
identified by the interviews and are included in Appendix B.
There were no violations or deviations noted in this area.
8.
Follow-up on previous inspection findings
a.
(Closed)
DEV 280,281/87-32-02, Failure to meet procedure generation
package commitments in generating emergency operating procedure for
natural circulation cooldown
This deviation stated that the licensee had not provided adequate
written justification in the step deviation document for the reduc-
tion in subcooling margin as a result of the RCS depressurization
to establish RHR flow.
The team reviewed the current procedure,
ES-0.2, Natural circulation cooldown, Rev. lA, and the related SOD,
and determined that they adequately addressed the concern about
subcaoling margin.
9.
Exit interview
The inspection scope and findings were summarized on April 12, 1990, with
those persons indicated in paragraph 1.
The NRC described the areas
inspected and discussed in detail the inspection findings listed below.
IFI-04 below was discussed by telephone between R. Schin and W. Benthall
during April 18-26, 1990.
No proprietary information is contained in this
report.
No dissenting comments were received from the licensee.
Item Number
IFI 280,281/90-09-01
IFI 280,281/90-09-02
IFI 280,281/90-09-03
IFI 280,281/90-09-04
IFI 280,281/90-09-05
IFI 280,281/90-09-06
\\
Description, Paragraph
Revise AP-12.01,
Loss of intake canal,
within one week, paragraph 3 and Appendix B
Initiate operator use of sound powered
phones within one month, paragraph 3
Provide
emergency
battery
power
to
operations radio repeaters, paragraph 3
Conduct IPE analysis of loss of intake canal
resulting from loss of all AC power to a
unit, paragraph 3
Revise 18 steps in 15 procedures to address
technical deficiencies within 12 weeks,
paragraph 3 and Appendix B
Revise
one
step in AP-22.01,
Loss of
refueling cavity level, prior to refueling
outage, paragraph 3 and Appendix B
12
IFI 280,281/90-09-07
Corrective actions for all technical and
human factors comments,
paragraph 3 and
Appendix 8
IFI 280,281/90-09-08
Corrective actions for all writer's guide
comments, paragraph 3 and Appendix C
IFI 280,281/90-09-09
Corrective actions for all nomenclature and
labeling items, paragraph 4 and Appendix D
1-E-O
l-E-1
l-E-2
l-E::*-3
l-ES-0.0
l-ES-0.1
l-ES-0.2
l-ES-0.3
l-ES-1.1
l-ES-1.2
l-ES-1. 3
l-ES-1.4
l-ES-1.5
l-ES-3.1
l-ES-3.2
l-ES-3.3
l-ECA-0.0
l-ECA-0.1
1-ECA-0.2
l-ECA-1.1
l-ECA-1.2
l-ECA-2.1
l-ECA-3.1
l-ECA-3.2
- 1-ECA-3.3
F-:-0
F-1
F-2
F-3
F-4
F-5
F-6
1-FR-C.l
1-FR-C.2
1-FR-C.3
1-FR-H.l
1-FR-H.2
1-FR-H.3
1-FR-H.4
APPENDIX A
PROCEDURES REVIEWED
Reactor Tri~ or Safety Injection
Loss of Reactor or Secondary Coolant
Faulted Steam Generator Isolation
Steam Generator Tube Rupture
Rediagnosis
Reactor Trip Response
Natural Circulation Cooldown
Natural Circulation Cooldown with Steam Votd *in
Vessel (With RVLIS)
SI Termination
Post LOCA Cooldown and Depressurization
Transfer to Cold Leg Recirculation
Transfer to Hot Leg Recirculation
Transfer to Cold Leg Recirculation from Hot Leg
Recirculation
Post-SGTR Cooldown Using Backfill
Post-SGTR Cooldown Usin,g Slowdown
Post-SGTR Cooldown Using Steam Dump
REV. lA
REV. lA
REV. lA
REV. lA
REV. lA
REV. lA
REV. lA
REV. lA
REV. lA
REV. lA
REV. lA
REV. lA
REV. lA
REV. lA
REV. lA
REV. lA
Loss of All AC Power
. REV. lA
Loss of All AC Power Recovery without SI Required REV. lA
Loss of All AC Power Recovery with SI Required
REV. lA
Loss of Emergency Coolant Recirculation
REV. lA
LOCA Outside Containment
REV. lA
Uncontrolled Depressurization of All Steam
Generators
SGTR with Loss of Reactor Coolant - Subcooled
Recovery Desi red.
SGTR with Loss of Reactor Coolant - Saturated
Recovery Desired
SGTR without Pressurizer Pressure Control
Critical Safety Functton Status Trees
Subcriticality (Status Tree)
Core Cooling (Status Tree)
Heat Sink (Status Tree)
Integrity (Status Tree)
Containment (Status Tree)
Inventory (Status Tree)
Response to Inadequate Core Cooling
Response to Degraded Core Cooling
Response to Saturated Core Cooling
Response to Loss ~f Secondary Heat Sink
Response to Steam ~enerator Overpressure
Response to Steam Generator High Level
Response to Loss of Normal Steam Release
Capabilities
REV. lA.
REV. lA
REV. lA
REV. lA
REV. lA
REV. lA
REV. lA
REV. lA
REV. lA
- REV. lA
REV. lA
REV. lA
REV. lA
REV. lA
REV. lA
REV. lA
REV. lA
REV. lA
Appendix A
1-FR-H. 5
1-FR-I.1
1-FR-I.2
1-FR-I. 3
1-FR-P.1
1-FR-P.2
1-FR-S .L
1-FR-S.2
1-FR-Z.1
1-FR-Z.2
1-FR-Z.3
1-FR-Z.4
AP-1. 01
AP-.3. 00
AP-5 .13
2
Steam Generator Low Level
High Pressurizer Level
Low Pressurizer Level
Voids in Reactor Vessel
Imminent Pressurized Thermal Shock
REV. lA
REV. lA
REV. lA
REV. lA
REV. lA
Response to
Response to
Response to
Response to
Response to
Condition
Response to
Condition
Response to
Response to
Response to
Response to
Response to
Response to
Anticipated Pressurized Thermal
Nuclear Power Generation/ATWS
Loss of Core Shutdown
Shock
High Containment Pressure
Containment Flooding
High Containment Radiation Level
Containment Positive Pressure
Rod Control System Malfunction
Control Rod Misalignment
Malfunctioning Individual Rod Position
Indicator (!RPI)
Conditions Requiring Emergency Boration
Nuclear Instrumentation Malfunction
Radiation Monitoring System Process Vent
REV. lA
REV. lA
REV. lA
REV. lA
REV. lA
REV. lA
REV. lA
REV. 00.01
05-18-89
01-28-88
06-23-88
05-29-89
Monitors ALERT/ALARM
Radiation Monitoring
Radiation Monitoring
Monitor
01-12-89
System Liquid Waste Monitor 02-13-87
02-12-87
Radiation Monitoring System CC Monitors A and B*
(RI-CC-105 & RI-CC-106)
02-13-87
Radiation Monitoring System Restricted Control
Area Monitors
Radiation Monitoring System Main Control Room
REV. 00.03
Area Monitor
02-07-89
Radiation Monitoring System Containment *Particulate
and Gas and Manipulator Crane
10-19-89
Radiation Monitoring System Condenser Air Ejector 02-13-87
Radiation Monitoring System SG Slowdown
02-18-88
Radiation Monitoring System Recirc Spray Coolerr
Service Water Outlet
REV. 00.01
Radiation Monitoring System Reactor Coolant
Letdown Monitors
Radiation Monitoring System Discharge Tunnel
Radiation Monitoring System Containment High
Range Area
Radiation Monitoring System Reactor Containment
Area Monitors and CHRRM
04-02-87
02-23-88
02-13-87
10-19-89
Radiation Monitoring System Process Vent
or Process Vent Flow Monitor Malfunction
Loss of RCP Seal Cooling
Monitor,
Station Blackout
08-04-88
REV. 1
REV. 00.03
Appendix A
AP-10. 03
AP-10. 05
Loss of Vital Bus I
Loss of *Vital Bus II
Loss of Vital Bus III
Loss of Vital Bus IV
3
Loss of Semi-Vital Bus 1 and AC Distribution
Loss of D. C. Power
Loss of Auto Load Shed
Partial Loss of Reserve Station Service
Partial Loss of Station Service
,
Service Water System Abnormal Conditions
Loss of Intake Canal
Loss of Component Cooling Water
Excessive RCS Leakage
AP-10. 07
AP-10 .10
AP-10 .11
AP-10 .12
-Auto Start Failure of EOG
AP-21. 00
AP-31. 00
Emergency Diesel Generator Fails to Accept
Electrical Load
Main Control Room Inaccessibility
Loss of Main Feedwater Flow
Response to AFW Check Valve Backleakage
Fuel Handling Abnormal Conditions
Loss of Refueling Cavity Level
Loss of Spent Fuel Pit Level
Minor SG Tube Leakage
Loss of Decay Heat Removal Capability
Increasing or Dec~easing RCS/PRZR Pressure
Control Room Security
Seismic Event
Abnormal Environmental Conditions
Natural Circulation of RCS
Non-Recoverable Loss of Instrument Air
Loss of Reactor Coolant Flow
Personnel Injury - Operations Response
Fire Protection - Operations Response
Loss of Domestic Water
FCA-1.00
Safe Shutdown Area Fire
FCA-1.01
Limiting MCR Fire
\\
08-23-88
08-23-88
08-23-88
08-23-88
07-21-87
04-11-89
01-28-88
REV. 00.03
I 04-14-87
REV. 00.01
REV. 00.00
04-02-87
06-14-89
06-23-88
06-28-89
04-02-87
04-02-87
04-19-88
01-10-89
10-13-88
10-26-88
02-12-87
09-27-88
03-03-87
01-28-88
02-07-89
REV. 00.05
03-03-87
05-30-89
02-12-87
09-11-87
04-21-89
01-08-89
REV. 00.09
REV. 00.05
APPENDIX B
TECHNICAL AND HUMAN FACTORS COMMENTS
This Appendix contains technical and human factors comments and observations.
Unless specifically stated, these comments are not* regulatory requirements.
Howeveri the licensee acknowledged that the factual content of each of these
comments. was correc.t a.s stated.
The 1 i censee. further agreed to eva 1 uate each
comment, to take appropriate *action, and to document that action. These items
will be reviewed during a future NRC inspection.
I. General Comments:
The licensee state.d that they would take corrective actions for Appendix B
comments per the following schedule:
1.
Prior to the exit interview:
Set the steam generator blowdown
radiation monitor alarms at the correct setpoints; AP-24, entry
condition 2.
2 .
Within one week of the exi't interview:
Revise AP-12.01, Loss of
intake canal; step 1 first asterisk, step one second ast:erisk, step
10a, step 15 table left column, and step 17 table right column.
(IFI
280,281/90-09-01)
3.
Within 12 weeks of the exit interview, with Rev. 2 to the EOPs:
Revise 18 steps in 15 procedures, as follows:
ECA-3.2 step 15a,
step 18c, and setpoint k.1; ES-0.3 step 3a; FR-I.2 step 4bl;
FCA-1.01 step 6; AP-21.01 step 4c; E-1 step 6d2 RNO;
step 10c; FR-H.1 step 28 RNO; ES-3.1 Attachment 2 step 16; ECA 0.0
Attachment 4 step 7; ECA-2 .1 step 8b; ES-1. 3 step 2a; ECA-0. 2
step 1 RNO and step 3 caution; E-0 step 4a RNO; AP-1.00 step 8 RNO.
(IFI 280,281/90-09-05).
4.
With Rev. 3 to the EOPs:
Complete corrective actions for all techni-
cal and human factors com"lents in Appendix B.
(IFI 280,281/90-09-07)
In this Appendix,
comments
on
individual procedures are arranged
numerically in the following order:
II. EOP comments:
E-, ES-, ECA-, F- , FR-
III.
AP comments:
AP-
IV. FCA comments:
FCA-
11. EDP comments:
1.
E-0
a.
Reactor trip or safety injection
Step 2c RNO:\\This step required the closure of MSR steam supply
valve, MOV-MS-.1000, which did not have local valve position
indication.
Appendix 8
b.
c.
d.
e.
f.
g.
2
Step 2c RNO:
This step directed the closure of MSR steam supply
valves.
Two of those valves, MOV-MS-lOOA and MOV-MS~lOOB, had
labels with only the valve numbers.
Step 2d:
This step directed the verification of the generator
output breakers in the open position. The corresponding local
indication was
11trip.
11
Step 3b RNO:
This step failed to address action to be taken if
the EOG did hot start 1 fol lowing step 3b2 R
1NO. i
PSTG DEV, Step 4a RNO:
This st~p failed to iist low PRZR level
as an i ndi cation which required SI.
This was not properly
addressed in the SOD,
Step 13b RNO:
This step failecj to direct that both SI reset
push buttons may be required to be pushed in order to reset SI.
Step 32b:
This step failed to specify the indications to be
addressed in order to check
11 radiatiori -
NORMAL.
11
h.
Step 34a: This step failed to specify which pressure was to be
checked .
. i.
Step 34b:
Same as step 34a above.
j.
Step 35a RNO:
This step directed the energizing of PRZR heaters
from the AC emergency buses.
PRZR heater groups 8, C, and D ~ere
not on the AC emergency buses.
k.
Step 35a RNO:
This step failed to direct that the energizing
of the semi-vital bus be conducted in accordance with AP-10.05 *
step 10.
1.
Continuqus action page:
Same as step 4a RNO above.
2.
E-1
Loss of reactor or secondary coolant
a.
Step 1 caution: This taution contained an action statement and
did not indicate the consequences of nqt complying with the
caution.
b.
Step 1 third note:
The note contained an action statement.
C.
d.
Step 3 caution: This caution did not indicate the consequences
of not complying wi~h the caution.
Step 3a RNO:
The adverse feed flow value of 492 GPM was too
detailed a value to use, since it was the total of up to three
flow indicators.
\\
Appendix B
3
e.
PSTG DEV, Step 6d2 RNO:
This step re qui red the operator to
11 Continue with Step 8.
WHEN PRZR level greater than 11 percent,
THEN close PRZR spray valves as necessary.
11
The corresponding
step in the ERG stated
11 00 NOT STOP SI PUMPS. Try to stabi 1 i ze
RCS pressure with norma 1 PRZR spray.
GO TO STEP 8.
11
The EDP
SOD did not address this deviation.
It was not clear to the
procedure writer if Step 7, which was SI Termination, would be
performed after the Step 6d2 RNO was completed.
f.
Step 10 caution~
Same as itep 1 caution above.
g.
Step 10a RNO:
This step required the operator to
11 GO TO Step
14.
WHEN pressure less than 14 psia, THEN do Steps 10 through
14.
11
It was not clear to the procedure writer why this step was
worded differently than step 6d2 RNO above.
h.
Steps 10d and e:
As written, these steps could have cawsed the
operator to isolate the A SW header in RNO 10d without insuring
that the B SW header was operational.
i.
Step 11:
The location of infrequently operated equipment was
not indicated.
j.
Step 13a RNO:
Same as Step 10a RNO above.
k.
Step 13b:
The previous step that reset CLS, step 10, could have
been bypassed as a result of step 9b, which stated "GO TO Step
13 .
11
There was no RNO step to "Reset CLS" if the response to
the action step to
11 Verify CLS reset
11 was negative.
1.
Step 14 caution:
Same as step 1 caution above.
m.
Step 16a RNO:
The CC pumps were not included on the list of
equipment that should be loaded on the AC emergency busses.
n.
Step 18a2:
The listed equipment was not organized to expedite
verification of power availability.
~
o.
Step 22cl:
There were no permanent position indicators on the
Key switches for the SI Accumulator isolation valves.
p.
Step 23 caution:
The consequences of not complying with the
caution were not included in the caution statement.
q.
Step 25:
The action column directed the verification of several
valve positions, but there were no RNO actions if the valves
were not in the desired positions .
Appendix B
4
3.
E-2
Faulted steam generator isolation
a.
Step 1:
This step did not identify bypass valve indication as
requiring local verification.
b.
Step 1 RNO:
This step identified steam generator bypass valves
as requiring manual
closure.
These valves required local
closure.
c.
Step 2:
This step was w~itten as a step with a single substep.
This was not consiste~t with the WG.
d.
Step 3a RNO:
This step indicated that there was no preferred
order for performance of the substeps, although it was clear
that an inspection of secondary lines would not be attempted
until the other items had been inspected.
e.
Step 3a RNO sixth bullet: This bullet required that decay heat
release lines be inspected to determine that an initiating break
had occurred.
Due to their location, it would not be possible
to inspect these lines after a break.
f.
Step 4a RNO:
This step *was actually the RNO for step 4al only.
g.
Step 4d:
The verb
11 veri fy
11 was used to indicate that this step
should be accomplished, if possible.
This was not consistent
w i th the de f i n i t i o n o f II v e r if y II i n the WG .
' h.
Step 5 RNO:
ThE step required the operator to transfer to
alternate AFW water supply without specifying the alternate
water supplies or the preferred sequence for the selection of
the source.
i.
Step 6a:
This step required a consultation with the TSC.
It was unlikely that the TSC would be manned at this point in
the procedure.
j.
Attachment 1 step 3a first and second bullets:
These actions
were preceded by an asterisk instead of a bullet, which was
inconsistent with the WG.
k.
Attachment 1 step 3b first and second bullets:
These actions
were preceded by an asterisk instead of a bullet, which was-
inconsistent with the WG.
1.
Attachment 1 step 3b
RNO first and second bul 1 ets:
These
actions were preceded by an asterisk instead of a bullet, which
was inconsistent with the WG.
Appendix B
4.
m.
n.
0.
p.
q.
r.
5
Attachment 1 step 4a first through sixth bullets: These actions
were preceded by an asterisk in stead of a bull et, which was
inconsistent with the WG.
Attachment 1 step 4b:
Keys 66, 67, and 68 were required by this
step, but were not identified.
Attachment 1 step 4d first through sixth bullets: These actions
were preceded by an asterisk and a place keeping mark instead of
a bullet, which was inconsistent with the WG.
