ML20151Q788
| ML20151Q788 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 07/12/1988 |
| From: | Haughney C, Konklin J, Vandenburgh C Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20151Q780 | List: |
| References | |
| 50-333-88-200, NUDOCS 8808110288 | |
| Download: ML20151Q788 (24) | |
See also: IR 05000333/1988020
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U.S. NUCLEAR REGULATOP.Y COMMISSION
OFFICE OF NUCLEAR REGULATORY REGULATION
Division of Reactor Inspection and Safeguards .
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Report No.. 50-333/88200
Docket No.: 50-333
Licensee: Power Authority of the State of New York
123 Main Street
White Plains, New York 10601
Inspection At: James A. Fitzpatrick Nuclear Power Plant
Inspection Conducted: May 23 through June 3, 1988.
Team Leader: Mh .' T-n-t'I
C.A.VanDenbufgh,TeamLeader Date Signed
Consultants: J. D. Wilcox, Jr., Prisuta-Beckman Associates
D. Schultz, Comex, Inc.
D. Jarrell, Battelle-Pacific Northwest Labs
0. Meyer, EG&G-Idaho
Other NRC Personnel Attending Exit Meet .gs: C. J. Haughney, Chief, RSIB, NRR;
J. Konklin, Section Chief, NRR; D. Langford, Project Engineer, NRR; J. Johnson,
Section Chief, RI; A. J. Luptak, Senior Resident Inspector; R. A. Plasse,
Resident Inspector.
Reviewed By: 7// 2./T7
ames E. Konklin, Chief Date S'igned
/SpecialTeamSupport
& Integration Section, NRR
Approved By: 1
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SUMMARY OF RESULTS
Scope:
On May 23 througn June 3, 1988 an NRC inspection team conducted an inspection
of the James A. Fitzpatrick Nuclear Power Plant (JAFNPP) Emergency Operating
Procedures (EOPs). JAFNPP is a BWR-4 with a Mark I containment. The
objectives of the inspection were to determine whether the E0Ps are technically
correct, whether the E0Ps can be physically carried out in the plant, and
whether the E0Ps can be correctly performed by the plant staff.
The team accomplished the inspection by performing a comp" ' the BWR
Owners Group Emergency Procedure Guidelines (EPGs) to the Pa..< Specific
Technical Guidelines (PSTGs); a comparison cf the PSTGs te 'ne E0Ds; a review
of the calculations performed to develop the plant specific curves, values, and
setpoints utilized in the E0Ps; a plant walkdown of all the E0Ps and the
Abnormal Operating Procedures (A0Ps) referenced by the E0Ps; a simulation of
two emergency scenarios using a full size control room mock-up; a human factors
review of the procedures and plant operations, including interviews of nine
licensed and non-licensed personnel who utilize the E0Ps and A0Ps; a detailed
review of the containment venting procedures; and a review of your on-going
program for evaluation of E0Ps. The inspection was primarily focused on the
adequacy of the end prcduct and not on a review of the process to develop the
E0Ps.
Results:
Based on a review of the E0Ps and supporting calculations, the inspection team
concluded that although the PSTGs have not been controlled and maintained
up-to-date as a design basis document, the E0Ps are a technically accurate
incorporation of the EPGs. The plant walkdowr.3, operator interviews and tne
simulated E0P scenario identified several minor deficiencies; however, the team
concluded that the E0Ps can be accurately accomplished using the existing
controls, instruments, and equipment. Based on the human factors review of the
E0Ps, Writer's Guide implementation, plant walkdowns, and the E0P simulation,
the team concluded that the E0Ps and associated procedures have the useability
to provide the operators with an effective accident mitigation tool and can be
correctly performed by the plant staff.
Several concerns were identified which will require further licensee action to
resolve. The most sigr.ificant concerns were:
1.) The E0Ps and the PSTGs had not been maintained a.c a design basis document
and theref ore have not been maintained up-to-date and appropriately
controlled. This resulted in several discrepancies between the PSTGs and
the E0Ps. (Section 3.a and 3.2)
2.) Plant process computer setpoints did not correspond to the E0P entry
conditions and potential confusion existed in the measurement and indica-
tion methodology of suppression pool level. In addition, outstanding
validation comments concerning the suppression pool level measurement
methodology had not been satisfactorily documented. (Section 4.a)
i 3.) In a few instances, information or e uipment necessary for the performance
of the E0Ps had not been provided. Section 4.b)
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4.) The E0P simulation adequately demonstrated that the minimum shift crew
described by Technical Specifications was sufficient to accomplish the
required actions of the E0Ps. However the team could not conclude that
sufficient personnel would be available to accomplish all of the actions j
required in an emergency, such as implementation of the Emergency Plan or l
activation of the Fire Brigade., coincidental with implementati6n of the l
E0Ps. In addition, a method of placekeeping was not used by the-operators i
during the performance of the E0Ps. Placekeeping methods have not been l
utilized during periodic training and were not supported by the ,
procedures. [Section 5.c.1 and 5.c.3)
5.) A response to the Safety Evaluation incorrectly indicated that action ,
statements would not be carried over from one page to another. (Section
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6.) Sufficient guidance was not provided in the E0P for Primary Containment
Control to describe the calculation of the Heat Capacity Temperature
Limit. (Section 6.2.c)
7.) An evaluation had not been performed to demonstrate the capability of the
Standby Gas Treatment System to operate under the anticipated accident
conditions of hign pressure and temperature during containment venting.
(Section 7)
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TABLE OF CONTENTS
EMERGENCY OPERATING PROCEDURE INSPECTION at
James A. FitzPatrick Nuclear Power Station
(Inspection Report 50-333/88200) . ;
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1. IhSPECTION 06JECTIVE......................................... 1
2. BACKGROUN0................................................... 1
3. PROCEDURE REVIEW............................................. 2
S.1 EPG/PSTG REVIEW......................................... 3
3.2 PSTG/EOP REVIEk......................................... 3
3.3 CALCULATION REVIEW...................................... 4
4. PLANT WALKD0WNS.............................................. 5
5. E0P SIMULATION USING CONTROL ROOM M0CK-UP............ ....... 9
6. HUMAN FACTORS REVIEW AND INTERVIEWS.......................... 12
6.1 WRITER'S GU!DE IMPLEMENTATION........................... 12
6.2 OPERATOR INTERVIEWS..................................... 13
7. CONTAINMENT VENTING.......................................... 16
8. Oh-GOING EVALUATION 0F E0FS.................................. 18
9. EXIT MEETING................................................. 19
10. REFERENCES
10.1 PERSONNEL CONTACTE0..................................... 19
10.2 PROCEDURES REVIEWED..................................... 20
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1. INSPECTION OBJECTIVE
The inspection team reviewed the licensee's Emergency Operating Procedures l
(EOPs), operator training and plant systems to accomplish the following
objectives in accordance with NRC Temporary Instruction (TI) 2515/92:
(1) Determine whether the E0Ps conformed to the vendor generic guidelines and
were' technically correct for the James A. Fitzpatrick Nuclear Power
Station.