Attachment.I step 4d first through third bullets: These actions
were preceded by an asterisk instead of a bullet, which was
inconsistent with the WG.
Attachment 1 step 5 RNO:
This step was not preceded by an
asterisk, although it was a step that would be performed at a
later time when conditions permitted.
Attachment 1 step 7 first and second bullets:
These actions
were preceded by an asterisk instead of a bullet, which was
inconsistent with the WG .
E-3
Steam generator tube rupture
a.
Entry conditions:
Asterisks were used in a manner inconsistent
with the WG.
b.
Step 3 caution:
The two items in this caution were written
indicating actions of a continuous nature.
Consequence was not
included as part of the caution.
c.
Step 4 caution:
This caution did not contain a statement of
consequences.
The
item was written implying a continuous
action.
d.
Step 4:
This step was written as a continuous action as denoted
by an asterisk. There was no formal method to aid operators in
remembering all of the open continuous action steps.
e.
Step 5 Caution:
This caution was written as an action with
conditional logic.
Also no consequence was stated in the
caution.
f.
Step 5:
Same as Step 4 above .
\\
Appendix B
g.
h.
6
Step Sa* RNO:
The breakers were i dent ifi ed with labels that
appeared to be pink rather than purple as was designated for the
proper color coding.
The operator on the walkdown agreed there
was an inconsistency in the color coding.
Access to one breaker
fequired entering a restri~tively narrow space* between a
contaminated area and the breaker boxes.
Due to the lighting
conditions, a flashlight was required to read the label on this
breaker.
Ste~ 7 caution:
This caution did not contain a statement of
~onsequence as per the WG.
i.
Step 7:
Same as Step 4 above.
j.
Step 9b:
In this step an action step was imbedded as a substep
under a check. This may increase the potential for the operator
to overlook the action.
k.
Step lOal RNO:
Performan~e of this action required o~eration of
controls approximately 15 to 20 feet above the floor.
The
operator on the walkdown indicated they would be accessed by
climbing pipes.
No ladder was evident in the area.
1.
Step 12 caution:
A continuous action and a conditional action
were contained in this caution .. This caution did not contain a
statement of consequence as per the WG.
m.
Step 15 note:
The first two items in this caution contained
conditional actions.
This caution did not contain a statement
of consequence as per the WG.
n.
Step 15a:
This step used the phrase ONE TIME which was not
sufficiently explained in the step and which was not defined in
the WG.
o.
Step 15b
RNO first and . second bull et:
This step provided
the operator with two options for action but did not provide
direction regarding preferred action.
The operator on the
walkdown indicated he believed there was a preferred action.
p.
Step 15b3 RNO:
This step required the operator *to stand on a
pipe to operate a large valve.
The valve was reverse seated.*
The operator on the wal kdown said he was trained on wh*ich
direction to turn the valve hand wheel for the desired result
but the valve was not labeled as such locally or referred to as
such in the procedure.
q.
Step 15b RNO third and forth bullets:
Same as Step 15b RNO
first and second bullets .
Appendix B
7
r.
Step 16 caution:
This caution did not contain a statement of
consequence as per the WG.
s.
Step 20a RNO:
This step required the operator to monitor four
parameters but provided no specific direction as to what changes
to look for or expect.
t.
Step 27: Same as Step 4 above.
u.
Step 28c2:
This step required the operator to locally throttle
a valve to 25 percent open.
No formal indication of valve
position was provided locally requ,_r,ng operator judgment to
determine an intermediate valve position.
v.
Step 30b7:
This step required the operator to verify that a
control valve was controlling temperature but did not ~rovide a
specific description of what indications to use to make this
verification.
w.
Step 31a RNO:
This step used IAW as a referencing term but IAW
is not defined as a referencing term in the WG.
x.
Step 33 caution: This caution implied a continuous action.
The
caution did not contain a statement of consequence as per the
WG.
y.
Step 33a:
This step contained a table in which some en.tries
were marked with asterisks indicating continuous action steps.
The operator on the walkdown indicated that he felt an asterisk
should be used with the step number.
z.
Step 33a:
This step contained items directing the operator to
raise flow, lower flow, depressurize,
and maintain pressure,
however provided no specific information concerning how to or
how much to change the parameters.
aa.
Step 34b:
This step contained two references but used no
referencing terms as ~escribed in the WG.
ab.
Step 35a RNO first and second bullets:
This step provided
the operator with two options for action but did not provide
specific direction regarding preferred or appropriate action.
The operator on the walkdown indicated that the appropriate
action was determined based on the operating status of the other
unit.
Appendix 8
8
ac.
Step 35a RNO first bullet:
This step directed the performance
of a relatively complicated local action requiring several steps
and communication between the two units outside the control
room.
No specific direction or reference was provided for
performance of the action.
The needed continuous communication
between NLOs on units one and two required radios, but that was
not specified in the procedure.
A gaitronics phone was located
in the area of the locally operated component on unit one but
was well out of reach from the component.
ad.
Step 35a RNO second bullet:
Same as Step 31a RNO.
ae.
Step 37a RNO:
This step required an operator to adjust a flow
controller but did not provide a specific value as per the WG.
af.
Step 38a2a RNO:
Same as step 31a RNO.
ag.
Step 39 RNO:
This step directed the operator to "Dump more
steam" but provided no specific direction on by what means or
how much.
ah.
Step 39:
The operator on the walkdown indicated that this step
would probably involve trending several parameters by hand.
The
procedure provided no space, attachment, or blank page for this.
ai.
Step 39 fifth bullet: This step used "based on" as a referenc-
ing term but this term was not specified for referencing in the
WG.
aj.
Step 41 seventh bullet:
This step used the acronym RSST which was
not listed in the WG acronym attachment.
ak.
Step 42:
This step provided the operator with three options for
branching to another procedure but provided no guidance on
preferred procedure or how to decide.
The operator on the
walkdown indicated that the TCS or SS would make the decision.
The procedure provided no reference to the TSC.
al.
Attachment 1 step 3, 4, and 7:
These steps used asterisks in a
manner inconsistent with the WG.
am.
Attachment 1 step 8: Same as Step 31a RNO.
an.
Attachment 2 step 6:
This step required local operation of two
valves in .the turbine deck dog house.
The label tag for one of
the valves was inaccessible under recently installed insulation
or was missing.
Hot pipes and limited space caused apparent
difficulty for operator access and operation of the valves .
Appendix B
9
ao.
Attachment 2 step 7:
This step required local operation of
three valves located high above walkways, piping and openings
- to the level below.
There was very limited room for ladder
placement and no apparent ladders in the area.
The operator on
the walkdown indicated he would climb nearby pipes and jump
across to piping near the valves.
A second operator on a
walkdown indicated he would obtain a safety belt to operate the
valves but only if he had time.
- ap.
Step 5 continuous action page:
This step directed the operator
to transfer to alternate AFW water supply but provided no
specific steps or reference for performing the action.
The
operator on the walkdown indicated the action could involve
local operation of a valve and operation of a valve from the
control room or as a last resort the use of FW.
aq.
Continuous actions page:
The
heading for this page was
not underlined and asterisks were used, in violation of \\'.JG
direction.
5,
Rediagnosis
a ..
Step 1:
This step was not written in the *form of a single step,
and was not consistent with the WG.
b.
Step 1 RNO first and second sub-bullets:
These steps indicated
that there was no preferred order for performance of the sub-
steps, although it was clear that the steps were dependent upon
the status of main steam line isolation.
c.
Step 3 second bullet:
The annunciator window for this step read
11Unit 1 AUX STM RAD ALERT /HI
11 *
This annunciator window was for
steam generator high radiation level although the annunciator
window did not refer to steam generator radiation level in the
engraving.
6.
Reactor trip response
a.
Step 6a RNO:
This step directed action which required that
BC-227
be
verified
closed per OP-50.1,
REMOVAL
OF
THE
BEARING COOLING WATER SYSTEM FROM SERVICE.
The operator stated
that he would reach valve BC-227 by climbing on various
equipment and piping.
The valve was located approximately 20
feet above the fl oar, and no appropriate 1 adder was readily
accessible.
b.
Step 6a RNO:
This step directed action which required that the
11 fire water to air compressor
11
valve be opened per OP-50.1,
REMOVAL OF THE BEARING COOLING WATER SYSTEM FROM SERVICE.
The
step did not, specify FP-170 but incorrectly stated BC-170.
'
.
Appendix 8
10
c.
Step 7a
RNO:
This step required verification of letdown
isolation.
No specific direction was provided for performance
of this step. This was inconsistent with the list of applicable
valves provided in step 7c.
d.
Step 8b RNO:
This step directed the
11defeat
11 of TAVG and
delta T control for the affected loop.
This terminology was
inconsistent with the switches used to perform the action (TAVE
e.
Step 8b RNO:
This step directed the use of auxiliary spray.
No
specific information was provided for performance of the step
(Open HCV-1311).
f .. Step 9c: This step was not properly identified as a continuous
action step.
g.
Step 17: This step directed the local closure of the 1st point
extraction steam. bre~ker, 181-343. * The normal position of
181-343 was
11 closed. 11
h.
Step 19b: This step directed the opening of AOV-CP-122 and the
closing of MOV-CP-100 in the condensate polishing system.* Prior
to this step in the procedure, the position of these valves was
dependent on the initial plant conditions. This step failed to
adequatel~ address the required action~ for all possible initial
valve positions.
i.
Step 21a2 RNO:
This step failed to clearly define communica-
tions in that it required consultation with
11 plant staff 11 *
7.
Natural tirculation Cooldown
a.
Cov_er Page Head~r:
The company name on the procedure did not
comply with the WG Section 6.2.1.a.l.
b.
~ntry Conditions:
Transition procedures were denoted with
asterisks instead of bullets as required by the WG.
c.
Step 1 Caut.ion:
There was inappropriate use of
11 should
11 instead*
of
11 shall,
11 i.e., if there was SI actuation, it would be
mandatory to implement procedure E-0.
d.
Step 6a RNO:
This step required that a tota 1 AFW fl ow of 492
gpm be read from a combination of gages that had small scale
units of 7 gpm.
These aw~ward scale units made the determina-
tion of feed flow to this precision impractical.
e.
Step 68 RNO:
The grouping of bullets below this step did not
have the word
11 or
11 between them.
Appendix B
11
f.
Step 14:
Attachment 3 was referenced when attachments 1 and.2
had not yet appeared in the procedure.
g.
Step 23c:
The overpressure mitigation key switches did not have
a label for
11enable
11 or
11disable
11 positions.
8.
Natural circulation cooldown with steam void in RX vessel
a.
Step 1 first caution:
The verb
11 implemented
11 was used in the*
caution statement but'it did not appear in the WG.
b.
Step 1 second caution:
The first 14 steps of ES-0.2 must be
performed prior to th~ beginning of this procedure in order for
this procedure to work.
The note did not identify these actions
as mandatory.
c.
Step 1 first note:
This note indicated that it- was not
mandatory to monitor the continuous action page.
This was not*
in accordance with the intent of the WOG guidelines.
d.
Step 2: This step was not preceded by an asterisk, although it
was a step that would be performed continuously throughout the
procedure.
e.
Step 3a: This step had an incorrectly referenced cooldown rate.
The procedure used the value of 100 degrees per hour, but the
setpoint docum~nt required a 50 degree per hour cool down rate.
For this step, the ERG used a cooldown rate of 100 degrees per
hour.
e.
Step 4b2 RNO first and second bullets: These steps were _listed
as having no preferred sequence,
This was inconsistent with
operator training.
f.
Step 6c RNO:
This step indicated that IA should be available.
The step did not specify containment IA.
g.
Step 6cl:
This step required keys to accomplish the step, but
did not identify the required keys (Keys 11, 12 and 13).
h.
PSTG DEV, General:
The licensee had no procedure ES-0.. 4,
Natural circulation cooldown with steam void in vessel without
The SOD justification for this was incomplete, in that
it did not consider RVLIS failure during or as a result of a
casualty to the plant that required a natural circµlation
cool down.
\\
Appendix B
12
9.
SI Termination
a.
PSTG DEV, Step 7: This step stated II ISOLATE HHSI TO COLD LEGS.
11
The corresponding step in the ERG stated II Isolate BIT.
11
The EOP
SOD did not address this deviation.
b.
PSTG DEV, Step 8b RNO:
This stp stated
11 IF PRZR level can
NOT be maintained greater than llJ.;, THEN do the following:
1.
Align SI fl ow path
2.
GO TO ES-1. 2 POST LOCA COOLDOWN
AND DEPRESSURIZA TION, STEP 1
11 *
The corresponding step in the
ERG, which was Step llb RNO, stated
11 If PRZR
level can not be
maintained THEN manually operate SI pumps as necessary.
GO TO
E-1,
LOSS
OF
REACTOR
SECONDARY
COOLANT,
Step
1.
11
The EOP SOD did not address this deviation.
c.
Step llcl RNO:
The indicator for RCP seal leakoff temperature
was not in this step, but was needed to insure that the
operators use the correct indicator.
d.
PSTG DEV, Step 12:
The sequence of steps in the EOP had been
changed and as written allowed the performance of ERG step 17
prior to the completion of ERG steps 13 through 15.
This was
contrary to the ERG required step sequence and the EOP SOD did
not address this deviation.
e.
Step 19a RNO:
Same as Step llcl RNO above.
f.
Step 20
RNO:
The
CC pumps were missing from the list of
equipment to be started when offsite power was restored, because
Step 19b directed the operator to this step if CC pumps were not
running.
g.
Step 21 caution:
This caution contained an implied action
statement and did not indicate the consequence of not complying
~ith the caution.
h.
Step 26:
The location of infrequently operated equipment was
not indicated in the EOP.
i.
Step 34: The same as step 26 above.
10.
Post LOCA cooldown and depressurization
a.
Entry conditions: Asterisks were used in a manner incon~istent
with the WG.
b.
Step 1 caution:
This caution contained a conditional logic
action and a transition and contained
no
statement of
consequences and was thus inconsistent with the WG .
Appendix B
13
c.
Step 2a
RNO:
This step indicated a continuous action but
was not identified by an asterisk, nor did it appear on the
continuous action page.
The word continue was used to indicate
transition and continuous action but was not defined as such in
the WG.
d.
Step 3:
The operator on the walkdown indicated that items in
this step were performed in parallel.
There was no indication
of this in the procedure.
e.
Step 3b RNO:
This step used IAW as a referencing term but IAW
was not defined as a referencing term in the WG.
f.
Step 5:
This step was indicated as a continuous action by an
asterisk.
There was no formal method to aid the operator in
remembering all of the open continuous action steps.
g.
Step 5 RNO:
This step used the condit i ona 1 statement IF
necessary, THEN ... but provided no specific guidance on-how to
determine if necessary.
h.
Step 6 caution:
This caution contained a conditional logic
action and contained no statement of consequences and was thus
inconsistent with the WG.
i.
Step 6:
Same as step 5.
j.
Step 7 caution:
This caution contained no statement of
consequence as per the WG.
k.
Step 7.
This step was indicated as a continuous action by an
asterisk.
There was no forma 1 method to aid the operator in
remembering all of the open continuous action steps.
1.
Step 7 items a and b:
The SG narrow range level indicators
required for this step had red and yellow lines scribed
horizontally across the face of the i ndi ca tor.
These 1 i nes
were not referred to in the procedure and the operator on the
walkdown did not know of any purpose for them.
m.
Step 8 note:
This note contained a continuous action and two
conditional logic actions and was thus inconsistent with the WG.
n.
Step 13d: Same as Step 3b RNO:
o.
Step 16e:
This step required the operator to read 25 gpm or
greater on a charging flow indicator.
The scale at the bottom
of the indicator was compressed and did not have a mark or clear
indication for 25.
The operator on the walkdown indicated he
controlled flow to the'\\?ext mark up which he believed was
40 gpm.
.
Appendix B
-
14
p.
Step 17 note: This note required the operator to determine when
RCS pressure stabilized but provided no direction on how to
determine this. The operator on the walkdown indicated he would
use judgment based on watching it for four or five minutes.
Stabilize was not defined in the WG.
q.
Step 19a RNO: Same as Step 3b RNO.
s.
Step 20:
Same as Step 7.
t.
Step 22a:
Same as Step 3b RNO.
u.
Step 22c RNO:
This step required the operator to
11Borate
11 but
provided ho specific information on how much or how.
v.
Step 24c RNO:
Same as Step 3b RNO.
w.
Step 25b:
This step referenced procedures but did not use
referencing terms as directed by the WG.
x.
Step 29 last bullet:
The operator on the walkdown indicated
this was a local action. It was not indicated as such in the
procedure.
y.
Step 33 RNO:
This step required* the operator to return to
Step 7.
An operator in an interview indicated he believed it
would be more technically correct to return to Step 6.
11.
ES-1.3 Transfer to Cold Leg Recirculation
a.
PSTG DEV, Step 1 Caution:
The ERG caution for loss of offsite
power following SI re.set was not included.
This was not
adequately justified in the SOD.
b.
Step 1 Note:
The ERG note to remind the operator that the
foldout page for the E-1 series should be open was not included.
This was not adequately justified i~ the SOD.
c.
Step 2a:
This st~p established fl ow to the RSHX s but did not
give the opera.tor a value for minimum or maximum acceptable
flows.
The minimum flow was identified as 2400 GPM in the
setpoint document.
d.
Step 4 Caution:
This caution did not list the value for the
RWST empty alarm.
e.
Step 4:
This step did not verify that the charging pump
mini fl ow valves were shut.
The consequences of these valves
being open could be injection of radioactive recirculation water
into sections of the let~own system, VCT, etc.
Appendix B
12 .
15
f.
Step 4c4:
This step did not include a caution for the operator
to monitor charging pump discharge flows when isolating the CHG
pump RWST suction valves.
g.
Step 4 RNO:
This step did not direct operators to operate
valves locally and there was no caution for potential highly
radioactive fluids being present during local operations.
h.
Step 3 RNO:
This step did not require the operator to start a
LHSI pump if at least one LHSI pump was not running.
i.
Step 4b RNO:
These steps were listed as having no preferred
sequence.