(2) Assess whether the E0Ps can be physically carried out in th! olant using
existing equipment, controls, and instrumentation, under tre expected
environmental conditions.
(3) Evaluate whether the plant staf f has been adequately trainto to perform
the E0P functions in the time available.
2. BACKGROUND
Following the Three Mile Island (TMI) accident, the Office of Nuclear Reactor
Regulation developed the "TMI Action Plan," (NUREG-0660 and NUREG-0737) which
required licensees of operating plants to reanalyze transients and accidents
and to upgrade Emergency Operating Procedures (E0Ps) (Item I.C.1). The plan
also required the NRC staff to develop a long-term plan that integrated and
expanded efforts in the writing, reviewing, and monitoring of plant procedures
(Item I.C.9). NUREG-0899, "Guidelines for the Preparation of Emergency Opera-
ting Procedures,' represents the NRC staff's long-term program for upgrading
E0Ps, and descr40es the use of a Procedures Generation Package (PGP) to prepare
E0Ps.
The licensees formed four vendor owner's groups corresponding to the four major
reactor types ir. the United States: Westinghouse, General Electric, Babcock &
Wilcox, and Combustion Engineering. Working with the vendor company and the
NRC, these owner's groups developed generic procedures that set forth the
desired accident mitigation strategy. For General Electric plants, the generic
guidelines are referred t.o as Emergency Procedure Guidelines (EPGs). These
EPGs were to be used by licensees in developing their PGPs. Submittal of the
PGP was made a requirement for the James A. Fitzpatrick Nuclear Power Plant by
Confirmatory Order dated June 12, 1984. Generic Letter 82-33, "Supplement I to
NUREG-0737 - Requirements for Emergt.nty Response Capability" required each
licensee to submit to the NRC a PGP wnich includes:
(1) Plant Specific Technical GuidElires (PSTGs) with justification for safety
significant ditterences from tne EPG.
(2) A Plant Specific Writer's Guide (PSWG).
(3) A cescriptico of the program to be used for the verification and
validation of E0Ps.
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(4) A description of the training program for the upgraded E0Ps.
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Plant specific E0Ps were to have been developed that would provide the operator
with directions to mitigate the consequences of a broad range of accident and
multiple equipment failures.
For various reasons, there were long delays in achieving NRC approval of many
of the PGPs. Nevertheless, the licensees have all implemented their E0Ps. To
determine the success of this implementation, a series of NRC inspections are
being performed to examine the final product of the program: the E0Ps. A
representative sample of each of the four vendor types has been selected for
review by four inspection teams from Regions 1, II, III and IV.
An additional 13 inspections have been scheduled at f acilities with General
Electric Nark I type containments. The latter inspections are being conducted
by the Office of Nuclear Reactor Regulation and include a detailed review of
the containment venting provisicns of the E0Ps. This inspection at the James
A. Fitzpatrick Nuclear Power Plant is the first of the 13 Mark I inspections.
3. PROCEDURE REVIEW
This portion of the inspection was performed to determine whether the JAFNPP
E0Ps have been prepared in accordance with the current Procedures Generation
Package (PGP) and the Plant Specific Technical Guidelines (PSTGs). The
inspection team compared Revision 3 of the BWR Owners Group Emergency Procedure
Guidelines (EPGs) to the PSTGs, and the PSTGs to the E0Ps. All differen m
were identified and reviewed to ensure that safety significant deviations were
identified in the PGP and that a documented basis existed for all deviations.
A review of selected calculations was performed to ensure that plint specific
values utilized in the E0Ps are correct and have been performed 1. accordance
with a documented engineering analysis. Section 10.2 of this report lists the
procedures reviewed,
a. Reccrd Control
in the process of reviewing the PSTGs and the E0Ps, the team identified
that the original copies of the E0Ps, EPG Calculations (Appendix C)
and the Plant Specific Technical Guidelines (PSTGs) were all being
temporarily stored in the Operations Department Administrative Office as
uncontrolled documents and were not being upgraded and maintained
up-to-date. HUREG-0899, "Guidelines for the Preparation of Emergency
Procedure Guioelines" indicates, in paragraph 4.4, that the PSTGs are the
primary basis for plant E0Ps and, as such, should be subject to examina-
tion under the plant's Quality Assudance (QA) Program, and are also
required to be accurate and up-to-date. Further licensee action is l
necessary to upgrade and maintain tne PSTGs as required. ;
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A review of Administrative Procedure F-AP-1.4. "Control of Plant l
Procedures," Plant Standing Order PS0-4, "Quality Assurance and Plant '
Operating Records," and JAFHPP Records Retention / Turnover Schedule, Rev.
3.1 indicated that the location and length of storage of the PSTGs and
supporting records were not in accordance with the plant's administrative
requirements. The QA Department had identified similar concerns with
record storage and the timeliness of turnover in QA Audit No. 584 dated
February 20, 1987 and QA Audit No. 646 dated April 22, 1988. However, at
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the time of the inspection, noce of the E0P records had been turned over l
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to Document Control. The initial upgraded (symptomatic) E0Ps were issued
on December 29, 1984. Further licensee action is necessary to maintain
and control these records as required.
3.1 EPG/PSTG Review ,
Four minor differences were identifieo between the EPGs and the PSTGs,'as
detailed below. The team concluded, based on a review of the PSTGs and the
differences identified, that significant discrepancies did not exist and that
there were no adverse affects on the adequacy of the JAFNPP E0Ps. Further
action should be taken to ensure that future revisions or upgrades of the E0Ps
correct these discrepancies,
a. The selection of 1090 psig as the lowest Safety Relief Valve (SRV)
pressure in PSTG step RC/P-2 did not consider the setpoint tolerances of
the SRVs. The EPG basis (Appendix B) indicates that the intent of RC/P-2
is to establish a control band for pressure control of the reactor at
which the SRVs will not cycle. The use of the lowest SRV setpoint
pressure without consideration of the setpoint tolerance of the SRV could
result in cycling the SRV if the reactor pressure is controlled at 1090
psig. Additional operator guidance such as a lower pressure control band
would ensure that reactor pressure is controlled below the pressure at
which the SRV would lift.
b. Other steam driven equipment available at JAFNPP such as Reactor Feed Pump
(RFP) turbines, RFP drains, steam seals, steam jet air ejectors and off
gas heaters have not been incorporated into PSTG step RC/P-2.
c. Amplifying information for Group 1 Isolation signals, such as initiating
conditions and isolation valve identification, has not been provided in
PSTG step RC/L-1.
a. Alternate injection systems which are available at JAFNPP, such as
demineralized water transfer to the Standby Liquid Control test tank, have
not been incorporated into PSTG contingency Cl-1, "Level Restoration."
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3.2 PSTG/EOP Review
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Six minor differences were identified between the PSTGs and the E0Ps, as
detailed below. The team identified that in each example the E0P w6s correct ;
and concluded that the discrepancies were the result of not maintaining the l
PSTGs up-to-date as required. No conditions which wculd adversely affect the
performance of the E0Ps were noted. Further action is necessary to correct
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these deficiencies and to control and maintain the PSTGs appropriately. I
a. F-E0P-1, "E0P Cautions," Caution No. 6 had a different Reactor Pressure
Vessel (RPV) water level specified than the PST6 Caution #6, because tne
E0P was revised af ter an evaluation identified that the drywell tempera-
ture referenced in the PSTG was incorrect.
b. F-EOP-2. "RPV Control (Boron Injection Not Required)," specified 335 psig
for the nigh/ intermediate RPV pressure region vice 300 psig as specified l
in PSTG Contingency Cl-E, because the E0P was revised after the PSTG was
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c. F-E0P-4, "Primary Containment Control," step 2.3 indicated that the
suppression pool scram temperature is 100 degrees F vice 110 degrees F as
l specified in PSTG SP/T-3 cue to a typographical error.
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i d. F-EOP-5, "Secondary Containment Control," Table F-E0P-5.1 listed dif ferent
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instruments and setpoints than specified in PSTG Table 1, because.
Subsequent plant modifications were not incorporated into the PSTGs.
e. F-EOP-5, "Secondary Containment Control," did not have an entry conoition
for Unit Cooler Temperature above 104 degrees F as specified in the
secondary containment control guidelines of the PSTG, because subsequent
plant modifications were not incorporated into the PSTG.
f. F-E0P-7, "RPV Flooding," did not utilize the Residual Heat Removal (RHR)
Keep Fill System to maintain RPV water level as specified in PSTG C6-4,
because the Keep Fill System had not been made operable at JAFNPP.
3.3 Calculation Review
Specific values from the E0Ps were selected for review to determine if the
values were correctly calculated based on the plant specific differences and
the guioance of the EPGs. The team identified that the calculations were
clear, orderly and performed in accordance with the guidance of the EPGs. Any
deviations were noted and substantiated. The calculations were observed to
include values for each of the cautions, steps and curves. However, as noted
in Section 3.a, the calculations have not been controlled as a plant basis
document in accordance with the requirements and guidance of NUREG-0899. The
following calculations were reviewed:
a. Hot Shutdown Boron Weight - This calculation was performed to support tne
values incorporated as part of F-E0P-3, "RPV Control," and documented in
Section 24, Appendix C, page C24-1 of the EPG Calculations.
b. Minimum Alternate RPV Flooding Pressure - This calculation was performed
to support tne values incorporated as part of F-E0P-3, "RPV Control (Boron
Injection Required)," and F-E0P-7, "RPV Flooding," and documented in
Section 21, Appendix C, page C21-1 of the EPG Calculations.
c. Suppression Pool Load Limit - This calculation was performed to support
tne values incorporated as part of F-E0P-2, "RPV Control (Boron injection
Not Required)," F-E0P-3, "RPV Control (Boron injection Required)," and
F-EOP-4, "Primary Containment Control," ano documented in Section 4,
Appendix C, page C4-1 of the EPG Calculations.
d. Heat Capacity Temperature Limit for the Suppressior Pool - This calcula-
tion was performed to support the values incorporated as part of F-EOP-4,
"Primary Containment Control," and documented in Section 3.3, Appendix C,
page C3-1 of the EP6 Calculations.
e. Maximum Drywell Spray Flow Rate Limit - This calculation was performed to
ensure that the evaporative cooling pressure drop was less than t'
permissible design to support a drywell spray flow rate of 10,000 gallons 1
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f. Drywell Spray Initiation Limit - This curve was developed to determine the '
pressure limit wnen spraying at the maximum drywell spray flow rate to
ensure that the reactor building to containment design negative pressure l
differential limit is not exceeded by initiation of drywell or wetwell
sprays. -
9 Maximum Primary Containment Water Level Limit - This calculatio'n~
determined that the maximum containment water level was based on the
highest primary containment vent elevation as opposed to the design
hydrostatic loading at the most limiting containment location.
4. PLANT WALK 00WNS
in order to assure that the E0Ps can be successfully accomplished, the team
performed in-plant walkdowns for all the E0Ps and referenced A0Ps. The team
verified that E0P instrument and control designations were consistent with the
installed equipment and that indicators, annunciators, and controls referenced
by the E0Ps were available to the operators. The location and control of E0Ps
in the Control Room was verified. With the assistance of licensed operators,
the team physically verified that activities which would occur outside of the
Control Room during an accident scenario could be physically accomplished and
that tools, jumpers, and test equipment were available to the operators. The
post accident radiation survey map was reviewed to ensure that remote
operations were not prohibited by environmental conditions. The procedures
reviewed are listed in Section 10.2.
During the performance of the plant walkoowns, the team identified a specific
strength in that, general area and equipment cleanliness was exceptional. The
plant walkdowns also identified several discrepancies which are broken down
into the following six areas.
a. E0P Entry Conditions
The team reviewed the entry conditions and associated instrumentation for
F-E0P-4, "Primary Containment Control," and identified the following
methods for determining suppression pool water level in the control room:
1. Utilizing the Safety Parameter Display System (SPDS) which indicates
13.95 feet normal torus water level at 100% power.
2. Utilizing the plant process computer which indicates level in Wide
Range (+72 in to '-72 in) and harrow Range (+6 in to -6 in).
3. Utilizing instruments at Main Control Board Panel 09-3, which
iacicctes approximately 14 feet normal torus water level at 100%
power.
The team no;ed that a narrow range level instrument which was not
referenced by the E0P was available.on Standby Gas Treatment System Panel
09-75, which is located in the back of the control room. However an
audible local or control room supervisory alarm was not available.
As a result of the review, the team identified that if the SPDS were not
available as inferred by NUREG-0737, Item 4.1.c, the operator's ability to
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identify entry conditions would be hampered because the plant process
computer alarm setpoints were +2.5 and -2.5 inches vice the E0P entry
conditions of 0.0 and -1.5 inches. The team also noted that E0P-4,
"Primary Containment Control," Section A, Item 6 listed drywell average
temperature above 135 degrees F as a condition for entry into the E0P.
Again, if the SPDS were net available as NUREG-0737 Item 4.1.c i_nfers, the
operator's ability to detect E0P entry conditions would be hampered
because the alarm setpoints for plant process computer points M085 and
M086 were 65 degrees F above the E0P entry condition of 135 degrees F
(i.e. 200 degrees F) and the 09-75 Panel did not have a audible local
alarm or a sur ervisory alarm in the control room. Further action is
necessary to revise the computer alarm setpoints to values which support
the E0P entry conditions,
in addition, the team identified a human engineering deficiency which had
the potential for cperator confusion with respect to the indication of
torus water level. Technical Specifications 3.7.A.1.a and 3.7.A.1.b
specify maximum and minimum vent submergence levels of 53 and 51.5 inches
respectively as the Limiting Conditions for Operation (LCO). However, the
E0P entry conditions were specified as 0.0 and -1.5 inches. Although
engineering correlatien exists between the TS LC0 (53.0 to 51.5 inches)
and the parameters monitored by the plant process computer and Panel 09-75
(0.0 to -1.5 inches) and SPOS (13.95 feet), the indication of torus water
level in different units and methods is confusing. Interviews with opera-
ting staff confirmed that the correlation was not immediately apparent.