This was inconsistent with operator training.
j.
Step 5d and Se:
There was no RNO column directing the operator
to locally close the valves if they failed to close from the
control room.
k.
General:
This procedure did not identify the m1n1mum contain-
ment sump level for running pumps on the reci rcul at ion sump.
This value was identified as 2.5 ft. in the setpoint document.
ES-1.4 Transfer to hot leg recirculation
a.
Entry cond.itions:
Asterisks were used in a manner inconsistent
with the WG.
b.
Step 2 caution:
This caution contained no statement of
consequences and thus was inconsistent with the WG.
c.
Step 5:
This step directed the operator to RETURN TO procedure
and step in effect".
In the WG
11 RETURN
11 TO is specified for use
in transitions but not for branching.
13.
ES-1.5 Transfer to cold leg recirculation from hot leg recirculation
a.
Cover page header:
The company name of the procedure did not
comply with the WG Section 6.2.1.a.l
b.
(NQ) was not in the WG list of abbreviations.
14. ES-3.l
Post-SGTR cooldown using backfill
a.
Cover page header:
The company name on the procedure did not
comply with the WG Section 6.2.1.a.l.
b.
Entry conditions:
Transition procedures were denoted with
asterisks instead of bullets as required by the WG.
c.
Step 1 Note:
There was inappropriate use of
11 should
11 instead of
"shall"; it was mandatory to use continuous action pages.
Appendix B
d.
e.
f.
g.
h.
i .
j.
k.
16
Step 2cl:
The key numbers were not designated for working the
SI accumulator key switches.
Step 4a RNO:
This step required that a tota 1 AFW fl ow of 492
gpm be read from a combination of gages that had
sma 11 seal e
units of 7 gpm.
These awkward scale units made the determina- *
tion of feed flow to this precision impractical.
Step 4b RNO:
The grouping of bullets below this step did not
have the wording
11 or
11 between theni.
Step 10a:
This step did not have an adverse containment
pressure value.
Step lOc:
The overpressure mitigation key switches did not have
a label for
11 enable
11 or
11disable
11 positions.
Step 13a Second Bullet:
The recorders for RCP leakoff flow Hi
and Lo Range, recorders FR-1-154A and FR-1-154B, referred to
reactor coolant pumps as 1, 2, and 3, instead of A, B, and C.
Attachment 2:
Asterisks were used in place of bullets, which
did not conform to the WG .
Attachment 2 Step lb:
There were no numbers provided for the
keys to be used in the SG blowdown permissive key switches, and
the switches did not have any labeling to indicate position.
15.
Post-STGR cooldown using blowdown
a.
Step 4a RNO:
The adverse feed fl ow va 1 ue of 492 GPM was too
detailed a value to use, since it was the total of up to three
flow indications.
b.
Step 4:
The location of infrequently operated equipment was not
specified.
c.
Step 10c:
This step directed the operator to Operating
Procedure 1-0P-32.1,
PLACING
SLOWDOWN
COOLING
IN
SERVICE.
This procedure ca 11 ed for several system lineups/checkoffs as
prerequisites for entering it.
The procedure, as written,
didn't support rapid implementation during a casualty.
d.
Step 14c:
There were no permanent position indicators on the
Overpressure Mitigation system key switches and the key number
was not included in the procedure.
Appendix B
17
16.
Post-SGTR Cooldown Using Steam Dump
a.
Step 2b:
The values for minimum RCS subcooling based on CETCs
did not include the natural circulation values.
b.
Step Sa:
The values for maximum cooldown rate in the RCS cold
legs did not include natural circulation values.
c.
Step Sa:
Cooldown .of the RCS must also be maintained in
accordance with p 1 ant TS pressure/temperature curves but no
reference was made for their use.
d.
PSTG DEV, Step 7:
This step required the operator to initiate
steps in Attachment 7.
These steps were plant specific and not
in the ERG.
They included taking the loop isolation valves off
their backseats, which operators estimated would take a licensed
operator at least 30 minutes and possibly over an hour to
comp 1 ete.
The performance of these steps would de 1 ay the
cooldown.
This potential delay was no~ adequately justified in
the SOD.
- e.
Step 9:
The -SG narrow range 1 eve 1 was 1 i sted as 25 percent
(32 percent) and was listed as 97 percent (32 percent) in other
parts of this procedure.
f.
PSTG DEV, Step 10:
This step required the operator to reduce
the ruptured SGs pressure by 100 psig. This was contrary to the
ERG which stated that steam should be released slowly from the
ruptured SG_to avoid a rapid decrease in pressure and sub5equent
reinitiation of break flow.
This was not justified in the SOD.
g.
Step* 12d:
Same as Step 2b.
h.
Step 14a:
The
RCS pressure for p 1 acing the overp_ressure
.mitigation system in service did not list the adverse contain-
ment value.
i.
Step 14c:
The overpressure mitigation key switches did not have
a label for the
11 enable
11 or "disable." positions.
j.
PSTG DEV, Step 14: This step placed the overpressure mitigation
system in service and was contrary to ERG steps.
The
SOD
justification for adding this step was inadequate.
It did not
address the potential actuation of the OPM system and subsequent
flow of unborated water from the SG into the RCS.
k.
Step 16a:
Same as Step Sa.
1.
Step 16c RNO:
This step required the operator to dump steam at
the "maximum rate
11 from *,he intact SGs.
The intent of this ERG
step was to dump steam at. a controlled rate. This step devia-
tion from the ERG was not justified in the SOD.
Appendix B
18
m.
Step 17:
This step incorrectly indicated that there was no
preferred order for performance.
n .. Cover page header:
The company name on this procedure did not
comply with the WG section 6.2.1.a.l.
o.
PSTG DEV, Step 1 Note:
This note indicated that it was not
mandatory to monitor the continuous action page.
This was not
in accordance with the intent of the ERG.
p.
Step 2c:
This step re qui red keys to accomplish the step, but
did not identify the required keys.
q.
Step 4a RNO:
This step required that a total feed flow of 492
GPM be read from a combination of gauges that had sma 11 seal e
units of 7 GPM.
The awkward scale units of FI-FW-200 A, B,
and C, made the determination of feed flow to this precision
impractical.
r.
Step 4b RNO:
The groupings of bullets below this step did not
have the wording
11 or
11 between them.
s.
Step Sb RNO:
These steps were listed as having no preferred
sequence.
This was inconsistent with operator training.
t.
Step Sb RNO:
There was no method for locally opening the PORVs
although the step required this action.
u.
Step Sb RNO third and fourth bullets: These steps were listed
as having no preferred sequence.
This was inconsistent with
operator training.
v.
Step 10:
Consideration of potential radiological
release
wa.s not reevaluated prior to initiation of this step, and
appropriate precautions were not identified in this step.
w.
Step 10 RNO:
The step did not direct the operator to manually
operate the ruptured SG PORV to decrease SG pressures.
x.
Step 14c:
The step did not identify the keys required to enable
the overpressure mitigation system.
y.
Step 16c RNO:
Same as Step Sb RNO.
z.
Step 16c RNO Thi rd and Fourth Bullets: Same as Step Sb RNO,
Third and Fourth Bullets.
aa.
Attachment 2 Steps la, le and ld:
These items were preceded by
asterisks instead of bullets.
\\,
Appendix B
19
ab.
Attachment 2 Step la, le, and ld:
The appropriate SFs were not
identified with the listed valves.
ac.
Step 13 Second Bull et:
The recorders for RCP lea koff fl ow HI
and LO range, recorders FR-1-154 and FR-1-154B, referred to
reactor coolant pumps as 1, 2, and 3, instead of a, b, and c.
ad.
Attachment 1 Step la:
This step called for using one of the
P-250 Analog trend recorders and the vertical board had a label
process computer trend recorder.
ae.
Attachment 2 Step lb:
Steam generator blowdown permissive key
switches did not have a position indication label.
They were
marked with a magic marker.
af.
Entry condition: These items were preceded by an asterisk
instead of bullets.
17.
ECA-0.0
Loss of all AC power
a.
Cover page header:
The company name on the procedure did not
comply with the WG Section 6.2.1.a.1.
b.
Step
read
gpm.
flow
4:
This step required that a total AFW flow of 492 gpm be
from a combination of gages that had small scale units of 7
These awkward scale units made the determination of feed
to this precision impractical.
c.
Step 5 Caution: There was inappropriate use of II shoul d
11 in stead
of
11 shal1
11 , i.e. an evaluation had to be made prior to taking
any action affecting no. 3 EOG.
d.
Step 6 Caution: There was inappropriate use of
11 should
11 instead
of
11 shall
11 , i.e. if an SI signal existed, it must be reset.
e.
Step 7a RNO:
In this step, if the other unit had AC power from
its diesel, operators were directed to cross tie power from
that unit's H bus to its J bus to close the condenser waterbox
valves on the affected unit.
This would take much longer than
70 seconds, thus per design calculation ME-0166, canal level
would be lost to below that needed to handle a LOCA.
The
licensee stated that the design for power to the condenser
waterbox va 1 ves was as fo 11 ows:
There were a total of eight
large eight foot diameter pipes supplying water by gravity flow
from the intake canal to the condenser waterboxes, four on unit
1 and four on unit 2.
Each pipe had two motor operated flow
i so 1 at ion valves, one on the condenser in 1 et and one on the
condenser out 1 et.
Eight of these va 1 ves ( four on unit 1 and
four on unit 2) were powered from either the unit 1 J bus or
the unit 2 J bus through an ABT.
Thus if either unit's J bus
had power, all eight valves would close automatically (when
intake canal level went below 23.5 feet) and this large flow of
water from the intake canal
could be
stopped within 70
seconds.
-
.
Appendix B
f.
g.
h.
i.
j.
20
The number three diesel supplied power to the J bus of either
unit, seeking the first one with no voltage. However, the power
supply arrangement for the other eight valves was substantially
different. The number 1 diesel supplied power to the unit 1 H
bus, and number 2 diesel supplied unit 2 H bus.
Four unit 1
waterbox valves were powered from the unit 1 H bus, and four
unit 2 waterbox valves were powered from the unit 2 H bus.
Thus; if only one unit's H bus had power, the waterbox valves
on the other unit wou.ld remain open .until they were closed in
this step.
There was no ABT from the two units 1 H buses to
close all eight waterbox valves within 70 seconds:
Step 7b:
This step did not include locally closing the
c;:irc water isolation valves.
If electrical power was not
obtained to close these valves, operators stated that they would
have difficulty in closing the circ water isolation valves locally
within one hour and intake canal level would likely be lost.
These were eight foot diameter valves.
Local closing required
an operator to manally turn a handwheel 3600 times for each valve.
The valves were not equipped with mechanical aids to assist the
operators in closing them quickly.
Step 10:
This step directed operators to cross tie charging
from the other unit to the affected unit to supply seal water
to the *RcPs.
However, the licensee was not well prepared to
cross tie charging in a timely manner; due to poor labeling and
accessibility of some valves (see FCA-1.00 Attachment 78). The
licensee would take substantially longer than 30 minutes to
establish RCP seal cooling.
Attachment 1 part 1 Step 2:
This step i dent ifi ed that pumps
2-CS-P-lA (breaker no 24J-5) and 2-RS-P-lA (breaker no. 24J-4)
were the A pumps but labels showed that these were the B pumps.
There was incorrect identification in this step.
Attachment 4:
Asterisks were used in place of bullets which ~as
inconsistent with the WG.
Attachment 4 Step 7:
This step required the unlocking of valves
- 1-CH-728 and 2~cH-447.
During the walkthrough the valves were
found to be neither locked nor chained.
The administrative lock
log did not require these valves to be locked.
18.
ECA-0.1
Loss of all AC power recovery without SI required
a.
Step 1 caution: This caution contained an action statemerit and
did not indicate the consequences of not complying with the
caution.
b.
Step 1 note:
This note contained an implied action statement.
c.
Step 1: The location of infrequently operated equipment was not
indicated.
Appendix B
d ..
e.
f.
21
Step 2 first caution:
The meter indication that provided the
correct indication of this value was missing in this step~
Step 2 second caution:
The referenced indicator scale was in
megawatts, but the value in the caution was given in KW.
Step 4e:
The action specified in this step would be difficult
to perform, because the meter had only three graduation marks
below 40 GPM.
g.
Step 6:
Same as step 1 above.
h .. Step 8 first caution:
This caution contained an action state-
ment and did not indicate the consequences of not complying with
the caution.
i .
j.
k.
1.
m.
n.
0.
p.
q.
r.
s.
Step 8 note:
This note contained an action statement.
Step 8:
Same.as step 1 above.
Step 11:
Same as step 1 above.
Step 12 first caJtion: This caution contained an action state-
ment and used the term
11 slowly
11 which was ambiguous and not in
accordance with the WG.
Step 12 s~cond caution:
Same as step 12 first caution above.
Step 12 third caution: This caution contained an implied action
statement.
Step 14bl:
Same as step 4e above.
Attachment 1 Part 1 Step 2:
The pump breakers were l-CS-P-18
and l-RS-P-18 vice 1-CS-P-lA and 1-RS-P-lA and the MCC lJl-1
Supply breaker was 14Jl-6 vice 14J-16.
Attachment 1 Part 2 Step 3:
The trip fuses were not normally
r~moved and that part of the acti~n step was not applicable.
Attachment 1 Part 2 Steps 5 and 7:
The Sync switch in
these steps did not exist and the steps were, therefore, not
applicable.
Attachment 1 Part 3:
This part contained action to restore
power to some equipment when the lJ Bus was returned to normal,
but there was no direction about what to do with the breakers
placed in PTL in Part 1, Step 1 .
\\
Appendix B
22
19.
ECA-0.2
Loss of all AC power recovery with SI required
a.
Step 1 RNO:
This step directed establishing cold leg recirc
valve alignment for LHSI and HHSI suction from the sump when
RWST level was less than or equal to 22 percent. This was prior
to.the automatic shift of suction.
The operator stated that he
would defer performance of this step to the automatic shift of
suction, and he would perform the step if the automatic shift of
suction did not occur at its proper RWST level setpoint.
b.
Step 1 RNO:
This step used paragraph designations
11 a
11 and
11b
11
in the RNO column without the corresponding paragraphs in the
ACTION/EXPECTED
RESPONSE column.
c.
Step 1 RNO:
This step directed the closing of LHSI recirc
valves. The LHSI recirc valves 1 switches were not arranged in an
orderly fashion on the MCR 1 s control board.
d.
Step 3 caution:
This step stated that energized emergency bus
load should not exceed 1200 amps. The MCR emergency bus current*
meter indications read 0-800 amperes with the aforementioned
limit off-scale.
e.
Step 3b RNO:
This step fa i 1 ed to address the performance of
step 3d.
f.
Step 6a:
This step failed to specify the relevance of
11 any
11
intact narrow range SG level.
g.
Step 6b
RNO:
This step failed to state the conditional
direction in accordance with the WG.
h.
Step 6c:
This step failed to identify which
SGs were
applicable.
i.
Step 8a RNO:
This step fa i 1 ed to address the performance of
step 8e.
j.
Step 11 caution: This step failed to specify what was meant by
11 excessive seal leakage
11 of an RCP. The operator stated that he
considered.
11 excess i ve 1 eakage
11 to be greater than 6 gpm which
was off-scale on the associated recorder, FR-l-154A.
20.
ECA-1.1
Loss of emergency coolant recirculation
a.
Entry conditions:
Asterisks were used; which was inconsistent
with the WG.
b .
Step 2 RNO:
This step failed to identify the components in
a particular train.
Appendix B
23
c.
Step 4a RNO:
This step fa i 1 ed to reference the procedure for
locally establishing fire water cooling to IA compressors,
OP-50.1, Removal
of the bearing cooling water system from
service.
d.
Step 4a RNO:
This step directed action which required that the
IA compressor local controller, 1-IA-C-l, be placed in auto per
Auto indication was obscured by the switch when auto
was selected.
e.
Step 4a RNO:
This step directed action which required that
BC-227 be verified closed per OP-50.1.
The
operator stated that
he would reach the valve by climbing on various equipment and
piping.
The valve was located approximately 20 feet above the
floor and no appropriate ladder was readily accessible.
f.
Step 4a RNO:
This step directed action whic~ required that the
11 fire water to air compressor
11 valve be opened per OP-50.1.
The
step did not specify FP-170 but incorrectly stated BC-170.
g.
Step 5a:
This step required maintaining cooldown rate prior to
step 5b initiating RCS cooldown .
h.
Step 5b RNO:
This step and local indication failed to address
the unconventional direction (counter clockwise to close) of
local steam dump handwheel operation.
Convenient means of local
communications to .the MCR during the high noise steam dumping
was not available.
i.
Step 6 RNO:
This step failed to direct CTMT fan operation if
desired by the TSC.
j.
Step 14 RNO:
This step failed to state the method
to be
used to thrott 1 e the SI valves which were not designed for
throttling.
k.
Step 21a:
This step failed to direct the warming of RHR if
desired by the TSC.
l.
Step 21c:
This step failed to direct the placing of RHR in
service if desired by the TSC.
m.
Step 22b RNO:
This step transitioned so as to omit step 22d.
Step 22d opened the SI accumulator isolation valve breakers, and
was important for operators to perform.
n.
Step 28b RNO:
Same as step 5b RNO above.
o.
Step 29a RNO:
Same as step 5b RNO above.
p.
Step 30b RNO:
This step transitioned so as to omit step 30d.
Appendix B
21.
24
q.
Step 31b RNO:
Same as step Sb RNO above.
r.
Step 33a:
Same as step 21a above.
s.
Step 33c:
Same as step 21c above.
t.
Step 34b RNO:
Same as step Sb RNO above.
u.
Attachment 2:
Same as entry conditions above.
v.
Attachment 2 Step 6:
This step required unlocking valves CH-728
and CH-447 which are not 1 ock.ed.
ECA-1.2
LOCA outside containment
a.
Entry conditions:
Asterisks were used in a manner inconsistent
with the WG.
b.
Step la:
This step re qui red the local operation of three
breakers. The NLO on the walk.down indicate-d that a control room
operator would have provided him the keys needed for the area.
The step did not indicate the need for keys .
c.