The licensee has indicated that this deficiency will be resolved with the
implementation of Revision 4 of the EPGs at wnich time all references to
torus water level will be in feet of water in the torus. Additional
action is necessary in the interim to ensure that operator confusion does
not exist.
The team attempted to determine whether inadequacies in the licensee's
program of E0P validation may have contributed to the above discrepanc)
related to tne torus water level. The plant specific validation of the
JAFNPP E0Ps was performed in December 1984 by a shift supervisor in the j
control room. The validation criteria of F-AP-2.2, "Procedures for
Emergency Operating Procedures," Appendix 0, "EOP Validation Checklist,"
were used to perform the validation. The completed checklist included
reference to the inconsistencies concerning the suppression pool level i
instrumentation, however objective evidence in the form of written I
resolution to the validation comment was not available in the licensee's
records. The team determined that val 106 tion comments of a typographical
nature were incorporated into the E0Ps; however, the resolution of coments
requiring engineering resolutions was not apparent. Future action is i
necessary to resolve this discrepancy as well as any other unresolveo '
validation comments,
b. Lack of Equipment or Inf ormation which could Af tect Performance of the
Procedure
F-E0P-4, "Primary Containment Control," Step 4.7, required the operators
to vent the containment in accordance with F-A0P-35, "Post Accident
Venting of Containment," to maintain pressure below the limits of Figures
F-E0P-4.6a and F-EOP-4.6b. The curves were specifically applicable to
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pressure instruments MENS 0R 16-1-PIT-104 and 27-PT-101A/B respectively.
Figure F-E0P-4.6a was not provided in Procedure F-A0P-35. Therefore, the
possibility for operator confusion existed, in that both pressure instru-
ments and their respective figures were not provided for use in F-A0P-35.
The licensee indicated that F-A0P-35 will be modified to control contain-
ment pressure using MENS 0R 16-1-PIT-104. _
During the walkdown of F-A0P-34, "Alternate Control Rod Insertion," the
team identified that tool cabinets containing the tools and equipment
required to perform the control rod withdraw line venting portion of the
procedure did not contain the necessary equipment for handling the venting
components in the anticipated accident environment. Alternate control rod
insertion is accomplished by venting the Hydraulic Control Unit (HCU) vent
valves with a flexible stainless steel hose. Venting even relatively
small flow rates of high temperature primary water would result in a
dangerous two-phase steam-water mixture through the flexible hose. A
caution in the procedure did not appear adequate in view of the potential
adverse impact on the performance of the procedure. As a result of this
concern, the licensee took immediate action to provide safety equipment
(i.e. welder's gloves) in the cabinets. The team also identifieo that the
venting procedure directed the tlexible hose discharge to a floor drain
and that no provision existed for securing the discharge end of the hose
against the reaction loading during venting. The licensee indicated that
a modification has been initiated to fabricate and install permanent drain
connections for this vent procedure.
During the walkdown of F-A0P-43, "Plant Shutdown From Outside the Control
Room," a potential deficiency in the performance of the procedure was
identifiec, in that a delay was experienced when the remote shutdown panel
could not be opened using the on-shift key ring. Subsequent investigation
identified that the necessary key was in the previously opened, staged
equipment box. The inspection team questioned the benefit of using a
separate key for the remote shutdown panel, in that no additional
protection is provided and the potential for confusion is increased.
Further licensee action is necessary to resolve this deficiency.
F-A0P-43, also required the operation of various Emergency Diesel
Generator (EOG) controls. Step D.2.2, which required the verification of
control power availability by checking the indicator lights, did not
indicate which lights would be energized to indicate that control power
was available to the EDG synchronization circuits. The location of the
inoicating lights was not apparent and should be clarified to prevent
confusion,
c. Uncontrolled Operator Aids
As a result of verifying that operator aids posted on plant instrumenta-
tion and control panels were the latest revision ano administratively
controlled, the team identified that Operations Department Standing Order
No. 21, "Posting of Operator Aids," paragraph 7.7, required an annual
review of all operator aids. The licensee was unable to demonstrate that
this review had been performed in 1967 and 1988 and as corrective action
initiated an immediate verification. As a result of this verification,
six operator aids were identified which required revision. None of these
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aids could have resulted in performing the E0P actions 'ncorrectly. The
licensee indicated that the audit of operator aids will be incorporated
into a new surveillance procedure to ensure the audit is accomplished as
required,
d. Referencing Errors in Procedures
The following discrepancies in referencing were identified during the
review and walkdown of the E0Ps. Based on a review of the context and
effect of these deficiencies, the team concluded that their effect on the
ability to adequately perform the E0Ps was minor.
1. F-E0P-5, "Secondary Containment Control," Table F-E0P-5.1,
incorrectly identified area temperature instruments, 23-MTU-202A and
23-MTU-202B (located at Panels 09-95 and 09-96 in the Reactor
Building) as 23-MTU-201C and 23-MTU-2010.
2. F-A0P-15, "Recovery from an Isolation," Paragraph I.C.2, incorrectly
identified the location of alarming Reactor Building radiation
monitors as Panel 09-12. The correct location for the alarmine
function is Panel 09-3.
3. F-A0P-36, "Stuck Open Relief Valve (50RV)," Paragraph A.6,
incorrectly indicated in the last parenthetical note that the
symptoms of a stuck open reliet valve are four energized solenoid
indicating lights. A single energized indicating light is the
correct symptom.
4. F-EOP-4,"Primary Containment Control," Step 5.2, incorrectly referred
to F-EOP-2, "RPV Control (Boron Injection fiot Required)," Step 4 for
emergency depressurization. The correct reference is Step 4.8 and
4.9. A similar reference occurred in Step 5.3.1 of F-E0P-4.
5. F-E0P-2, "RPV Control (Boron Injection flot Required)," Table
F-E0P-2,1, incorrectly referenced step 4 in the third action item
under pressure high / level decreasing. The correct reference is to
Steps 4.8 and 4.9.
A similar reference existed for the fourth action
item in the same section of Table F-E0P-2.1, in that the reference j
should hve been to Steps 4.6 and 4.7 vice Step 4. I
e. Incorrect / Inadequate labelling
The following minor examples or incorrect or inadequate labelling were
identitied. The team concluded that these examples did not adversely
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affect the performance cf the E0Ps. l
1. F-A0P-36, "Stuck Open Relief Valve (50RV)," required the removal of l
fuses at Panel 09-47 in the Relay Room. Four fuses were not labelled
and eight fuses were labelled with temporary marking (i.e. Dynotape).
The licensee indicated that the fuse location prevents permanent
marking of the fuses and that further action would be taken to remove
the temporary markings and ensure that the fuses are adequately
identified by an cperotor aid.
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2. F-0P-37, "Nitrogen Ventilation and Purge; Containment Atmosphere
Dilution (CAD); Containment Vacuum Relief and Containment Differ-
l ential Pressure Systems," Section G.1.b.4, required operation of the
27-M0V-121 valve on Panel 27PCP in the Relay Rocm. The valve switch
was labeled. "Bypass Valve", however the correct name is ' Purge
Exhaust fan Bypass Valve." _
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3. Several additional minor examples of informal marking (i.e. use of
black markers) and temporary labelling (i.e. adhesive labels) were
ioentifieo during the plant walkdowns. The team observed that
specific actions have been undertaken to upgrade the equipment
labelling throughout the plant. Although no examples were identified
which would prevent the procedures from being accurately performed,
further actions are necessary to upgrade the labelling of instruments
ano components. In particular, the Hydraulic Control Unit (HCU) vent
valves used in F-A0P-34, "Alternate Control Rod Insertion," shoulo be
permanently labelled. These valves were identified witn temporary
marking (i.e. magic markers).
5. E0P SIMULATION USING CONTROL ROOM M0CK-UP
To ensure that the E0Ps could be correctly implemented under emergency
cer.ditions, two accident scenarios were developed and conoucted in a 6-hour
session utilizing licensed operators. The accident scenarios were accomplished
to determine whether the E0Ps provide the operators with sufficient guidance
such that their required actions during an emergency were clearly outlined; to
verify whether the E0Ps could cause the operators to physically interfere with
each other; to verify that the procedures did not duplicate operator actions
unless required; and to verify that transitions from one E0P to another or to
other procedures were accomplished satisfactorily,
a. Control Room Mock-Up
Because JAFNPP does not have a site specific simulator and the plant was
operating at full power thereby making extensive control room access
difficult, a full scale photographic mock-up of the control room was
utilized for tne scenario. To ensure a realistic test of the E0Ps using
the control room mock-up, the following provisions were verified:
1. The photographic reproduction replicated the actual control room with
enough fidelity 50 as not to cause confusion or detract measurably
from the ability of the operating crew. To ensure fidelity, the
mock-up was examined by the team and the licensee. Although mincr
differences were identified, the team concluded that the simulation
would not be excessively impaired.
2. Realistic scenarios which preserved a true time line were required to
be developed and executed in a believable manner. The licensee
supported these efforts by providing a reactor analyst who assisted
the NRC operator examiner to produce two scenarios based on an
integrated analysis of computer generated transients. The time
response of the reactor, primary, and secondary parameters were
caref ully followed. The scenarios were then examined by tne i
licensee's training staff to confirm the adequacy and accuracy of the l
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system responses. To implement and control the scenario, a certified
training staff member and the NRC operator examiner functioned as
controllers, providing data input as required by the operators in
accordance with the timed and scripted accident scenarios.
3. Licensed operators were required to simulate actions and responses
based on inputs from the controllers. The shift crew was requested
and fully supported the mock-up requirements to simulate obtaining
data from the appropriate instrumentation and locations. Three NRC l
team members were used to monitor the operators' response and imple- l
mentation of the E0Ps. Based on the operators' actions and the
te6m's review, the use of the mock-up was determined to be a adequate l
simulation for performance evaluation of the E0Ps. l
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b. Scenario Description
The team developed two simulated accident scenarios. Both were selected i
to exercise parallel E0P paths and contingency procedures, with a special
emphasis on reaching and utilizing the containment venting procedure. The
specific paths were designed to invoke PRA-based risk significant operator
actions as a means to demonstrate E0P adequacy.
The first scenario involved the loss of the main condenser with an )
anticipated transient without scram (ATWS) and delayed Standby Liquid
Control System (SLC) actuation. The transient was initiated with an
unrecoverable failure of the main condenser. The ATWS condition was then
required to develop sufficient internal energy in the primary to produce a
containment venting situation later in the scenario. The workability of
the text-type E0Ps in a multiple path sequence was demonstrated by
requiring the operators to simultaneously follow F-EOP-2, "Reactor
Pressure Vessel (RPV) Control," for reactor power, reactor pressure, and
reactor level control. As alternate rod insertion techniques were being
pursued, primary containment precedures and later SLC injection (with
SQUIB valve failure) were entered due to increases in torus temperature.
Power control with vessel level was then utilized to minimize power while
alternate SLC injection methods were tried. On approaching the
containment venting pressure, alternate rod insertion (withdraw line
venting) and CR0 boron injection were successfully implemented to allow
reactor shutdown and plant recovery. Following containment venting,
shutdcwn cooling was established and the scenario was terminated.
The second scenario involved an unisolable LOCA in the Secondary
Containment. This sequence postulated that the inboard isolation valve in
the secondary containment portion of the Reactor Water Cleanup (RWCU)
system f t 51ed to close. No alarms sounced initially, but parametric
values indicative of a 0.1 square f oot breach of the primary system were
given to the operators. This condition was intended to initiate a problem
solving mcde (i.e. an event based evolution) prior to any E0P entry ;
c/ dition. Area high temperature, and high radiation alarms were used to )
e. ;ablish secondary containment and radiation control E0P entry with
feergency Plan actions. Secondary containment conditions were severe
enough that emergency depressurization was required to minimize the
radioactive release. Rapid plant shutdown and cooldown proceeded with all l
plant components operating normally. l
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c. Observations and Conclusions
The operators' cooperation under the difficult simulation circumstances
involving the control room mock-up was excellent. Cornmunications between
crew members during the evolution were clear, and overall response to the
scenario situations resulted in moving through a very complex scenario
without any significant procedural errors. The scenario timeline was
maintained with a deviation of less than five minutes at the one hour
point, which resulted in a realistic event sequence. Under these real
time conditions, the useability of tne E0Ps was demonstrated to be satis-
factory. The shif t from event based (i.e. abnormal operating parameters)
to condition based E0Ps in the second scenario was made smoothly, com-
pletely and without hesitation, as was the transition from secondary
containment centrol to the emergency depressurization sequence. Although
no significant procedural errors were evident, the following concerns were
identified by the team.
1. Placekeeping Method - The team was concerned with the method utilized
for placekeeping curing the performance of the E0Ps. The operators
used multiple loose leaf copies of the procedures to follow the
multiple paths required. Through direct observation and interviews,
the team determined that loose leaf procedures would not be used in
the control room. The control room copies are of a "lay flat" design
and two copies of the E0Ps are maintained to support multiple entry
conditions. Although the operators did not lose their place in the
procedure during either scenario, the team was concerned that the
Plant Specific Writer's Guide, Operating Department Standing Order s,
operator training and the E0Ps themselves did not provide or support
a preferred method of placekeeping. Further action should be taken
to evaluate and identify a preferred method of placekeeping. This
method should be procedurally supported and trained on a periodic
basis.
2. Use of Cautions - The team was concerned that at no time during the l
performarice of the scenario were the operators observed to review
F-EOP-1, "EOP Cautions." This E0P contained 23 Cautions, whicn were '
referenced throughout the E0Ps. Specific reference to these cautions
were made by numeric and abbreviated reference. Althougn none of the l
specific requirements of this E0P were violated or overlooked during l
the scenario demonstrations, the team remained concerned that the
collection of all cautior.s into one location outside the normally
performed flow path of the E0Ps could result in overlooking a
significant caution.