Step le:
A colon was used in a manner inconsistent with WG.
d.
Step 2a, d, and g:
Same as Step le.
e.
Step 2 continued:
Same as Step le.
22.
ECA-2.1
Uncontrolled Depressurization of All Steam Generators
a.
Cover page header:
The company name on the procedure did not
comply with the WG Section 6.2.1.a.l.
b.
Step
1
Note:
This
note
indicated that it was
not
mandatory
to monitor the
continuous action
page.
This
was not in accordance with the intent of the ERG.
c.
PSTG DEV Step 2c:
This step had the operator check for RCS hot
leg temperatures decreasing.
The intent of the ERG was to have
the operator check. for RCS hot leg temperatures stabilized.
This EOP step deviation from the ERG was not justified in the
SOD.
.
d.
Step 2b:
This step required the operator to maintain SG level
less than 50 percent but did not identify an adverse containment
value.
e .
Step 29:
No caution was listed to warn the operator of the
minimum PZR level for reenergizing PZR heaters, 24 percent
(55 percent), and the operator indicated that PZR heaters could
be reenergized if PZR level was .greater than 17 percent.
Appendix B
25
f.
Step 2a:
The values for maximum cooldown rate in the RCS cold
legs did not include natural circulation values.
g.
Step 28:
The required VCT levels in the action statement and
RNO were not the same value. 27 percent versus 34 percent.
h.
Step 3a:
This step did not list a minimum HHSI flow required
for continuation or exiting from this step. The operator could
not provide a specific value.
i.
Step 3 Note:
The recorders for RCP l eakoff fl ow Hi and Lo
range, recorders FR-1-154 and FR-1-1548, Referred to reactor
coolant pumps as 1, 2, and 3 instead of A, B, and C.
j.
Step 3e RNO:
No direction was given to the operator to locally
close the Chg. pump miniflow valves.
k.
Step 3 Note:
The instruments for RCP seal injection fl.ow
referred to reactor coolant pumps as 1, 2, and 3 instead of A,
B, and C.
l.
Step 4 RNO: Alternate AFW supply sources were not listed.
m.
PSTG DEV, Attachment 1 Step 6a:
This step did not caution the
operator that if offsite power was lost after SI was reset,
manual action may be required to start safeguards equipment.
This was not in accordance with the ERG.
n.
Attachment 1 Step 2a:
This step directed the operator to
11 Do
Not Continue" if containment pressure is greater than 14 Psia.
This statement would stop operator actions in this procedure.
o.
Attachment 1:
Various steps in this attachment used asterisks
instead of bullets.
p.
Attachment 1 Step 4b:
The steam generator blowdown permission
key switches did not have a position indication label.
They
were marked with a magic marker.
q.
Attachment 1 Steps 4a, c, and d:
The appropriate SG(s) were
not identified with the.listed valves.
r.
Step 6b:
The operator stated that he would initially verify
the alarm windows to satisfy this step.
The main steamline
radiation monitor alarms were identified as Unit 2 MSTM. ABC.
RAD. MON., and were physically located on the Unit 2 side of the
alarm panel.
s.
Step 6b:
The operator stated that he would initially verify the
alarm windows to satisfX this step.
The TD AFW pump exhaust
radiation monitor alarm was identified as Unit 1 Aux. Stm. Rad.
and was physically located on the Unit 2 side of the alarm
panel.
.
Appendix B
26
t.
Step 7 Caution:
This caution contained a conditional logic
action and contained no statement of consequences and was thus
inconsistent with WG direction.
u.
Step 7b:
This step directed the operator to reset SI if
necessary.
The operator was unable to specify what would
ful fi 11 the term
11 necessary
11 *
v.
Step 8g:
There was no
RNO step to direct the operator to
locally close the CS discharge valves.
w.
Step 8h:
There was no RNO step to direct the operator to
locally close the caustic supply valves.
x.
Step 9:
This step checked the RWST 1 eve l greater than the 22
percent recirculation switchover setpoint and did not have an
asterisk to identify the -step as a continuous action item.
y.
Step 8b:
This step had a less than 12 psia containment pressure
setpoint .for resetting CLS and Step 13a had a less than 14 Psi a
containment pressure setpoint for resetting CLS. It was also
noted that neither step documented the minimum pressure for
operation of the spray system which was i dent ifi ed in the
setpoint document as 10 psia.
z.
Step lOC:
This step required keys to accomplish the step, but.
did not identify the required keys.
aa.
Step lla RNO:
SI may have been terminated in Step 78 or attach-
ment 1 Step 1.
The action statement was an SI reinitiation
criteria which made the RNO statement inadequate~
ab.
Steps 11 band c RNO:
The band c RNO steps were reversed.
ac.
Step* llc: SI may have been terminated in Step 7b or attachment
Step 1.
The action statement was an SI reinitiation criteria
which made the RNO statement inadequate.
ad.
Step 12:
Same as Step 6a Attachment 1.
ae.
Step 14a RNO:
This step did not direct the operator to restore
intake canal level by going to AP 12.01, loss of intake canal
level. It also did not specify the minimum canal level which was
identified as 24 ft. in the setpoint document.
af.
Step 14cl RNO:
The seal leakoff line had been isolated by the
SI signal and seal leakoff would flow through the letdown relief
in containment.
The temperature instrument used by the operator
was located downsteam of the relief valve and would not see
leakoff temperature.
As a result, the operator may not close
TV-CC-107 as required by this step and after starting the CC
pump, the RCP seals could be cooled at greater than 1 degree per
minute, which could result in seal failure.
Appendix B
27
ag.
Step 18 RNO:
This step did not direct the operator to locally
operate various valves.
ah.
Step 19 RNO:
This step did not direct the operator to locally
operate various valves.
ai.
Step 14c5 RNO:
This step failed to reopen TV-CC-107 after
returning the CC system to operation. This could result in
failure to provide adequate cooling to the RCP seals.
aj.
PSTG DEV, Step 22:
Both subset steps in the action statement
are SI reinitiation steps.
The RNO steps did not direct the
operator to manually initiate SI.
During the walkthrough, the
operator walked through manually aligning valves.
This step was
not in accordance with the ERG, and was not justified in the
SOD.
ak.
Step 26:
While reviewing the letdown piping diagrams, the NRC
noted that RV 1203 on print 11448-FM-088C did not have a relief
setpoint. The operators were unable to readily provide the
relief setpoint.
aj.
Step 26 RNO:
This step did not direct the operator to locally
operate various valves.
al.
Step 28 RNO:
This step did not direct the operator to locally
operate various valves.
am.
Step 30 RNO:
This step could result in RCP seal failure because
CC may not have been restored and the seal leakoff temperature
instrument used by the operator was not in the flowpath with
MOV-1381 closed.
an.
Steps 30 c and d:
These steps did not list normal CC flows.
ao.
Step 31:
This step directed the ope.rator to verify all AC buses
energized by offsite power.
The operator had some difficulty
determining which instruments to review for verification of th1s
step.
ap.
Step 37 RNO:
This step did not direct the operator to locally
operate various valves as necessary.
aq.
Step 37:
11 Reset Auto-Start Inhibit
11 was not i dent ifi ed as a
1 oca 1 action.
ar.
Step 39:
Same as Step 22.
as.
Step 40b:
Same as Step 22.
at.
Step 40c:
Same as Step ')oc.
au.
St~p 40d:
Same as Step 100.
Appendix B
28
23.
ECA-3;1
SGTR with loss of rea~tor coolant - subcooled recovery
a.
. Cover page header:
The company name on the procedure did not
comply with the WG Section 6.2.1.a.l.
b.
Entry Conditions:
Transition procedures were denoted with
asterisks instead of bullets as required by the WG.
c.
Step 1 caution:
This caution inappropriately used
11 should
11
instead of
11 shal1
11 *
It was necessary to align the SI system for
cold leg recirculation.
d.
Step Sc:
This step required the operator to 11veri fy
11 CLS set,
and the operator stated that if the CLS had not been set that
action would be taken to change this condition.
Verify as
defined in the WG required the operator to observe that* a
condition existed but did not direct the
operator to take
action to change the condition.
The misuse of
11 verify
11
was
generic to all of the EOP procedures.
e.
Step 10a RNO:
This step required that a tota 1 AFW fl ow of 492
gpm be read from a combination of gages that had sma 11 sea 1 e
units of 7 gpm.
These awkward scale units made the determina-
tion of feed flow to this precision impractical.
f.
Step 10b RNO:
The grouping of bullets below this step did not
have the wording
11or
11 between them.
g.
Step 17 Caution:
This step contained inappropriate use of
llshould
11 instead of
11 shall
11 ; RCPs must not be started without
prtor status evaluation.
h. *
Step 28dl:
Key numbers were not designated for working SI
accumulator key switches.
i".
Step 36a second bullet:
The recorders for RCP leakoff flow Hi
and Lo range, recorders FR-l-154A and FR-1-1548, referred to
reactor coolant.pumps as 1, 2, ~nd 3 instead of A, B, and C.
24.
ECA-3.2
SGTR with loss of reactor coolant - saturated recovery
a.
b.
C.
Cover page header:
The company name on the procedure did not
co~~ly with the WG sectio~ 6.2.1.a.l.
Step 1 first note:
The fir.st 13 steps of ECA-3.1 must be
performed prior to the beginning of this procedure in order for
this procedure to work.
The note did not identify these actions
as mandatory.
Step 1 second note:
This note indicated that it was . not
mandatory to monitor the continuous action page.
This was not
in accordance with the intent cif the ERG.
...
Appendix B
29
d.
Step la:
LIC-CS-200A and LI-CS-2008 did not have the appropri-
ate power supply designations on the meter scale.
e.
Step lb:
The steps that were used to make up to the RWST were
incorrectly identified as not having a* preferred method of
performance.
f.
Step lb first and secontj bullets: These bullets required local
actions, but were not identified as requiring local actions.
g.
Step 2 caution:
Actions were contained in this caution, which
was inconsistent with WG section 6.4.4.a.S.
h.
Step 4a RNO:
This step required that a total feed flow of 492
gpm be read from a combination of three gauges that had small
scale units of *7 gpm.
The awkward scale units of FI-FW-200-A,
B, and C made the determination of feed flow to this precision
impractical.
i.
Step 4b RNO:
The groupings of bullets below this ste*p were not
divided into groups of three with the statement
11or
11 between
them .
j.
Step S second and third notes: Actions were contained in these
cautions, which.was inconsistent with WG section 6.4.4.a.S.
k.
Step Sc RNO first and second bullets: These- steps were listed
as having no preferred sequence.
This was inconsistent with
operator training and normal engineering practice of opening the
steam dumps before' using the PORVs.
l.
Step Sc RNO first bullet:
There was no metho~ for locally
opening the PORVs although the step required this action.
m.
Step Sc RNO third and fourth bullets: These steps were listed
as having no preferred sequence.
This was inconsistent with
operator training.
n.
0.
Step 8:
This step required a switch position of
11off
11 *
pressurizer heater group C did not have an off position.
switch did have a position "pull to lock
11 *
The
The
Step 13b RNO:
This step required the operators to manually open
the charging line isolation valve.
The operators interviewed
indicated that this valve actually had to be opened locally.
T~ere was no local method for accomplishing this step .
\\
Appendix B
- - - - - - - - - -
30
p.
Step 13e:
This step required the use of FI-2-122A.
The scale
on the instrument was compressed _on the lower end of the scale.
This made a flow reading of very close to 25 gpm difficult to
distinguish.
q.
Step 15a:
This step used the verb
II secured
11 *
This statement
was inconsistent with the operators* action of placing the
system in
11 auto
11 *
r.
Step 17:
This step required the verification that five
parameters were decreasing or stable.
No
indication was
available to indicate this information.
This information
required trending to determine the appropriate status.
s.
Step 18c:
This step required the pressurizer heaters to be
turned on without determining if they were actually needed.
t.
Step 21d RNO:
This step indicated that IA should be available.
The step did not specify containment IA.
u.
Step 2ldl: This step required keys to accomplish the step, but
did not identify the required keys (Keys 11, 12 and 13) .
. v.
Step 22b second bullet:
The procedure title was inconsistent
with the preceding bullet, in that it did not identify that
actions were required .to be performed from the MCR.
w.
Step 24b:
The~ verb
11 veri fy
11 was used to indicate that this step
should be accomplished, if possible.
This was not consistent
with the definition of
11 verify
11 in the WG.
x.
Step 25:
This step was not written in the form of a single
step, which was not consistent with the WG. *
y.
Step 27:
This step indicated that the substeps could be
performed in any order.
Under some conditions it would be
necessary to perform the first bullet before the second bullet.
z.
Step 27 seventh bullet:
The abbreviation RSST did not appear on
the approved WG abbreviation list.
aa.
SOD step 5 forth note:
The reference to set point E .13 was
incorfect.
The set point was applicable to the third note.
ab.
Setpoint B.8:
This calculation concerned the RCS
pressure
correspcindi ng to the* shutoff pressure of the 1 ow head SI pumps
plus allowance for normal channel accuracy.
The setpoint was
based upon the SI pump shutoff head for an RWST level of zero.
The calculation failed to include the additional head for an
Appendix B
31
RWST level of 100 percent.
As a result, the calculated shutoff
head setpoint was 366 ft instead of approximately 408.5 ft.
The equivalent RCS pressure setpoint was 250 psig instead of
approximately 270 psig.
The licensee 1 s setpoint contractor
stated that this setpoint error would be corrected in the next
EOP setpoint document revision, scheduled to be completed by
the contractor in about four months.
Additionally, the portion
concerning dPelev was written incorrectly.
The calculation as
written gave the result -711.7 psi. The mathematically correct
answer was 7.3 psi. This setpoint was used in eighteen steps in
the EOPs.
ac.
Setpoint B.9:
This calculation concerned the RCS pressure
corresponding to the shutoff pressure of the low head SI pumps
plus allowance for normal channel accuracy and post-accident
transmitter errors.
The setpoi nt was based upon the SI pump
shutoff head for an RWST level of zero. The calculation failed
to include the additional head for an RWST level of 100 percent.
As a result, the calculated shutoff head setpoint was 366 ft
instead of approximately 408.5 ft.
The equivalent RCS. pressure
setpoint was 525 psig instead of approximately 545 psig.
The
licensee 1 s setpoint contractor stated that this setpoint error
- would be corrected in the next EDP setpoint document revision,
scheduled to be completed by the contractor in about four
months.
Additionally,The portion concerning dPelev was written
incorrectly.
The calculation as written gave the result 23.7
psi.
The mathematically correct answer was 7.3 psi.
This
setpoint was used in eighteen steps in the EOPs.
ad.
Setpoint K.1:
This calculation concerned RCS subcooling margin
and used a combination of RCS temperature measurement and
pressure measurement uncertainty converted into a temperature
based on saturated conditions.
Under certain conditions, at
lower RCS pressures (below 400 psi) the RCS subcooling value of
30 degrees would indicate that the operator was operating in
the subcooled region; however, based on the uncertainties of the
calculation the operator may have been operating at saturated
conditions.
This phenomenon would increase the probability of
operation in the saturated region as RCS pressure decreased and
increase the probability of vessel voiding.
This calculation
was based upon a three standard deviation uncertainty range
for the subcooling margin at .a reactor coolant system pressure
of 400 psi.
However the subcooling margin calculation was
assumed to be valid at lower pressures by maintaining the
calculated subcool ing margin and increasing the uncertainty
in the value of subcoo 1 i ng and subsequently increased the
probability of operation in the saturated condition.
As RCS
pressure decreased to 100 psig the probability of operating
in the saturated region would have increased to a l eve 1 of
Appendix B
32
approximately one standard deviation and the uncertainty in the
subcool ing margin value would have correspondingly decreased.
This exceeded the minimum acceptable probability limit for
errors of 95 percent (2 standard deviations) per NRC Reg.
Guide 1.105.
This setpoint was used in 16 steps in the EOPs.
25.
ECA-3.3
SGTR without pressurizer pressure control
a.
PSTG DEV, Step 4:
The ERG step was deleted with the justifica-
tion that establishing auxiliary spray flow would terminate SI.
However, the ERG step included other actions and it was possible
that SI may not be required. The EDP SOD did not address these
aspects of the deviation.
b.
PSTG DEV, Step 4:
The corresponding step in the ERG stated
11 IF
narrow range level in any intact SG continues* to increase.
11 The
word
11 intact
11 was not included in the procedure and the EDP SOD
did not address this deviation.
c.
Step llc RNO:
The indicator for RCP seal leakoff temperature
was not in the step, but was needed to insure that the operators
use the correct indicator .
d.
e.
Steps 19a and d RNO:
Same as step llc RNO above.
Step 21:
The location of infrequently operated equipment was
not indicated.
f.
Step 22cl:
There were no permanent position indicators on the
SI Accumulator switch keys and the key numbers were not included
in the EOP.
g.
Step 26 caution:
This caution contained an implied action
statement.
h.
PSTG DEV, Step 30 note:
This ERG note which stated
11The upper
head region may void during RCS depressurization if RCPs are not
running.
This may result in a rapidly increasing PRZR l~vel
11 ,
was deleted from the EOP.
The justification failed to consider
the fact that PRZR heaters had been secured since step 13 and
that the note only stated that a rapid rise in pressurizer level
may occur.
i .
PSTG DEV, Step 31a:
The corresponding step in the ERG was
divided into two separate steps with different RNO actions.
The
justification for this deviation in the EDP SOD stated that it
simplified the logic without unnecessary delay.
Since the EOP
RNO step could send the operator to procedure steps that mtght
not be applicable, the deviation could introduce delays .
\\
Appendix B
33
26.
F-0
Critical safety function status trees
a.
General comment:
The ERF computer simulation could not display
the correct path if an erroneous instrument reading was input to
the system. There was no method for bypassing incorrect inputs
and displaying the correct path.
The status or path color or
alternate paths were not displayed.
The operators interviewed
stated that they would use the F-0 procedure and would not use
or rely on the E~F computer output when using the EOPs.
27.
FR-C.l
Response to inadequate core cooling
a.
PSTG DEV, Step 8b:
This step stated
11 H2 concentration -
LESS
THAN 4~~
11 *
The corresponding step in the ERG stated "Hydrogen
concentration -' LESS THAN 6.0~~ in DRY AIR.
11 The EOP SOD did not
address this deviation.
b.
PSTG DEV, Step 9a RNO:
This step stated "Maintain total feed
flow* greater than 350 gpm until narrow range level greater
than ...
11 *
The corresponding step in the ERG stated
11 Increase
total feed flow to restore narrow range level greater than ...
11
The EOP SOD did not address this deviation .
C.
PSTG DEV, Step 20:
The EOP step added the extra words "USING
STEAM DUMPS" at the end of the high level step. This statement
was not included in the ERG.
This step deviation was not
justified in the EOP SOD.
d.
Step 21b:
There were no permanent position indicators on the SI
Accumulator isolation key switches and the key numbers were not
included.
28.
FR-C.2
Response to degraded core cooling
a.
Step 1 caution:
This caution contained no statement of
consequences as per the WG.
The caution contained a continuous
monitoring action and a reference to another procedure.
b.
Step 2a
RNO:
This step provided five alternative actions
(bullets one through five) to the operator but provided no
guidance on necessary or sufficient conditions.* No specific
direction was provided on how to perform the alternatives. The
operator on the walkdown indicated either of the first two
alternatives as sufficient, the third and forth alternatives
together as probably sufficient, and required referencing part
of attachment 7b to FCA-1.00 to perform the fifth alternative.
FCA-1.00 was not referenced in this step and the operator took
considerable thinking and time to find it .
- ':.**:: ., . ~. *, ..... .
'
.* **-* -.
- '
... *.
. .*, .~ . -. : .
Appendix B
34
c.
Step 4 note:
This note inappropriately included a caution
statement.
d.
Step 10b RNO:
This step required the operator to locally
crosstie to turbine bldg air but provided no specific direction
or reference for performing the action .
. e.
Step llb~ This step requjred the operator to dump steam to the
main condenser but provided no direction on how to perform this.
The operator on the walkdown indicated this was a relatively
complicated action.
f.
Step llb RNO:
This step provided the operator with two
alternatives for action but provided no direction as to
preferred action.
The operator on the wa lkdown indicated he
believed there was a preierred action.
g.
Step 13b RNO:
This step used IAW as a referencing term but IAW
was not defined as a referencing term in the WG.
h.
Step 14 caution:
This caution was located near the bottom of
the page while most of the related steps were on the following
page.
i.
Step 15b RNO:
Same as Step llb RNO.
j.
Step 16 RNO:
Same as Step 2a RNO.
k.
Step 17 RNO:
Same as Step 13b RNO.
l.
Genera 1 comment
on
FR-C. 2:
This procedure contained many
transitions (eight transitions on pages five and six alone).
This could increase the probabi 1 i ty of operator errors and
place keeping problems, especially when the operator is under
psychological stress during an emergency.
29.
FR-C.3
Response to saturated core cooling
~
a.
No comments.
30.
FR-H.1
Response to loss of secondary heat sink
a.
PSTG DEV, Entry Conditions: The EDP stated that this procedure
could be entered when an ORANGE path existed.
The ERG did not
contatn that provision and the F-0.3 Heat Sink CSF Status Tree
did not contain the provision.
b.
Step 1 second caution:
The caution contained an action state-
ment and did not indicate the consequences of not complying
with the caution.
.. - ,., .. - '
Appendix B
35
c.
PSTG DEV, Step 4e RNO:
This step stated
11 GO TO Step 10.
11
The corresponding step in the ERG stated
11GO TO Step 7.
11
The EDP SOD did not address this deviation.
d.
Step 11 caution: This caution did not indicate the consequences
of not complying with the caution .
. e.
Step 11 first and second notes:
These notes contained action
statements.
f.
PSTG DEV, Step 14a RNO:
The EDP directed the operator to
continue to step 16.
This deviated from the ERG required step i
sequence, but the EDP SOD did not address this deviation.
This
was a 1 so an examp 1 e of the use of the phrase
11Cont i nue wi th
11
that required the operator to do all of the steps that were
initially by passed in order to comply with the ERGs.
If the
operator had only done the steps specified in the EDP, then step
15, which established IA to containment, would not have been
done.
See comments in E-1 step 6d2 RNO and step 10a RNO above
for additional information.
g.
PSTG DEV, Step 18:
The corresponding step
11Maintain PRZR PORVs - AT LEAST TWO OPEN.
11
not included in the EDP. The EDP SOD did
deviation.
in the ERG stated
This statement was
not address this
h.
PSTG DEV, Step 23b RNO:
The corresponding step in the ERG
stated
11THEN start one LHSI pump if none running.
11
The EDP just
stated
11THEN verify one LHSI pump running.
11
The EDP SOD did not
address this deviation.
i.
PSTG DEV, Steps 25a and b RNO:
The corresponding step in the
ERG stated
11 BIT
11 and the EDP stated
11SI
11 *
The EDP SOD did not
address this deviation.
j.
Step 28 RNO:
The step did not include the action to "leave one
in AUT0
11 after PRZR PORVs which was included in step 25.
This
step would improperly leave all PRZR PORVs isolated.
32.
FR-H.3 * Response to steam generator high level
a.
No. comments.
33.
FR-H.4
Response to loss of normal steam release capabilities
a.
Note 1:
The intent of Note 1 was to insure pressure was main-
tained low enough to prevent steam rel ease through a safety
valve.
The 1085 psig limit may have been too high to prevent
weapage or subsequent actuation following an initial actuation,
since subsequent safety valve actuation has been found to
typically occur at lower pressures.
Appendix B
36
b.
Step 2 RNO:
The operator was told to manually or locally dump
steam.
Under this step the SG PORVs were listed. The SG PORVs
could not be locally operated.
34.
FR-H.5
Response to steam generator low level
a.
Step 2:
The verb
11 verify
11 was used to indicate that this step
should be accomplished, if possible.
This was not consistent
with the definition of
11 v,erify 11 in the WG.
b.
Step 2 RNO:
This step indicated that an action should be
accomplished manually.
The step actually had to be accomplished
locally.
c.
Step 4:
This step required that a feed flow of 168 gpm be read
from a gauge that had small scale units of 7 gpm.
The awkward
scale units of FI-FW-200-A, B, and C made the determination of
feed flow to this precision difficult.
35.
FR-I.1
Response to pressurizer high level
a.
b.
C.
d.
e.
f.
Step 1:
This step was not written in the form of a single step,
and was not consistent with the WG.
Step 2a2 RNO:
This step used the verb verify.
This use of
verify in this step was inconsistent with the WG definition. If
the pump miniflow valve was not opened the GHG/SI pump could not
be started.
Step 2b RNO:
This step indicated that there was no preferred
order for performance of the substeps, although it was clear
that establishing CTMT IA was the preferred method.
Step 2b RNO first bull et:
This substep contained no action
verb.
11 did not clearly indicate the required
action.
Step 2c2 RNO:
'This step used the verb verify.
This use of
verify in this step was inconsistent with the WG definition. If
the pump miniflow valve was not opened the GHG/SI pump could not
be started.
Step 4bl:
This step used the verb verify. This use of verify
in this step was inconsistent with the WG definition. There was
no previous step that would have established a charging flow of
greater than 25 gpm.
Charging flow was established at 25 gpm in
step 2.4.c .
\\
Appendix B
37:
g.
Step 4b3:
This step used the verb verify. This use of verify
in this step was inconsistent with the WG definition.
This
step did not have an RNO which directed the operator to close
the letdown orifice isolation valves.
36.
FR-I.2
Response to pressurizer low level
No comments.
37.
FR-I.3 Response to voids in reactor vessel
a.
Step 2b second bullet RNO:
This step required the cross tying
of containment IA and turbine building instrument air.
This
step was performed by locally manipulating two locked valves.
These two valves and key numbers were not denoted in the step.
b.
Step 3b:
This step did n.ot have an RNO to locally open the
valve.
c.
Step 17 a, b, and c: This step required starting of fans.
Fans
in Step a could be started manually but those in Steps band c
must be started locally. The Step did not indicate manually or
locally.
38.
FR-P.1
Response to imminent pressurized thermal shock condition
a.
Step 8c RNO:
This step failed to reference the computer point
required to access the RCP seal leakoff temperature.
b.
Step 15c:
This step and its associated RNO were split between
two pages.
39.
FR-P.2
Response to anticipated pressurized thermal shock condition
a.
Step lel RNO:
This step was a local action but was not
identified as such in the step.
b.
Step le2 RNO:
This step was a local action but was not
identified as such in the step.
c.
Step le4 RNO:
This step directed the operator to isolate
feedwater to faulted SG(s) unless necessary for RCS temperature
control but provided no specific information on how to determine
if necessary for temperature contra l.
The operator on the
wa l kdown indicated this was determined by operator judgment
based on knowledge of plant conditions .
Appendix B
38
40.
FR-S.1
Response to nuclear power generation/ATWS
a.
Step 5b RNO:
This step failed to indicate the direction of
operation of the trip lever when the local indication was
obscured.
The operator stated the incorrect direction for
operation.
b.
Step 11:
This step failed to specify the requirements for the
isolation of AFW line(s).
c.
Step
12b:
This step failed to reference OP-lF for SOM
calculation.
d.
Step
13 caution:
The continuation of boration to obtain
adequate shutdown margin was incorrectly stated as a caution
statement.
41.
FR-S.2
Response to Loss of Core Shutdown
a.
Step 1:
This step required the operator to verify intermediate
range flux decreasing but did not identify an acceptable rate.
The ERG background information 1 isted an acceptable value as
greater than -.2DPM.
b.
Step 2 RNO: This step required the operator to emergency borate
which w.as inconsistent with Step la RNO which required the
operator to borate.
The ERG required a preferred method of
borating the RCS.
42.
FR-Z.1
Response to containment high pressure
a.
Step 1 caution:
This caution did not contain a statement of
consequences as per the WG.
b.
Step 6 RNO:
This step directed the operator to manually or
locally close the MSTVs and bypass valves._ The MSTVs could not
be closed locally.
c.
Step 7 caution:
This caution suggested action based on a
conditional statement and did not contain a statement of
consequences.
d.
Step 8 note:
This note contained an action.
e.
Step 8a and c RNO: This step used IAW as a referencing term but
IAW was not defined as a referencing term in the WG.
f.
Step 9b:
procedure
specified
This step directed the operator to "RETURN
TO
and step in effect".
In the WG,
11 RETURN T0 11
was
for use in tra:\\!,s it ions but not for branching.
- Appendix B
39
g.
Step 10a RNO items 1 and 2:
Item 1 presented a conditional
action to perform steps 10 through 12 when CTMT pressure was
less than 14 psig.
Item 2 directed the operator to
11 RETURN TO
procedure and step in effect.
11
The way the procedure was
written, the operator may have some confusion over whether to
wait for the pressure to be less than 14 psig before taking any
further action or continuing with the steps in effect.
Also
there was no formal place keeping method to remind the operator
of item 2 and which procedure would be in effect in response
to item 2.
This step appeared to be a continuo~s actio~
requiring an asterisk as per the WG.
h.
Step 10a RNO item 2:
Same as step 9b.
i.
Step 10d and e:
The operator on the walkdown made the observa-
tion that the procedure had the operator verify A SW header
components operating before isolating the A SW header, but did
not have the operator verify B SW components before isolating
j.
Step 12 RNO:
This step required the operator to perform one of
three options within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from the initiating event.
No
specific guidance was provided regarding preferred action and
there was no method of place keeping to avoid overlooking these
actions.
k.
Step 13:
Same as step 9b.
43.
FR-Z.2
Response to Containment Flooding
a.
General Comment:
The concern with flooding in containment was
that critical plant components needed for plant recovery could
be damaged and rendered inoperable.
The procedure did not
identify a maximum level, the flood level, or what equipment
would be lost at what levels.
b.
Step 1:
The intent of this step was to identify sources of
water to the containment sump, and to isolate them if possible:
1)
The sources listed in Step 1 were not complete.
2)
Instrumentation
used
to
identify unexpected
sources
of water were not listed.
3)
Isolation valves were not listed.
c.
Step ld: This step had the control room consult with the TSC to
determine required actions:
1)
The TSC may not be *~anned.
Appendix B
44.
40
2) , The step did not include options to isolate leakage.
d.
Step 2:
This step requested chemistry to sample the containment
sump:
1)
The chemistry procedure (l-COP-168) only covered sampling
for activity.
2)
Following a loss of offsite power CC would not be readily
available and as a result, chemistry would not be able to
sample the containment sump.
Consequently, this procedure
could not be completed.
.
e.
Step 3:
This step referenced TSC personnel for exchange of
information and subsequent actions, and the TSC may not be
manned.
f.
Steps 2 and 3:
The containment sump pumps may not be available,
following
containment flooding,
for sampling purposes or
transfer of sump water.
The containment sump
pump
upper
temperature limit for pumped liquid was 115 degrees F, per the
manufacturer's technical manual, PG8.
Also the pump electrical
connections may be under water .
FR-Z.3
Response to containment high radiation level
No comments.
45
FR-Z.4
Response to containment positive pressure
a.
b.
C.
d.
Step 2bl:
This step had no direct method of determining the
required RCP sea 1 * return temperature.
The
seal return
temperature instrument was not in a flow path with seal return
isolated.
Specific guidance was
required to prevent the
operator from reading an incorrect temperature and as a result
chill shocking the RCP seal.
Step 3 RNO al:
This step used the verb verify.
This use of
verify in this step was inconsistent with the WG definition.
Step 14a RNO:
This step used the verb verify.
This use of
verify in this step was inconsistent with the WG definition.
If the valve MOV-CS-lOOA was not opened 1-CS-P-lB could not be
started.
Step !Sa:
This step used the verb verify. This use of verify
in this step was inconsistent with the WG definition.
If the
valve MOV-CS-lOOA was not opened 1-CS-P-lB could not be started .
\\
'
- Appendix B
41
III. AP comments
1.
Rod control system malfunction
a.
Step 2b RNO:
This step incorrectly referenced the Reactor
trip/safety injection procedure as EP-1.00 vice l-E-0.
This
error was consistent throughout the procedure.
b.
Step 8 RNO:
This step directed that reactor be matched and
stabilized at less than 75 percent powe~. Previously, in step 7
the procedure directed that th'e turbine power be less than : or
equal to 70 percent.
2.
Control rod misalignment
a.
Step 1 note:
This step failed to specify whether reactor or
turbine power was assumed to be 70 percent or less~
b.
Step 5 caution:
This step failed to specify the 75 percent
power limit as reactor or turbine power.
C.
d.
Step 7:
This step directed the placement of
11all
11 disconnect
switches to open vice the affected bank 1 s disconnect switches ..
Step 10:
This step failed to specify the actions required to
reset affected bank P/A converter.
3.
AP-1.02 Malfunctioning individual rod ~osition indicator (IRPI)
'
a.
Step 1 caution:
This step incorrectly directed verification of
rod to band vice rod to bank position.
b.
Step 4 RNO:
This step failed to specify from when the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
time limit was to be commenced.
c.
Step 5:
This step failed to indicate which of the actions were
to be performed by the instrument shop.
4.
Conditions requiring emergency boration
a.
Attachment 1:
This attachment was not referenced in the
procedure and appeared to serve no useful purpose.
5.
AP-4.00 Nuclear instrument malfunction
a.
General:
Asterisks were used throughout the procedure instead
of bullets .
\\
. , .. * ~ -*-*. ***** ..
,
Appendix B
42
6.
Radiation
monitoring
system
process
vent monitor(s)
alert/alarm
a.
Step 1:
This step was not numbered.
b.
Step 3b:
This step failed to specify to whom the information
was to be provided.
c.
Step 3b:
This step required the reading of flow rate indica-
tion.
The
associated.
indication,
FI-GW-100,
was
not
referenced in th~ step and was poorly incremented in divisions
of 7 CFM.
d.
Step 3b:
This step required the reading of the Kaman effluent
monitors,
RI-GW-130-1
and
RI-GW-130-2.
The
indicator's
parameters were displayed numerically (i.e., 47 corresponds to
micro-curies/second).
There were no procedures in place to
state the parameter associated with a particular number.
7.
Radiation monitoring system liquid waste monitor
a.
Step 9:
This step directed checking the flow monitor vice the
flow rate monitor.
The entry conditions described it as the
flow rate mon,toring device.
8.
Probable causes and references
a.
Step 3 RNO:
This step had improper lettering of substeps.
9.
Radiation mon~toring system CC monitors A & B
a.
Entry conditions: The information was incomplete. The equipment
that would alarm was not listed.
b.
Attachment 1:
This attachment was not referenced in the
procedure and appeared to serve no useful purpose.
10.
Radiation monitoring system restricted control area monitors
a.
General: Asterisks were used throughout the procedure instead of
bullets.
b.
General:
The procedural steps did not address all of the
monitors listed in the entry conditions.
c.
Step 3:
This step had the operator dump the control room
bottled air without verification of radiation levels .
'
Appendix B
- 43
11.
Radiation monitoring system main control room air monitor
a.
Attachment 1:
This attachment was not referenced in the
pro~edure and appeared to serve no useful purpose.
12.
Radiation monitoring system containment particulate and gas
and manipulator crane
a.
Step 10:
The step directed the checking of three valves, but
did not indicate w~at position the valves should be in.
Since
the positions would vary depending on the tause of the alarm,
more information was required in this step.
b.
Step 13:
No guidance was provided to the operator if the
radiation level was high, but had not increased by a factor of
1000.
c.
Attachment 1:
This attachment was not referenced in the
procedure and appeared to serve no useful purpose.
13.
Radiation monitoring system condenser air ejector
a.
General: Asterisks were used throughout the procedure instead of
bullets.
14.
Radiation monitoring system SG blowdown
a.
Entry conditions:
The alarm meters that this procedure applied
to were not inaicated.
b.
Attachment 1:
This attachment was not referenced in the
procedure and appeared to serve no useful purpose.
15.
Radiation monitoring system recirc spray cooler service
water outlet
a.
b.
C.
d.
e .