3. Minimum Snif t Staf fing - Corwnunications outsioe the control room
Tdispatener, plant o.anagement, NRC, and Emergency Plan notifications)
were not simulated in the first scenario due to the circumstances of
the control room mock-up. Corrcunications were adequately simulated
in the second scenario, and crew communication was excellent. The
Emergency Plan was not available at the simulator, but was called for
by the operators. The E0P simulation demonstrated that the minimum
shift crew could imGement all steps nf the E0Ps. However, the team
was concerned that sufficient control room personnel might not be
available to concurrently perform all required actions, including
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implementation of the Emergency Plan and activation of the Fire
Brigade.
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The operatnrs confirmed that their.first responsibility was to
perform the E0Ps and that additional actions would be required to be
performed by personnel not specified in the minimum shift _ crew (i.e. ;
security guards). In addition, the operator interviews (discussed in l
Section 6.2.a) indicated that the operators believed that the present
staffing levels were adequate. The team remained concerned that all
actions required to be performed in an emergency would not be able to
be accomplished by tne minimum shift crew defined in Technical
Specifications. Further licensee and NRC action is required to
resolve this concern.
6. HUMN FACTORS REVIEW AND INTERVIEWS
In order to determine the adequacy of the E0Ps witn respect to the guidance
provided in NUREG-0899, "Guidelines for the Preparation of the Emergency
Operating Procedures," a review of the Plant Specific Writer's Guide (PSKG) and
tne E0Ps was performed to determine the extent to which the PSWG nas been
implemented. In addition, structured interviews were conducted with relevant
JAFNPP personnel. The results of these efforts are detailed below.
6.1 Writer's Guide Implementation
Administrative Procedure F-AP-2.2, "Writer's Guice," was reviewed to ensure
that the human factor's guidance provided was incorporated during the develop-
ment of the E0Ps. Two specific strengths and four weaknesses were noted:
a. Paragraph 4.3, page 27, provided concise, distinctive specifications for
the content of Cautions and Notes which was similar to the guidance of
paragraph 5.7.9 of NUREG-0899. This application of human factors
principles was used by JAFNPP to transform Cautions #4, 5 and 10 of the
EPGs into either Notes er Action statements in the E0Ps. This reduction
in the total number of cautions creates a reduction in the eperator's
buroen and is considered an improvement in the clarity of the E0Ps.
b. Tne use of miniature figures within the body of the E0Ps was identified as
an innovative method to minimize branching outside the procedure withcut
reducing the technical adequacy of the references. As identified in
Section 6.2.e, the operator interviews confirmed that whenever extrapola-
tions were required, the full sized figures attached to the E0Ps were
used.
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C. JAFhPP's response to comment B.1 of tne Interim Safety Evaluation dated l
September 11, 1985 states that an action step will be completed on tne l
page where it begins. A review of tne E0Ps identifleo that operator '
action statemer,ts were oftEN Continued from one page to another witn no
consister4 in format. This practice sometimes resulted in part of a l
logic statement on the first page and the remainder on the second page. '
For example, F-E0P-4, step 4.4 (pages 21-23), step 4,6 (pages 23-25) and
step 5.3.1 (pages 31-35) each contained examples or three different
formats. Beseo on the response to the safety evaluation and t;ie operator
ccncerns retarding continuity and placekeeping identified in Section
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6.2.g the Writer's Guide shculd be revised to provide a human factored
format for the continuation of action statements to the following page
when completion of the action statement on the first page is not possible.
In addittoa, the incorrect response to the SER comment should be
identifieo and resolved. .
d. Paragraph 4.8, page 29, specified that capitalization will be use'd for
emphasis in specifted instances. A review of the E0Ps indicated that the
term "upper case type" is prob'bly meant instead of "capitalization" in
the Writer's Guide and only the action statements within the contingency
statements were to be in upper case type. Tne Writer's Guide should be
clarified to provide consistent guidance, in addition, Caution No. 22 of
F-EOP-1, "EOP Cautions," should be changed to conform to the requirements
of paragraphs 4.2 and 4.8, such that the "if" logic term is in upper case
and is located at the beginning of the logical condition.
e. Paragraph 4.6, page 28, concerning the referencing and branching to other
procedures or steps, provided no guidance as to how referencing or
branching to Abnormal Operating Procedures (A0Ps) or Operating Procedures
(ops) should be handled. The review of the E0Ps identified that the prac-
tice utilized within the E0Ps was not consistent, in that the A0Ps or ops
were not always referenced by both procedure number and name. NUREG-0899,
paragraph 5.2.2 indicates that the specific system procedures should be
referenced in the E0Ps.
t. Paragraph 4.7, page 26, indicated that the equipment names referenced in
the E0Ps may not always match the engraved names on the panels, but will
be complete and in operator language. Althougn there were several minor
examples identified ouring the plant walkdcwns in which equipment names
did not match their labels, on-going labelling efforts, identified in
Section 4.e. are anticipated to correct this discrepancy. Further action
should be taken to revise the Writer's Guide to reflect the current
labelling philosophy.
6.2 Operator Interviews
interviews 'were conducted by the human factors member of the team with
individual' members of the plant staff as classified below:
ifob Classification (License) Number
$nift Supervisors (SRO) 2
Asst. Shift Supervisors (SRO) 3 t
Senior Nuclear Operators (RO) 1
Nuclear Control Operator (k0) 1
Auxiliary Operator (unlicensed) 1
Training Coordinator (SR0 certifieo) 1
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A four page interview guide with 8 major topics was used for each interview and l
was reviewed by both parties. Discussions were open-ended, in that the
licensee representative was encouraged to volunteer comments which were
relevant. Each person was advised that the objective of the interview was to
develop information on the effectiver.ess of the E0Ps and not to examine the.
qualifications of the individual. The length of the individual interviews were
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approximately one hour. Two majo changes were in progress at JAFNPP which the
operators anticipate will have a positive impact on the effectiveness of the
E0Ps. The first is the stort-up of a site specific training simulator late in
1988 and the second ;s the change to a flow chart E0P format in 1989. However,
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the interviews were confined tc the context of the presently implemented E0Ps.
The results of this process are identified below. -
a. Role / Task Definition - There was an established, uniform practice for the
conduct of plant operaticns during the execution of E0Ps. These
operations were governed by Operations Department Standing Order No. 2,
"Operating Principles ano Philosophy," and further developed during
control room crew team training during simulator and on-the-job training
(0JT). Ef fective execution of the E0Ps was also aided by crew stability
(typically crew membership has been the same for 3 years) and by a
recently initiated training program on effective oral communications. The
control room task assignments were well defined. The consensus of
opinions was that the current control room staffing level achieved the
proper balance between assuring adequate staffing and avoiding confusion.