Entry condition 1: This entry condition was not written in the
same format as other AP entry conditions.
Step 6:
These actions were preceded by an asterisk instead of a
bullet, which was inconsistent with the WG.
Step 7:
These actions were preceded by an asterisk instead of a
bullet, which was inconsistent with the WG.
Attachment 1 step 1:
These actions were preceded by an asterisk
instead of a bullet, which was inconsistent with the WG.
Attachment 1 step 2:
This reference was preceded by an asterisk
instead of a bullet, wht~~ was inconsistent with the WG.
r
Appendix B
44
16.
Radiation monitoring system reactor coolant letdown monitors
a.
General:
Asterisks were used throughout the procedure instead
of bullets.
17.
Radiation monitoring system discharge tunnel
a.
Step 1 RNO:
This step referred to step 5 which re qui red the
monitor to be verified operable.
When the transition to step 5
was accomplished from the RNO column a PT was required to verify
the monitor operable.
When the same step was transitioned to in
the AER column a PT was not required.
b.
Step 2:
This step referred to closing valve HCV-LW-104A and B.
The method did not direct the operator to set the pot to zero.
There was no closed position on the pot.
c.
Step 2:
This action was preceded by an asterisk instead of a
bullet, which was inconsistent with the WG.
d.
Step 4:
These actions were preceded by an asterisk instead of a
bullet, which was inconsistent with the WG.
e.
Step 5:
Refer to comment for step 1 RNO.
f.
Step 6:
These actions were preceded by an asterisk instead of a
bullet, which was inconsistent with the WG.
g.
Attachment 1 step 1:
This reference was preceded by an asterisk
instead of a bullet, which was inconsistent with the WG.
h.
Attachment 1 step 2:
This reference was preceded by an asterisk
instead of a bullet, which was inconsistent with the WG.
18.
Radiation monitoring system containment high alarm range
area
a.
User block:
The nomenclature in this block was inconsistent
with other APs.
b.
Step 6:
These actions were preceded by an asterisk instead of a
bullet, which was inconsistent with the WG.
c.
Step 7a:
These actions were preceded by an asterisk instead of
a bullet, which was inconsistent with the WG.
d.
Attachment 1 step 1: These actions were preceded by an asterisk
instead of a bullet, which was inconsistent with the WG.
e.
Attachment 1 step 2:
This reference was preceded by an asterisk
instead of a bullet, which was inconsistent with the WG.
Appendix 8
19.
'
45
Radiation monitoring. system reactor containment area
monitors and CHRRM
a.
General: Asterisks were used throughout the procedure instead of
bullets.
20.
RM system process vent monitor or process vent flow monitor
malfunction
a.
Step 2b RNO third and jforth bullets: These actions could only
be preformed locally. This was not specified.
b.
Step 4a RNO:
This step identified a transition to AP-5.24;
however, AP-5.24 did not list AP-5.16 as an entry condition.
c.
Attachme~t 1 step 1:
The substeps were preceded by asterisks,
which was inconsistent with the WG.
d.
Attachment 1 step 2:
The substeps were preceded by asterisks,
which was inco~sistent with the WG.
21.
Loss of RCP seal cooling
a.
Entry Condition:
This
section did not list all entry
conditions.
b.
Step 8:
Prior to entry into this step there was no caution
about the maximum allowable cooldown rate of 1 degree F/minute.
c.
Step 12:
This step did not caution the operator to check seal
1 eakoff temperatures 1 ess than 235 degrees F and if not to
es tab 1 i sh a coo 1 down rate of 1 ess than or equa 1 to 1 degree
F/minute.
22.
AP-10.00 Station ~lackout
a.
Entry conditions: This section lacked specifics as to when this
procedure, instead of the EOPs, would be the proper procedure to
use in a loss of power casualty.
b. . Genera 1 comment:
The procedure did riot inc 1 ude any actions to
restore communications capabilities.
The procedure did not
mention the time sensitivity of some recovery actions and the
cross connect capabilities that could be used to mitigate the
adverse affects of this casualty.
Appendix B
46
c.
Step 4c:
The fact that the operation of the stub bus tie was a
local operation was not indicated in the procedure step.
d.
Step 9:
This step lacked the detail necessary to insure that
none of the actions necessary to carry out the step would be
overl ook.ed.
e.
f.
g.
h.
Step 11:
Step 15:
Step 16:
Step 18:
numbers.
remember
Same as step 9 above.
Same as step 9 above.
The undefined abbreviation
11TS
11 was used in this step.
This step referred to two Attachments by their
Their titles were not included to help the operators
how the attachments were used.
i.
Step 20:
The specific indication lights, that must be checked
on the mimic bus, were not included in the step to insure the
correct bus status was obtained.
j.
Step 21:
Same as step 20 above .
k..
Step 21 RNO:
The undefined abbreviation
11 RSSP was used in this
step.
1.
Step 23:
Same as step 20 above.
m.
Step 25:
Same as step 20 above.
n.
Steps 26a and c RNO:
The verb
11 verify
11 was used in these steps,
but no action statement was included to cover the situation when
the break.er was not found in the desired position.
o.
Step 26 RNO:
The local operations were not indicated in the
step as required.
p.
Step 28:
Same as step 20 above.
q.
Step 29:
Same as step 20 above.
r.
Step 31:
Same as step 20 above.
s.
Step 32 RNO:
Same as steps 26a and c RNO and step 26 RNO above.
t.
Step 33:
Same as step 26 RNO above.
u.
Step 36 RNO:
Same as steps 26a and c RNO and step 26 RNO above .
\\
Appendix B
47
v.
Attachment 3 Steps lA and B, Steps 2A and Band Steps 3A and B:
Same as step 20 above.
w.
Attachments 3 and 4:
Same as steps 26a and c RNO above.
x.
Attachment 4 Steps lA and B ~nd Steps 2A and B:
Same as step 20
above.
y.
Attachment 5:
This attachment was not referenced in the
p1ocedure and appeared to serve no useful purpose.
23.
Loss of vital bus I
24.
25.
26.
27.
a.
General:
Asterisks were used in the procedure instead of
bullets.
b.
Step 3 RNO:
The old numbering scheme was used for the reactor
trip procedure.
c.
Step 21 Note:
The note was misleading in that it stated that
flow was lost, instead of flow
11 indication
11 was lost.
AP 10.-02
Loss of vital bus II
a.
General:
Same as AP 10.01 general comment.
b.
Step 3 RNO:
Same as AP 10.01 Step 3 RNO.
AP 10. 03
Loss of vital bus I II
a.
General:
Same as AP 10.01 general comment.
b.
Step 3 RNO:
Same as AP 10.01 Step 3 RNO.
Loss of vi ta l bus IV
a.
General:
Same as AP 10. 01 general comment.
b.
Step 3 RNO:
Same as AP 10.01 Step 3 RNO.
Loss of a semi-vital bus and AC distribution panel
a.
Entry conditions:
This section did not list the busses that
this procedure applied to and did not include the specific
conditions when this procedure, instead of the EOPs, should be
used to respond to the situation.
b.'
Step 8:
The local operation was not indicated in the step as
required.
Appendix 8
c.
48
Step 9:
The step lacked the detail necessary to insure that
none of the actions required to carry out the step would be
overlooked.
ct.
Step 10a:
Same as step 8 above.
e.
Step 10a:
The undefined abbreviation
11 SV8
11 was used.
f.
Step 11:
The title of the attachment was not included to help
the operators remember how the attachment was used.
g.
Attachment 1:
The SVBs were incorrectly indicated as SVBI and
SVB2 vice lSVBl and 2SV81.
h.
Attachment 4:
This a.ttachment was not referenced in the
procedure and appeared to serve no useful purpose.
28.
Loss of DC power
a.
General :
Asterisks were used in the procedure in stead of
bullets.
b.
Step 1:
The operator indicated that he would be checking for
reactor trip breaker indication lights not lit, but the step
called for the reactor trip breaker to be open.
c.
Step 1:
The operator stated he would observe the valve indica-
tion not lit and the TDAFW pump running.
The step required the
operator to check the TDAFW pump steam supply valve open, but
the indicating light may not have power.
ct.
Step 2:
This step required the operator to verify the generator
output breaker open, but the indicating lights may not have
power.
Breaker position would have to be checked locally.
e.
Step 6d:
The procedure identified valve SOV-()-113A power
available and this valve designation did not exist on the
control board.
29.
Loss of auto load shed
a.
General :
Asterisks were used in the procedure in stead of
bullets.
b.
Step 1: The instrumentation necessary to verify that two station
service buses were energized from one RSS transformer was
unclear to the operator.
\\
Appendix B
49
30.
Partial loss of reserve station service
a.
Step 4:
This step required the verification of
11 RCS circula-
tion -
forced
11 *
The
RNO for this step did not involve the
attempt to start or restart a RCP,
but went directly to
AP-39.00, Natural Circulation of RCS or AP-27.00, Loss of Decay
Heat Removal.
b.
Step 6:
This step required the operator to
11 verify unit condi-
tions - stable
11 , without defining the normal parameters for loss
of reserve station service.
c.
Step 8:
This step required the operator to
11 check radiation
monitors -
normal
11 without specifying the specific monitors to
check.
d.
Step 8
RNO first through third asterisks:
These actions
were preceded by an asterisk instead of a bullet, which was
inconsistent with the WG.
e.
e.
Step 9 first through third asterisks:
preceded by
an
asterisk instead of a
inconsistent with the WG .
Step 10 first through third asterisks:
preceded by
an
asterisk instead of a
inconsistent with the WG.
These actions were
bull et, which was
These actions were
bull et, which was
f.
Attachment 2 step 2 first through forth asterisks:
These
statements were preceded by an asterisk, which was inconsistent
with the WG.
31.
Partial loss of station service
a.
Step 2:
The verb
11verify
11 was used to* indicate that this step
should be accomplished, if possible.
This was not consistent
with the definition of
11 verify
11 in the WG.
b.
Step 6:
These actions were preceded by an *asterisk which was
inconsistent with the WG.
c.
Step 7:
Due to non-vi ta l nature of this step it was not
appropriate for inclusion in this AP.
ct.
St~p 7 first and second asterisks:
These actions were preceded
by an asterisk instead of a bullet, which was inconsistent with
the WG.
e .
Step 8:
Due to the non-vital nature of this step, it was not
appropriate for inclusio~, in this AP.
Appendix B
f.
50
Step 7 first through third asterisks:
preceded by an asterisk instead of a
inconsistent with the WG.
These actions were
bullet, which was
g.
Attachment 2. step 1:
This action was preceded by an asterisk,
which was inconsistent with the WG.
h.
Attachment 2 step 5:
This action was preceded by an asterisk,
which was inconsistent with the WG.
i.
Attachment 3 step l:
This action was preceded by an asterisk,
which was inconsistent with the WG.
j.
Attachment 3 step 2:
This action was preceded by an asterisk,
which was inconsistent with the WG.
32.
Service water system abnormal conditions
a.
b.
C.
d.
e.
f.
Step 3a first and second hollow bullets:
This symbol was not
defined in the WG.
Step 3a
RNO:
This step was a local action, but was not
identified as such .
Step 3a first and second hollow bullets:
This symbol was not
defined in the WG.
Step 3a
RNO:
This step was a local action, but was not
identified as such.
Step 4 note first and second hollow bullets:
This symbol was
not defined in the WG.
Step 5a5 RNO:
This step required the operator to cross tie to
the unaffected units CHG pump SW system if both of the affected
units CHG pump SW pumps were air bound after only attempting to
vent the affected system one time.
Depending on the volume of
air in the system this could compound the problem by allowing
air to enter the unaffected units CHG pump SW system, poten-
tially air binding the CHG pump SW pumps on the unaffected unit.
g.
Step 6:
This step to read certain values on the ERF computer
did not have an RNO, although some set points were occasionally
not available, including the point selected by the team.
h.
Step 6 first and second asterisks: These actions were preceded
by an asterisk instead of a bullet, which was inconsistent with
the WG .
i .
Step 6 second asterisk first and second hollow bullets:
This
symbol was not defined in the WG.
Appendix 8
51
j.
Step 9 second asterisk:
The minimum required service water
inlet temperature for the co~trol room chiller, 85 degrees, did
not appear to be realistic.
The inlet temperature at the time
of the inspection was 58 degrees.
k.
Step 9 forth . asterisk:
This step addressed the IA system
without specifying the turbine building IA system.
1.
Step 11:
This step required that the operator check
II strainer
delta-P - normal
11 without identifying the appropriate parameters
to check.
m.
Step 12:
This step required that the operator check .suction
pressure -
normal" and
11discharge pressure -
normal" without
identifying the appropriate parameters to check.
n.
Step 121
RNO:
This step addressed the IA system with out
specifying the turbine building IA system.
o.
Step 16:
This step required the operator to "check vacuum
priming pump seal recirc temperature - normal
11 , but there was no
clear method for accomplishing this step .
p.
Step 17 first and second asterisks:
These actions were preceded
by an asterisk, which was inconsistent with the WG.
q.
Step 18 first and second asterisks: These actions were preceded
by an asterisk, which was inconsistent with the WG.
r.
Step 19 first through third asterisks:
These actions we.re
preceded by an asterisk, which was inconsistent with the WG.
s.
Attachment 1 step 1 first through fifth asterisks:
These
actions were prec~ded by an asterisk, which was inconsistent
with the WG.
t.
Attachment 1 step 2 first asterisk: This action was preceded by
an asterisk, which was inconsistent with the WG.
33.
Loss of intake canal
a.
Step one first asterisk:
According to design calculation
ME-166,
the condenser waterboxes must be isolated within a
specific time (70 seconds) to prevent loss of the intake canal
level (ultimate heat sink) to less that that required to handle
a LOCA.
ME-166 also required other actions to be completed
within certain times, such as isolation of many service water
flowpaths within one hour.
These times were not specified in
this procedure.
Also, operators were not aware of these time
requirements.
\\
Appendix B
52
b.
Step 1 second asteri sl<.:
This step required the CW pumps to be
started from the MRC.
With the normal equipment line up, the CW
pump mode switches were in local. With the switches in local it
was only possible to start the pumps locally.
This procedure
step and the associated RNO both required normal power to be
available to start the circ pumps.
If this procedure was
entered as a transition from AP-10.00, neither the AER nor the
RNO could be performed.
c.
Step 3:
This step directed the operator to go to step 22 if the
canal level was stable or increasing.
When directed to step 22,
the operator would proceed to step 23 which directed him to
verify canal level was increasing. If canal level was stable,
step 23 RNO was then applicable. This directed the operator to
continue efforts to restore level.
These efforts were not
applicable since any efforts to restore level were bypassed in
the transition from step 3 to step 22.
Additionally, there were
no procedural steps that would have caused the canal at stable
level to increase.
d.
Step 5:
These actions were preceded by an asterisk instead of a
bullet, which was inconsistent with the WG.
e.
Step 6:
This action contained one. substep instead of being
written as a single step, which was inconsistent with the WG.
f.
Step 10a: This step required the operators to start
able ESW pumps.
Design calculation ME-166 required
pumps be started within a certain time (two hours).
was not specified in the procedure.
all avail-
that these
This time
g.
Step 10b first and second ho 11 ow bull et:
This action was
required to .be performed within a specific time by design
calculation ME-166.
This time was
not specified in the
procedure.
h.
Step 10 first and second hollow bullet:
This symbol was
undefined in the WG.
i.
Step 11 first through eighth hollow bullets: These symbols were
undefined in the WG.
Additionally, the place keeping markers
following each hollow bullet were not in accordance with the WG.
j.
Step 13:
These actions were preceded by an asterisk instead of
a bullet, which was inconsistent with the WG.
k.
Step 15 table left column first entry:
This column was not
clear to the interviewed operators in that the definition of at
power was different than the definition giving during general
training and the definit{?n used in other procedures .
Appendix B
53
1.
Step 17 table left column first entry:
This column was not
clear to the interviewed operators in that the definition of at
power was different than the definition giving during general
training and the definition used in other procedures.
m.
Step 17 table right column case 1 second entry:
This step was
not identified as a continuous action step and had no method of
place keeping to remind the operator of required actions 24
hours later.
This step required that the CCW Hxs with service
water be throttled to a SW dP of between 0.75 and 1.00 inches
per heat exchanger.
This action was required to be performed
within this time by design calculation ME-166.
This step did
not have a method to advise the operator to perform it at the
required time.
n.
Step 22:
These actions were preceded by an asterisk instead of
a bullet, which was inconsistent with the WG.
o.
Step 24
RNO:
This step was not necessary based upon the
accepted operator definition of the verb
11verify
11 as used in
step 24.
p .
q.
Step 26i These actions were preceded by an asterisk instead of
a bullet, which was inconsistent with the WG.
Attachment 2 step 3d5:
level intake structure.
The step referred to a G bus in the low
There was no bus with this designation.
r.
Attachment 2 step 4b:
The equipment required to perform this
step was not available at the low level intake structure, nor
was it transported to the low level intake structure when the
operator went to the structure.
s.
Attachment 2 step 5b:
This step required local river water
level to be checked without a quantitative method of accom-
plishing the task.
34.
Loss of component cooling
a.
Entry conditions:
There were very few conditions leading up to
the loss of component cooling in th~s section of the procedure.
It did not include the related alarm response procedures.
b.
Step 1:
This appeared to be an immediate action step, but it
was not designated as such in the procedure.
c.
Step 3:
This step lacked the detail necessary to insure that
none of the actions necessary to carry out the step would be
overlooked .
Appendix B
54
d.
Step 7:
Same as step 3 above.
e
Step 8:
Same as step 3 above.
f.
Step 12:
Same as step 3 above.
g.
Attachment 3:
This attachment was not referenced in the
procedure and appeared to serve no useful purpose.
35.
Excessive RCS leakage
a.
b:
C.
d.
e.
f.
Entry conditions:
There was no transit.ion step from ARP-B-A-3,
CTMT SUMP HI LVL, to AP-16.00.
Step 4:
This step had a misplaced continuous action step
asterisk.
Step 6 RNO:
This step failed to direct a reactor trip prior to
SI.
Step 10:
This step directed the notification of chemistry that
all primary sampling was secured. The operator stated that he
would dispatch an operator to verify a 11 sarnp 1 i ng va 1 ves were
closed. This step did not direct this action.
Step 18b:
This step described HCV-1137 as the excess letdown
HCV, while step 18f described HCV-1137 as the flow control valve
Step 22
RNO:
This step failed to underline WHEN/THEN and
IF/THEN in accordance with the WG.
g.
Step 23:
This step incorrectly directed checking PDT influent
vice PDT influent level.
36.
Auto start failure of EOG
a.
b.
C.
d.
General:
Asterisks were used throughout the procedure instead
of bullets.
Step 1:
The operators had difficulty in finding the a 1 arms
which were located on Alarm Panels BA2, CG6, and VSP-C5.
Step 14:
This step directed the operator to place the selector
switch in remote but it was labelled
11 auto.
11
Step 10 RNO:
A hose was not readily available to connect the
wall drain tank and the base tank .
\\
Appendix B
55
37.
Emergency diesel generator fails to accept electrical load
a.
General:
Asterisks were used throughout the procedure instead
of bullets.
b.
Step 2: This step did not identify that local verification was
required.
38.
Main control room inaccessibility
a.
Step 4:
This step required the evacuation of the control room;
however, it failed to require the operators to bring APP R key
ring, FCA procedures, and steam tables which were required for
the successful completion of subsequent procedure steps.
39.
Loss of main feed flow
a.
Step 1 RNO:
After a reactor trip, the operator was transitioned
to EP-1, and not to E-0.
b.
Step 9:
This step required the operator to verify SG levels
trending to
11 NOL
11 *
An operator stated that NOL meant normal
operating level.
This was inconsistent with the WG, which
defined NOL as normal operating limit.
c.
Attachment 1 step 1 fifth bullet:
The word across was spelled
11cross
11
Also, asterisks were used in place of bullets.
40.
Response to AFW check valve backleakage
a.
Step 1:
The use of verify in this step was inconsistent wi-th
operator training.
b.
Step 2:
This action was preceded by an asterisk, which was
inconsistent with the WG.
c.
Step 3:
These actions were preceded by an asterisk, which was
inconsistent with the WG.
d.
Step 4c:
This step was not clear on how many valves could be
shut without SNSOC approval.
The operators interviewed answered
zero, one, and three when asked to interpret the number of
valves that could be closed before SNSOC approval was required.
e.
Step 5:
These actions were preceded by an asterisk, which was
inconsistent with the WG.
f.
Step 5 second asterisk: This step was not preceded by a caution
to describe the personnel danger involved in opening this vent -
exposure to steam.
Appendix B
56
g.
Step 6:
These actions were preceded by an asterisk, which was
inconsistent with the WG.
h.
Step 6 second asterisk: This step was not preceded by a caution
to describe the personnel danger involved in opening this vent
exposure to steam.
i.
Step 9:
This step required the AFW pump to be returned to
operable status; however, it had not been declared inoperable.
j.
Step 10 first through third asterisks:
These actions were
preceded by
an asterisk instead of a bullet, which was
inconsistent with the WG.
k.
Attachment 1 AFW header temperature:
This step did not clearly
define the expected operator actions at 165, 185 and 200
degrees.
l.
Attachment 1 required surveillance forth unnumbered step:
m *
The operators gave conflicting interpretations of the statement
11 No additional logging requirements 11 *
Attachment 2 step 6:
This step required that the AFW pump be
cooled, but did not specify the required final temperature of
the AFW pump.
41.
Fuel handling abnormal conditions
a.
Step 3:
These actions were preceded by an asterisk instead of a
bullet, which was inconsistent with the WG.
b.
Step Sb:
This procedure step required that valves VS-103-A and
VS-103-B be placed in the closed position.
The valves had no
indicated closed position.
c.
Steps 10 and 11:
These steps would have to be performed in the
reverse order if t~e affected equipment was needed to put fuel
in the safe condition.
d.
Step 13:
These actions were preceded by an asterisk instead of
a bullet, which was inconsistent with the WG.
42.
Loss of refueling cavity level
a.
Step 3a RNO first and second bul 1 ets:
These bul 1 ets aligned
alternate suction. If the first bullet "LHSI suction from CNTMT
11 was aligned then only step 3b applied and not the step 3b
RNO.
If the second bullet "RWST x-tie from unaffected Unit.
to CHG/SI pump suction" was aligned only step 3b
RNO was
applicable.
These ste~ were written as equally acceptable
steps, but could not be u'sed that way in the AER and RNO format
as they were written.
Appendix 8
57
b.
Step 3c and step 3c RNO:
These steps required the operator to
perform the same step in the AER and RNO column.
c.
Step 7: This step required the evacuation of containment, which
based on current operator training involved sealing of the
containment. However, steps 8, 9, and 11 required some opera-
tions to be accomplished in containment.
d.
Step 10:
These actions were preceded by an asterisk instead of
a bullet, which was inconsistent with the WG.
e.
Step 11:
This step required the determination that cavity
1 eve 1 was restored; however, there was no direct method for
determining the required level.
f.
Step 12:
These actions were preceded by an asterisk instead of
a bullet, which was inconsistent with the WG.
g.
Attachment 1 step 1:
This reference was preceded by an asterisk
instead of a bullet, which was inconsistent with the WG.
h.
Attachment 1 step 2:
This refer~nce was preceded by an asterisk
instead of a bullet, which was inconsistent with the WG.
43.
Loss of spent fuel pit level
a.
Genera 1:
Asterisks were used in the procedure instead of
bullets.
44.
Minor SG tube leakage
a.
Entry condition 1:
This entry condition was an increase in the
activity on the condenser air ejector radiation monitor.
However, the re 1 ated annunciator response procedure did not
direct the operator to AP-24.00.
b.
Entry condition 2:
This entry condition was an increase in the
activity on the steam generator blowdown radiation monitor.
However, control room records indicated that the alert and alarm
setpoints on the blowdown radiation monitors were set at more
that two decades above the current meter readings~
Meter
RI-SS112
RI-SS113
Reading
8.0xlOE2
l.lxl0E2
Alert
1. 2xlOE5
1. 2xlOE5
Al arm
2.4xlOES
2. 4xlOES
Appendix B
58
The licensee's procedures stated that, for process monitors, the
alert setpoint should be two to five times the background
reading, and the alarm setpoint should be 10 to 15 times the
background reading.
The control room records stated that these
alert
and
alarm
setpoints
had
last been
adjusted
on
July 11, 1978, over 11 years ago.
Since 1978, the steam
generators had been replaced and other events had caused the
background reading of these blowdown radiation monitors to vary.
The licensee stated that process radiation monitor setpoints were
not included in a periodic maintenance program.
Operators looked
at the blowdown radiation monitor setpoints daily, but did not
check that they were set at the appropriate levels.
With the
setpoints set so far above background, the alarms would not
function to alert the operator in support the APs or EOPs.
Prior to the end of this fnspection, the licensee stated that the
blowdown radiation monitor setpoints had been corrected and they
were reviewing the initiation of a periodic maintenance program.
c.
Step 1 RNO:
After a reactor trip, this step directed the
operator to go to EP-1.00, and not E-0.
d.
e.
Step 3:
The list of instruments to be used to identify the
affected SG did not include the steam line radiation monitors .
Step 4 fourth bullet:
The applicable valve numbers were not
included for locally isolating the affected SG supply to the
turbine driven AFW pump.
Also, asterisks were used in place of
bullets.
45.
Loss of decay heat removal capability
a.
Step 3 RNO:
The verb
11 verify
11 was used in the step, but no
action statement was included to cover the situation when the
valve was not found in the desired position.
b.
Attachment 2:
This attachment was not referenced in the
procedure and appeared to serve no useful purpose.
46.
Increasing or decreasing RCS/PRZR pressure
a.
Steps 4 and 5:
The location of these steps in the middle of the
procedure caused unnecessary transitions.
b.
Step Sb:
This step required the operator to return all controls
to
11 NORMAL
11 *
However, none of the related contra ls had a
position labeled NORMAL.
\\.
..
Appendix B
59
C.
Step 7 RNO:
After a reactor trip, this step directed the
operator to EP-1.00, and not E-0.
47.
Control room security
a.
Step 1:
This step failed to address obtaining the padlocks
required to lock the doors.
b.
Step 6: * This step contained an improperly placed comment in the
RNO column.
c.
Step 6 RNO:
This step fa i 1 ed to direct
11Use Base Radi 0
11 as done
in the step 3 RNO.
d.
Step 4:
This step failed to address the conduct of* a search of
the MCR.
48.
Seismic event
a.
General:
Asterisks were used in the procedure instead of
bullets.
49.
Abnormal environmental conditions
a.
Step 2:
This step lacked the detail necessary to insure that
all of the areas containing components necessary for safe
operation were listed and placed under appropriate scrutiny to
provide for early detection of potential flooding conditions.
Contingency actions were not developed, in advance, to combat
the most likely flooding situations that could occur.
b.
Step 3 note: The word
11 note
11 was not underlined as required in
the WG.
c.
Step 4 caution:
The word
11 caut i on
11
was not underlined as
required in the WG.
d.
Step 4 note:
Same as step 3 note above.
e.
Step 8 note:
Same as step 3 note above.
f.
Step 9:
This step also covered the situation of either unit on
RHR and did not reflect that fact.
g.
Step 11 caution:
Same as step 4 caution.
h.
Step 12a:
This step referred to an area as *
11 1ow level ,
11 but it
was not clearly indicated as a location at the Intake Structure .
.,
Appendix B
60
i.
Step 17:
This step appeared to be a simple step, but it was in
fact a very involved task. Detailed information to help the
operators perform the step was not included.
j.
Step 21:
The undefined abbreviation "ISO" was used in this
step.
k.
Attachment 1 Step 2:
The step lacked the detail necessary to
insure that none of the actions necessary to carry out the step
would be overlooked.
1.
Attachment 1 Step 3:
The step 1 acked the addi ti ona 1 guidance
necessary to properly carry out the intent of the step.
m.
Attachment 1 Steps 4 and 5:
Same as Attachment 1 step 3 above.
n.
Attachment 1 Step 8: The undefined abbreviation
11 RSD
11 was used
in this step.
o.
Attachment 3 Step 12:
The seventh and ninth door on this list
had the same title. Apparently one of the titles was incorrect
or one of the doors was incorrectly listed twice.
p.
Attachment 4 Step 1:
The verb
11 veri fy" was used in the step,
but no action statement was included to cover the situation when
the equipment/valve was not found in the desired position.
q.
Attachment 4 Steps 2, 3 and 4:
Same as Attachment 4 step 1
above.
r.
Attachment 5 Step 10:
Same as step 21 above.
s.
Attachment 7:
This Attachment was not referenced in the
procedure and appeared to serve no useful purpose.
50.
Natural circulation
a.
Entry conditions:
The one condition listed was very general ih
nature and did not clearly define the conditions when this
procedure vice an EOP would be used.
Two Abnormal Procedures,
AP-37.01 Step 16 and AP 27.00 Step 19, directed the operator to
this procedure, but were not listed as entry conditions.
b.
Step 5:
The words "NARROW RANGE" were missing after the word
"ALL".
C.
Attachment 1 Second bullet after IMPENDING:
The delta
11 P
11 at
the end of the sentence was incorrect -
delta: "T" was needed .
\\
Appendix B
61
d.
Attachment 2:
This Attachment was not referenced in the
procedure and appeared to serve no useful purpose.
51.
Loss of reactor coolant flow
a.
Step 3:
This step was inconsistent with the ES-0.1 step 8b RNO.
b.
Step 5:
This step failed to reference the procedure( s) to be
used for performance of unit shutdown.
52.
Fire protection - operations response
a.
Step 1:
Performance of this step re qui red the ga i-tron i cs
system to be operable.
An RNO step to call out the fire brigade
without gai-tronics available was not included.
b.
Step le:
To be consistent with operator training this step
would have been repeated.
c.
Step 6:
The use of
11 check
11 was not appropriate in this step in
that all non-affected areas would be checked for certain area
fires.
d.
Step 6:
These actions were preceded by an asterisk instead of a
bullet, which was inconsistent with.the WG.
e.
Step 10:
This step only applied. when the RNOs in steps 4 or 5
were applicable.
f.
Step 14:
These actions were preceded by an asterisk instead of
a bullet, which was inconsistent with the WG.
53.
Loss of domestic water
a.
Entry conditions:
Alarms or other system conditions that would
indicate a possible loss of domestic water were not listed in
this section of the procedure.
IV. FCA comments
1.
FCA-1.00 Attachment 2 Local operation of EOG 1 and 2. *
a.
General: Asterisks were used in the procedure instead of
bullets.
b.
Caution:
The use of a scribe mark for a control setting
appeared to be inappropriate.
2.
FCA-1. 00 Attachment 3 #3
EOG alternate fast start
No comments.
Appendix 8
62
3.
FCA-1.00 Attachment 4 Individual transfer of components to auxiliary
shutdown panels
a.
Parts A and B:
These parts incorrectly identified the transfer
switch position AUX PNL.
All 34 local switches were labeled
AUX P.
4.
FCA-1.00 Attachment 6 Communications
a.
Part 8:
There was no step at the *top of this part of the
procedure to direct the operator, who would man the J8 COMM 7
panel, to bring a list of Beeper assignments.
The operator
would need the list to page people from the J8 COMM 7 panel.
b.
General comment:
The security department had an emergency
diesel generator that automati~ally started on loss of AC power
and energized the security portable radio repeaters.
However,
the licensee had no provision for. operators to take advantage of
the security portable radios *to en.able them to have communica-
tions needed to implement EOPs.
C.
General comment:
A sound powered telephone system existed in
the plant and was used by the Instrumentation and Co.ntrol
Department.
No plans or provisions had been implemented for
operators to use this system during pl ant casualties such as
loss of AC power, when communication problems were known to
exist.
5.
FCA-1.00 Attachment 78
Establishing charging pump cross connect
a.
Genera 1 :
Asterisks were used in the procedure instead of
bullets.
b.
C.
d.
e.
Step 4:
This step was no longer necessary because *the valves
were no longer locked.
Step 3:
Valves 21168 and 21150 were not clearly labelled to
assist the operator in quickly locating them.
Step 6:
Valve l-CH-729 was not clearly labelled and the
operator had difficulty finding it. Also it was in a contami-
nation overhead area; the operator needed to find a ladder, and
then it was still hard to reach the valve for operation.
Steps 5 and 7:
Valve l-CH-728 was difficult to operate because
of piping interfe.rence .
\\
Appendix B
63
6.
FCA-1.00 Attachment 8 Alternate steam release
No comments.
7.
FCA-1.00 Attachment 12A
Establishing RCS letdown
a.
Notes Page 1:
All the provisions of these notes were not
incorporated into the body of the two parts of this procedure~
unlike the notes in the EOPs.
8.
FCA-1.00 Attachment 18
Cross connecting emergency busses
No comments.
9.
FCA-1.01
Limiting MCR fire
a.
Step 6:
This step required the evacuation of the control room;
however, it failed to require the operators to bring the steam
t.ables. Additionally, the procedure required only one APP R Key
ring, when two key rings were required to accomplish thi*s
procedure if both units* remote shutdown panels were manned.
b.
Step 10b:
This step required that the MSTVs be pl aced in the
disable position. The switch did not have a disable position -
the appropriate switch position was labeled 11 FIRE EMERG. CLOSE"
c.
Step 10c: This step required that key switches be placed in the
disable position. The switch did not have a disable position -
the.appropriate switch position was labeled 11 EMERG CLOSE 11
d.
Step 16a:
A non-permanent adhesive label was used to indicate
valve position on the HSP.
e.
Step 19:
A non-permanent adhesive label was used to indicate
valve position for the PRZR heater group A and B transfer
switches on the HSP.
f.
Steps 21a, b, and c:
The verb
11 verify
11 was used to indicate
that the first two steps should be accomplished, if possible.
The third step required the operator to verify the same actions
that had been accomplished in the previous steps.
The. use of
the verb verify in these steps was inconsistent.
g.
Step 22:
The controls of the AFW pumps on the remote shutdown
panel were not in consistent order.
The upper level had the
controls arranged
11C11 ,
11A
11 ,
118
11 ,
and
11C
11 *
The lower level
controls were arranged
118
11 and
11A
11 :
h ..
Step 25b first bullet: The procedure step required one SW pump
control switch in hand and one in auto.
The switches were
labeled CHG PP SW PP SW-P-lOA and CHG PP SW PP SW-P-108 .
. "*"* .. * .
--
. *'~
\\ ..... ;*-* ... .- .
Appendix B
64
i.
Step 27:
Steam tables were required to perform this step but
were not available.
j.
Step 33 note 1:
This note included a list of 10 action steps,
which was contrary to the. WG.
These steps were not properly
prioritized and also were not separately included in procedure
steps.
k.
Step 33 note 2:
The prioritized order for reactor coolant pumps
was inconsistent with the priority in other procedures.
1.
Step 33:
The OP referenced for establishing conditions to
start a RCP, OP-5, could not be accomplished by operators when
attempted from outside the control room.
There was no procedure
for accomplishing this function from outside the control room.
m.
n.
Step 54:
This step, which put the overpressure mitigation
system in service, appeared to be impractical since the PORVs
were di sab 1 ed in step 10a for RCS i nte"grity.
If operators
attempted to place the overpressure mitigation system in service
in the normal manner, they would receive indications that it was
in service, when in fact it was not because the PORVs were
disabled .
Various Steps:
The name plate was missing from steam generator
pressure channel Bon the remote shut down panel .
\\
APPENDIX C
WRITER'S GUIDE COMMENTS
This appendix contains writer's guide (WG) comments and observations. Unless
specifically stated, these comments were not regulatory requirements.
However,*
the licensee acknowledged that the factual content of each of these comments
was correct as st_ated.
The 1 i censee agreed to eva 1 uate each comment, to take
the appropriate action, and to document that action.
These items will be
reviewed during a future NRC inspection.