At JAFNPP an off duty control room crew was assigned to "standby" and may
be called to the site upon the declaration of an Unusual Event. However
the avail 6bility of off duty personnel does not resolve the team's concern
with the minimum shift crew manning as discussed in Section 5.c.3.
b. Use of E0Ps - The control room resources for use of the EUPs were
considered adequate. Two sets of E0Ps were kept ir he control room. A
space was assigned for the lay-down of the "open-fl ' E0P manuals. A
cart was provided for abnormal and normal operating procedures, Technical
Specifications, Emergency Plan, etc.
c. Technical Adequacy - Concerns were expressed about the ability to reliably
execute, under accident conditions, Step 5.2 of F-EOP-4, "Primary Contain-
ment Control," which required the combined use of Figure F-E0P-4.1, "Heat
Capacity Temperature Limit," and Figure F-E0P-4.7, "Heat Capacity Level
Limit." There was no explicit direction in Step 5.2, or in the E0Ps
overall, as to how to obtain the value of the abscissa for Figure
F-EOP-4.7. Paragraphs 5.6.9 and 5.7.8 of NUREG-0899, "Guidelines for the
Preparation of Emergency Operating Procedures," provide guidance which
suggests that step-by-step direction should be added to F-EOP-4, Step 5.2,
for the proper combined use of the two curves. This direction would
improv[ethehumanreliabilityassociatedwiththeoperationofthesafety
systems affected by Step 5.2.
Concerns were expressed about the ability within the time available to
bypass the low RPV water level Main Steam Isolation Valve (HSIV) isolation
interlocks in accordance with F-A0P-38, "EOP isolation / Interlock
Overrides," when directed by step 2.2.2 of F-E0P-3, " RPV Control (Boron
injection Required)." The operators suggested that F-EOP-3, "RPV Control
(Boron Injection Required)," should specify controlling RPV water level at
e control pcint above the automatic MSIV isolation setpoint to prevent
unnecessary isolation of the RPV (and loss of the ultimate heat sink)
during the installation of override jumpers.
d. Use of Cautions - The responses of the persons interviewed indicated that l
the first three cautions of F-E0P-1, "EOP Cautions," were not readily !
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recalled and were therefore not fulfilling their intended function.
F-EOP-1, contained all cautions applicable directly to the E0Ps. The
cautions were listed by serial number and consisted of a title ano text.
Numbers and titles of the cautions were used throughout the E0Ps when
reference to the specific caution was required, however the first three
cautions were applicable to all E0Ps and were consequently not referenced
within the texts of otner E0Ps. These cautions required the operators tc
monitor overall plant conditions, nicnitor multiple indications and to
confirm safety functions of automatic equipment. This concern is not to
st6te that the requirements of the cautions would not be applied by the
operators in the performance of the E0Ps, however the effectiveness of the
use of a separate volume of cautions (as discussed in Section 5.b.2) is
questionable.
The effectiveness of the E0P Cautions would be enhanced if the texts on
some of the associated figures were more explicit. For example, label the
cross-hatched areas as "Prohibited Region," or label the figures associa-
ted with Caution # 8 with a directly worded caution such as "Do not
operate pumps unless..." instead of the non-specific, "Observe NPSH
Limits...".
There was a general concern among the operators over the number of
cautions. The PSTGs must be observeo with respect to the incorporation of
tecnnical restrictions into the E0Ps, however application of human factors
guidelines has reduced the burden on the operator of several technical
restrictions (as discussed in Section 6.1). As previously identified,
Cautions # 4, 5 and 10 of the EPGs were not incorporated as caution
statements into the E0Ps because they did not meet the criteria for
cautions in NUREG-0899, paragraph 5.7.9. Similar further application of
the criteria of accuracy, conciseness, and consistency, could reduce the
impact of a large number of cautions,
e. Miniature Figures - The interviews confirmed that the use of Miniature
figures within tne E0Ps were used by the operators to the extent that the
figures can be safely interp-eted. If the small size of the figures ,
caused any doubt, the full size figures of the E0Ps were used. The
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miniature figures did not appear to create any additional possible error
mechariism and the operators considered them useful aids,
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f. Need for a Basis Document - The interviews identified the operator's
desire to add the basis for the opEdating limits of the E0Ps into the l
procedures. Although the addition of this information would generally '
interfere with the clarity of the operating instructions and is therefore
rot recommenced, the need for this information in some form is apparent.
The basis of the operating limits were supported by operator training,
nowever the identification of a neea for a "basis document" underscores
the requirements (identified in Section 3.a) for a reference document
which is traceable ar,d maintained up-to-date,
g. Transitions and Placekeeping Methods - The interviews indicated that the
metnoa of hanaling transitions witnin the E0Ps as well as place-keeping
within the E0Ps and the reliability of the metnods utilized has been
receiving considerable attention by the JAFNPP staff. A standard method
of placekeeping has not been developeo or trained. Each crew was free to
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implement its own method of placekeeping. The use of separate binders and
color ccding of the E0Ps, as well as the "lay flat" capability were the
operator aids in use. Based on the suggestions for additional methods,
the activity involved to improve tnis ability, and the concerns identified
in Section 5.c.1, further consioeration should be given to revising the
E0Ps to include support for check-off spaces and the adoption of- a
practice of writing directly in the action copy of the E0Ps.
h. Communication - The interviews indicated that the present communications
methods were adequate both within the control room and to the local
stations, but that a recent program for improving communications w;s
appropriate and productive. Expectations for improvements in
communications training due to the use of the site specific simulator were !
also noted.
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i. Contro! Rocm Environraert - The interviews identifieo that the provisions l
for lignting witnin the control room and at local stations in the event of
a station blackout and the provisions for control room habitability in the
event of an on-site raoiation release or in the event of smoke or toxic j
gasses in the control room have been difficult to incorporate into 0JT l
training and that further training with the site specific simulator should I
be pursued. '
j. Balance of Plant / Local Control Stations - The interviews identified the i
use of an "operator aids" program wnicn produced validation and ;
improvement of the human f6ctors at local stations including significant i
efforts to upgrade the labelling of components. In summary, the
implementation of E0P operations at local control stations was considered
adequate.
k. Validation and Verification - The interviews indicated that the operators
were includeo to a limited extent in the process of developing, verifying,
and validating the E0Ps. Operators were aware that there was a procedure
for initiating suggested upgrades to the E0Ps and that training exercises
were expected to help identify possible discrepancies. However, operators
were unclear as to their personal responsibilities for initiating
resolutions of possible discrepancies as evidenced by several instances in
which potential discrepancies identified by the operators in the interview
had not been formally identified by the operator for resolution. Further
action is necessary to clarify the operator's role in E0P upgrades.
7. CONTAINMENT VENTING
The team reviewed the EPCs and the Appendix C, Calculation Procedure No. 14,
"Primary Containment Pressure Limit," to determine if the PST6 values were
computed correctly. The team also reviewed tne method, flow path, and feasib1-
lity of the containment venting procedure.
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The attributes of the vent paths are detailed in Calculation Procedure No. 14
"Primary Containment Pressure Liniit," and are summarized as follows:
Suppression Chamber Vent Paths
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Path A(v?) P(oi) P(ci) E(vi) --
27-A0V-117,118 0.736 79.3 79.3 29.5
27-MOV-117,123 0.021 56.0 56.0 29.5
Drywell Vent Paths
Path A(vi) P(oi) P(ci) E(vi)
27-A0V-113,114 0.697 9.2 9.2 105
27-MOV-113,122 0.021 56.0 56.0 105
A(vi) - Minimum vent path area (f t2)
P(oi) - Maximum containment pressure the vent valve can open
against(psi)
P(ci) - Maximum containment pressure the vent valve can
close against (psi)
E(vi) - Elevation of the vent patn containtnent penetration
referenced to torus bottom (ft)
From the above information, JAFhPP determined that the suppression chamber vent
path via A0V-117 and A0V-118, is the only path capable of operating at the
design pressure of the containment and meeting the criteria of removing decay
heat.