L Inadequacies in the Writer's Guide
The writer's guide did not thoroughly address each aspect of the
procedures nor did it define restrictively the methods designated for use
in order to assure consistency within and between procedures and to *retain
that consistency over time and through personnel changes.
The following
weaknesses led to problems and inconsistencies in the EOPs or allowed for
future inconsistencies in the revision or development process:
1.
Section 6.4.8(a), Transitions,. did not define the use "Do" and
"Continue 11 as used in transitions, however, these terms were used
in numerous cases to direct transitions.
These terms were also
used inconsistently, and consequently they were used both when a
restricted set of steps or substeps were to be completed and also
when the procedure was correctly performed without such restrictions.
2.
Section 6.4.8(c), Referencing, failed to clearly restrict the terms
which were acceptab 1 e for use in referencing other procedures.
Consequently several referencing terms (e.g., IAW) were identified
in the procedures which were not identified* in the guidance on
referencing.
3.
Section 6.4.8 of the writer's guide failed to state if the procedure
name was required in addition to procedure number when procedures
were identified in branching or referencing.
4.
Section 6.4.8(c), Referencing, stated that specific steps of supple-
ment,al procedures should be considered for incorporation into the
original procedure instead of referencing. The writer's guide failed
to provide restrictive criteria for consideration.
5.
Section 6.4.8(a), Transitions, stated:
"If steps that are to be
repeated are not lengthy, then consider repeating the information and
not using the transition." The writer's guide did not establish the
criteria for this decision and therefore was nci~restrictive.
6.
Section 6.4.4 stated:
"Do not include actions within cautions.i' In
practice this guidance was. not fo 11 owed and numerous in stances of
action statements within caut i ans were i dent ifi ed.
Operators were
trained to expect action steps to be numbered and located in sequence
in the AER and RNO columns of the procedure.
Actions that were
located in a caut{on or note were likely to be missed due to operator
expectations about'how the procedures worked.
" ..... ,
.*.**.***
Appendix C
2
7.
The writer 1 s guide stated:
11A step identified in the Westinghouse
ERG background as a Continuous Action Step should be identified by an
asterisk to the left of the step number, unless the procedure text
clearly indicates that it is a continuous action by using continuous
action verbs (e.g., Maintain, Monitor), or is specified in a Note or
a Caution.
11
The team had several concerns with this statement:
(1) The use of the term
11 should
11 was nonrestrictive.
(2) Use of
asterisks to identify Continuous Action Steps helped ensure that
these steps were not overlooked, but this advantage was lost when
steps with continuous action verbs were not marked with an asterisk.
(3) Insufficient guidance was provided as to whether placement of
asterisks was only allowed on high level steps, or was acceptable
for substeps and in the RNO column.
(4) The definition suggested
that the continuous action was impJied in notes and cautions. This
definition appeared to contradict section 6.4.4 of the writer 1 s
guide which* stated that notes and cautions shall not contain action
steps. (5) The use of WHEN THEN logic terms indicated a continuous
action and were not addressed as such in the writer 1 s guide.
8.
Continuous Action Steps were defined as
11 performed any time after
they are presented or are steps that require continuous monitoring
before an action can be accomplished
11 *
This definition included
non-sequential steps_ in the definition of Continuous Action Steps and
could result in confusion and difficulty in procedure control.
9.
Section 6.3.4 stated that equipment mark numbers should be included
as substeps when needed for clarification, the noun name and equip-
ment mark numbers should be used when i dent ifyi ng 1 oca 1 equipment,
and a reference to a piece of equipment should be used consistently
throughout the procedures. The use of the term
11 should
11 rather than
11 shall
11 implied that there were criteria for noncompliance with this
guidance.
The writer 1 s guide failed to state that criteria.
10.
Section 6.4 provided direction for writing action steps.
No guidance
was provided for acceptabi 1 i ty and/or format for action substeps
under high level CHECK steps.
11.
Section 6.4.2 of th~ writer 1 s guide failed to address a method for
i dent ifyi ng an order of preference for alternative actions when
alternative actions were listed in a step.
12.
The writer I s guide stated, II_As a genera 1 rule, branching to other
procedures should take place from the RNO column.
The writer 1 s guide
failed to restrictively define the criteria for when branching to
other procedures would be acceptable in the AER column.
13.
The writer
1s guide stated that
11 an underline space shall bi provided
at the high-level step for a check mark to signify the action of the
step has been completed.
11
This guidance failed to consider place
keeping for portions of proce~ures which were repeated.
, .. -.*,
- .**.**
Appendix C
3
14.
The writer 1 s guide failed to include, in the list of parentheses
usage, the use of parenthesis around the letter
11 s
11 at the end of a
word to indicate that the singular or plural from of the word may
apply.
15.
Section 6.4.4 stated that a caution may be repeated at any time
throughout the procedure but failed to define the criteria for
caution repetition.
16.
Section 6.2.2 stated that
11 the instruction pages shall have a border
on all sides of the page
11 but did not define the size of the border.
II. Deviations from the EDP writer 1 s guide.
A sample of the EOPs was evaluated for deviations from the Surry writer 1 s
guide. Types of deviations noted were characterized in this section and
accompanied by a list of examples of the specific deviations.
Note that
some steps contained more than one deviation.
1.
The following steps used terms for transitions between steps within
in a procedure, or terms for branching or references to other
procedures/references in a manner not specified by the WG.
E-3, Step 35a RNO
E-3, Step 38a2a RNO
E-3, Step 39 fifth bullet
E-3, Stop 8
E-3, Step 31a RNO
FR-C.2, Step 13b RNO
FR-C.2, Step 17 RNO
FR-Z.l, Step 5
FR-Z.l, Step 8a & c RNO
FR-Z.l, Step 9al RNO
FR-Z.l, Step 9b
FR-Z.l, Step 10a2 RNO
FR-Z.l, Step 13
ES-1.2, Step 3b RNO
ES-1.2, Step 13d
ES-1.2, Step 19al RNO
ES-1.2, Step 22a
ES-1.2, Step 24c RNO
ES-1.4, Step 5
2.
The fol lowing steps contained a reference to another procedure but
did not include a reference term.
E-3, Step 34b
ES-1.2, Step 25b
\\
.. ,
Appendix C
4
3.
The following step used words with a meaning important to the action
that were not defined in the WG.
E-3
Step 15a
4.
The following steps contained one or more of the following violations
of the WG: (1) conditional or continuous actions in cautions or
notes, (2) references or transitions in cautions or notes, (3) no
statement of consequences in a caution, (4) a caution in a note.
E-3, Step 3 caution
E-3, Step 4 caution
E-3, Step 5 caution
E-3, Step 12 caution
E-3, Step 15 note
E-3, Step 33 caution
FR-C.2, Step 1 caution
FR-Z.l, Step 1 caution
FR-Z.l, Step 7 caution
FR-Z.l, Step 8 note
ES-1. 2, Step 1 caution
ES-1.2, Step 6 caution
ES-1. 2, Step 8 note
ES-1.4, Step 2 caution
5.
The following steps used asterisks in a manner inconsistent with the
WG.
E-3, Entry conditions
E-3, Steps 3,4 & 7
ECA-1.2, Entry conditions
ES-1.2, Entry conditions
ES-1.4, Entry conditions
6.
The
following
steps used punctuation/underlining in a manner
inconsistent with the WG.
E-3, Continuous action page heading
ECA-1.2, Step lb & c
ECA-1.2, Step 2a,d, & g
ECA-1.2, Step 2 continued
7.
The following step used an acronym that did not appear in the WG
acronym list.
8.
E-3, Step 41 seventh bullet
The following caution used a caution at the bottom of a page when
most of the related steps were on the following page in violation of
WG direction.
\\
FR-C.2, Step 14 caution
. .,.. . . . ..
~ *. : - .. ,. .
APPENDIX D
NOMENCLATURE
This Appendix contains basic nomenclature weaknesses.
For example, instances
where writer's guide application to the EDP would cause the reader to expect an
exact nomenclature match with component labeling, yet there was no match.
It
also includes. instances where a complete match was neither required nor found
and the mismatch or lack of label was sufficient to cause concern.
In addi-
tion, inadequate labeling methods such a use of dymotape or hand written labels
was noted.
The licensee agreed to evaluate each item and make the appropriate.
changes.
These items will be reviewed during a future NRC inspection~
Procedure
FCA-1. 00
FCA-1.01
Step/pg.
part A
part B
10a
19
EDP nomenclature
1-FW-P-3A
1-FW-P-3B
1-CH-P-lA
1-CH-P-lC.
1-CH-P-lB
1-CH-P-lC
2-FW-P-3A
2-FW-P-3B
2-CH-P-lA
2-CH-P-lC
2-CH-P-lB
2-CH-P-lC
PCV-( )455C
LCV-()460A
PCV-()456
HCV-() 137
PRZR HTR
11A
11 BANK
-. :**. .
~ ..
Component nomenclature
11 3B
11
CHARGING pp
II iA"
CHARGING PP
11 lC
11
CHARGING PP
11 18
11
CHARGING PP lC
CUB. 25H4 AUX SiEAM
GEN. FD PUMP A
CUB. 25J4 AUX STEAM
GEN. FD PUMP B
- CUB. 25HS CHARGING
pp A
CUB. 25H6 CHARGING
pp C
CUB. 25JS CHARGING PPB
CUB. 25J2 CHARGING PPC
P~ESS RELEIF VV 1455C
HI PRESS LETDOWN STOP
VV 1460A
PRESS RELEIF VV 1456C
EXCESS. LETDOWN HEAT
EXCHANGER FLOW 1137
PRZR HEATER GROUP A
.... * * .. **:** -*::;,, .,
- -'
Appendix D
2
E-3
4
lbl-343
181-3-43
PRZR HTR 11811 BANK
PRZR HEATER GROUP E
FR-H.4
2c RNO
TCV-MS-205A
TCV-MS-205B
TCV-MS-105A
TCV-MS-1058
lMS-TCV-1058
ECA-2.1
28b
MOV-11158
MOV-LCV-11158
MOV-1115C
MOV-LCV-1115C
MOV-11150
MOV-LCV-11150
MOV-1115E
MOV-LCV-1115E
37
MOV-MS-lOOA
MOV-SD-lOOA
MOV-MS-1008
MOV-S0-1008
MOV-MS-lOOC
MOV-SO-lOOC
MOV-MS-1000
MOV-SD-1000
ES-1. 3
4bl
Phase 1 Status
Status
4cl
Phase 2 Status
Status
7b RNO
Atmos Stm Dump VVs
Att. IA
P-250 analog
Process computer
trend recorders
trend recorders
ES-3 .1
Att. IA
P-250 analog
Process computer
trend recorders
trend recorders
ECA-3.1
5g
Caustic supply
refuel WTR chem add
valves
TK outlt VV a and b
E-1
la
HHsi to cold leg
Cold Leg HHSI line flow
Sa
PRZR PORV block
Relief Line
valves
isolation vvs
13c
caustic sup~,ly
Refuel WTR CHEM ADD
valves
TANK OUTLT vvs
. . .. ~ : *~ . . . . ., . . . ' .
-*
Appendix D
ES-1. l
ECA-0.l
ECA-3.3
FR-H.l
. "*-. *< . ~*--:~ . ;
29b
la
2b
2c
13b
16b
16c
15d
21
1
1
7b
7e
3a
3b
3c
3c
9
3
FW isolation reset
RCP seal injection
isolation valve
Safeguards area
exhaust fan
CROM fans
RCP seal return
valve
CHG pump VCT
suction valves
CHG pump RWST
suction valves
RWST crosstie
1st PT Extraction
steam MOV
Turbine drain
valves
Reheater vents
Permissive Status
light C-21
Permissive Status
light F-1
Purge supply and
exhaust valves
Purge supply fans
Instrument Air
valves
AOV-IA-( )03
Air recirc fans
\\
RCP seal water
injection line
isolation valve
l-VS-F-40A/408
CONT ROD CLG 480v
fans
RCP seal LK OFF
return OTSD Trip vv
CHG pump SUCT FROM
VCT !SOL vvs
CHG PUMP SUCT FROM
RWST ISOL vvs
ISOL vv
Turbine INLT & CROSS
UNDER ORN vvs
Reheater ORN vents
RCVR ISOL vvs
SI BLOCKED PRZ
LO PRESS
LOT AVE SI BLOCKED
STM FLOW & PRESS
RX CTMT valves
CTMT fans
CTMT SUCT vvs
SOV-IA-( )03
CTMT AIR recirc fans
~. :a*,. -..... ," ...... .
- '*,.
- ;
- ~ ~.... . .
..*** :* " .,
. : .* .. _ ...... -.
,. *.' '*'* ~--..
Appendix D
4
E-0
Att. 1
12b
Att.5
step 6
step 8
18d
1
2b
5
9
19
35a RNO
120 RNO
Charging pump CCW
pump
Control room
chillers
Control room AHUs
Shroud cooling
fans*
RCP Oil Cooler
Thermal Barrier
loop drain valve
MRC To Annex
Unit 1 exit to
Stairwell
Unit 2 exit to
Stai rwe 11
PERM
FEED REG valves
CHG pump CC pump
reactor banks
Shroud cooling fans
bypass interlock*
switch
Charging pump CLG
WTR pump
REL & CONT RM WTR
CLR
Control room CHILL
WTR A/C FDR
REL RM CHILL WTR A/C
FDR
CONT ROD cooling
fans
FLOW OTSD TRIP vv
RCP Thermal Barrier
CC OUTLT FLOW OTSD
TRIP vv
LOOP ORN HOR ISOL VVS
CONTROL ROOM DOOR 1
. CONTROL ROOM DOOR 3
CONTROL ROOM. DOOR 4
N
CHRG PP CLG WTR PPS .
34.5 KV Rx CORR
CONT. ROD CLG FAN 480V
STM DUMP CONTR
BYP INTERLOCK
' ,, ' .* , ,**, *.
- '
- - '
,. *, '*. ;*~ -~.r r
- ,* * *
~
'
_-* *. t
- .. , \\ ;*;' ,**,.~ ** :** * * *I ,.
'** '.,',*,
~ * (' *
- ,.
- r' *
- :
..
Appendix D
5
AP-16.00 *
17
CCW head tank
CC SURG TK LVL
level
18b
excess letdown HCV
EXCESS LETDOWN FLOW
HCV-1137
SETPT HCV-1137
1
LW MONITOR
LIQ WASTE DISP
AP-1. 01
5
ROD DROP
DROPPED ROD
AP-1. 00
4
NIS DROPPED ROD
. NIS DROPPED ROD ROD
STOP AND TURB
STOP AND TURB RNBK
RNBK
\\
. . . *t~
. ,, .* '._ ... ;*"* .. , ..
. . : ,. *. '. -*** ** : ... ~-
- . ** .I "'..... * .. \\' *:. ~. *****.' ...... *
.*
- ~* *
. ;
, ... :* ... ~ *- -*- *-
...
AER
ATWX
AUX
BRK
cc
CHG
CLS
cs
CROM
cw
DEV
DP
DP ELEV
EOG
FCA
GPM
IF!
LHSI
MON
MSTV
NI
NR
NRC
NQ
OP
APPENDIX E
ABBREVIATIONS
Alternating Current
Action Expected Response
Abnormal Procedures
Alarm Response Procedures
Anticipated Transient Without Scram
Auxiliary
Breaker
Component Cooling
Charging
Consequence Limiting Safeguards
Corporate Nuclear Safety
Condensate Storage Tank
Control Rod Drive Mechanism
Critical Safety Function Status Tree
Containment
Circulating Water
Direct Current
Deviation
Differential Pressure
Differential Pressure Due to Elevation
Emergency Operating Procedure
Emergency Response Facility
Emergency Response Guidelines
Fire Contingency Action
Flow Control Valve
Gallons Per Minute
Hand Control Valve
Health Physics
Heat Exchaner
Instrument Air
Instrumentation and Control
Inspector Followup Item
Low Head Safety Injection
Loss of Coolant Accident
Motor Control Center
Main Control Room
Monitor
_
Motor Operated Valve
Main Steam Trip Valve
Non-licensed Operator
Nuclear Instrumentation
Narrow Range
Nuclear Regulatory Commission
Not Qualified
\\
Overpressure Mitigation System
Operating Procedure
._.. t *
- **
- ,
.**
,~ *** ,
- * *
' *.' ****** .-*. *,*.
,
- -.. _**.* .. *-*,:*-
- ."7".:
II
/
)
.~
- .~
Appendix E
Pp
PSID
PSTG
PSTG DEV
PRZR
RECIRC
R/hr
RNO
RX
SOD
SNSOC
ss
Tavg
Tc
TD
Th
Tref
TRANS
TS
TURB
VPAP
2
Procedures Generation Package
Power Operated Relief Valve
Pump
Probalistic Risk Assessment
Pressure Square Inch-differential
Pressure Square Inch-guage
Plant Specific Technical Guidelines
PSTG Deviation
Pressurizer
Quality Assurance
Quality Control
Reactor Coolant Pump
Recirculation
Roentgen/hour
Radiation Monitor
Response Not Obtained
Reactor Operator
Reserve Station Service Transformer
Resistance Temperature Device
Reactor Vessel Level Instrumentation System
Refueling Water Storage Tank
Reactor
Station Blackout
Step Dilation Document
Safety Evaluation Report
Spent Fuel Pit
Steam Generator Tube Rupture
Safety Injection
Station Nuclear Safety and Operating
Committee
Safety Parameter Display System
Senior Reactor Operation
Shift Supervisor
Steam
Temperature-average
Temperature-Cold
Thermocouple
Transition Driver
Temperature-Hot
Trend Recorder
Temperature-Reference
Transformer
Technical Specification
Turbine
Verification & Validation
Volume Control Tank
Virginia Power Administrative Procedure
--
. . . . .* .. ~ *. '.
, . . :-.
- . ..~.
. .- ., .. *
. *. --- :'*'
- -
Appendix E
WG
X-Tie
3
Writer's Guide
Westinghouse Owner's Group
Writers Guide
Cross Tie
\\
- -
- , ... -:* .--
- ,* *.** ,.,*-~ ***.* ': .
- * !" ... *:-.
- .:.*, .*,:-*.
.*, ...... _ *****.~