Venting of the containment was controlled by F-A0P-35, "Post Accident Venting
of the Primary Containment." Two flow paths are possible: a small bore path
and a large bore path. Containment venting would be initiated via the small
bore path, and if not effective in restoring pressure to less than the limit,
venting would be continued througn the large bore path. The initial path is
via valves MOV-117, M0V-123 and M0V-121. Due to piping and valve size and a
flow restricter between valves MOV-117 and MOV-123, this flow path will not
control the containment pressure under accident conditions.
The large bore path is via valves A0V-117, A0V-118, and MOV-120. Valves
A0V-117 and A0V-116 are 20 inch valves that discharge tc a 30 inch carbon steel
header. Valves M0V-121 anc h0V-120 are 6 inch and 12 inch valves respectively
which discharge in parallel through a 24 inch carbon steel header to one of two
Stacdby Gas Treatment Systein (SBGT) trains. Tne SBGT trains are tested at 1
psig, and have a working pressure of 0.5 psig. The vent patn starts near the
top of the torus. When torus level increases to 29.5 feet, the vent path
becomes unusable due to flooding. An alternate path from the drywell air space
is not available under hign pressure conditions because isolation valves in
" potential vent paths are not able to be stroked (closed) under high
differential pressure conditions.
The team confirmed that figure F-E0P-4.6, "Primary Containment Pressure
Limits," was a result of considering the maximum constant pressure condition
during the air purge and during tne steam vent. During air purge, constant
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pressure is the result of volumetric air flow out of the vent being equal to ;
the volumetric addition of steam generated by decay heat. During the steam I
vent, the steady-state condition exists where energy out of the vent is equal i
to energy generated by decay heat. The limiting structural component in the
primary containment is the 48 inch manway to the torus. For any conditions of
vacuum that may develop in the torus or drywell, vacuum breakers are-provided
between tne torus and the drywell and between the torus and the reactor
building.
Calculations were not included as part of the EPG calculations to determine the
pressure which the SBGT system would be subjected t under venting conditions.
The SBGT filter units are located in an enclosure adjacent to the reactor
building and 1solated from the environment by a non-seismically qualified door
with ventilation louvers. Without further procedura l precautions or hardware
modifications, it is possible that the SBGT train would rupture due to the high
pressure steam being vented. In the event of a f ailure, the vent path would
then release into the environment via this unmonitored path througn the SBGT
room. Further evaluation is required to ensure that the SEGT train is not
anticipated to rupture under the postulated pressures and temperatures
associated with the containment venting sequence.
The vent paths discussed above pass through readily iccessible portions of the
secondary containment. Although venting would result in increased radiation
levels, the team concluded that the operators could carry out other duties
simultaneously with venting.
F-A0P-35 woulo be clearly implemented by control room operators without further
direction after the accident mitigation strategies of the E0Ps have failed. ;
The operators were directed by the procedure to vent, "... irrespective of
radioactive release." Appropriate cautions were included concerning
implementation of the Emergency Plan ano Dose Assessment. The team identified
one concern with the implementation of the containment venting procedures, in
that a Special Procedure, F-SP-02, "Post LOCA Venting of Containment &
Operation of the Main Steam Leakage Collection System," was identified to be an
active plant procedure. The vent paths described in F-SP-02 were identical to
those specified in F-A0P-35, however the initiation pressure for containment
venting was approxim6tely 45 psig lower than the pressure specified in
F-A0P-35. The licensee indicated that the containment venting portion of this '
procedure was based on old event-based operating procedures and was l
inadsertently issued during the last revision of the section applicable to the
operation of the Main Steam Leakage Collection System. During the inspection
period, F-SP-02 was withdrawn and revised to remove tne non-applicable portions 1
of containment venting.
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8. ON-G0ING EVALUATION OF E0Ps
hUREG-0899 Paragraph 6.2.3, indicates that licensees should establish a program
for on-going evaluation of the E0Ps. This program should include: evaluations
cf the technical adequacy of the E0Ps in light of operational experience and ,
use, training experience, and any simulator exercises and control room '
walk-throughs; evaluation of the organization, format, style and content as a
result of using the procedures during cperations, training, simulator
exercises, and walk-throughs; and evaluation of staffing and staff
qualifications relevant to using the E0Ps.
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The team reviewed the Administrative Procedures which control the use of l
procedures at JAFNPP. F-AP-1.4, "Control of Plant Procedures," established the
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requirements for initiation, review, approval, revision, temporary change, I
withdrawal, and control of procedures and was applicable to all operational i
procedures. F-AP-1.2, "Plant Operating Review Comiittee," specified a schedule
for periodic review of procedures, which included a biennial review of EOPs.
Plant Standing Order No. 28, described the procedure by which internally and
externally generated operating experience is evaluated, reviewed and, if
necessary, incorpcrated into plant procedures or design changes. Although each
of these progran.s or procedures was considered to be applicable to tne E0Ps, a
specific program for the on-going evaluation of the E0Ps did not exist. As a
future upgrade of the E0Ps, the licensee is in the process of upgrading the
E0Ps to Revision 4 of the Vendor's EPGs and has scheduled verification and
implementation of flow chart E0Ps by July 1989.
9. EXIT MEETING
The inspection team conducted an exit meeting on June 3,1988, with licensee
management to identify the inspection fir. dings and provide the licensee with an
opportunity to question the observations. The scope of the inspection was
discussed and the licensee was informe' of the conclusions identified in the
course of the inspection. Mr. C. J. Haughney, branch Chief, Special Inspec-
tions Branch, NRR, and Mr. Jon Johnson, Section Chief, Division of Reactor
Projects, Region I represented NRC management at the final exit meeting.
10. REFERENCES
10.1 Personnel Contacted
A large number of personnel were contacted during the inspection. The
following is a list of the JAFNPP personnel involved:
- R. Converse, Resident Manager
"W. Fernandez, Superintendent of Power
- D. Lindsey, Operations Superintendent
- 0. Burch, Reactor Analyst Supervisor
- R. Patch, Quality Assurance Superintendent
- V. Walz, Technical Services Superintendent
- 0. Simpson, Training Superintendent
- J. Catella, liuclear Training Manager
P. Brozenich, Shift Supervisor
R. Pike, Asst. Shif t Supervisor
G. Davis, Reactor Operator
L. Shaffer, Reactor Operator
G. Frank, Training Department
D. Jonnson, Waste fianagement General Supervisor
J. Lazarus, Assoc. Plant Engineer
K. Moody, Plant Engineer .
0. Ruddy, Plant Engineer Supervisor l
G. Tasick, Quality Assurance Supervisor l
B. Robinson, Quality Assurance Engineer l
J. Prokop Jr., Quality Assurance Engineer .
O. Squires, Shitt Supervisor '
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R. Thomas, Assistant Shift Supervisor
W. Hendricks, Reactor Operator
- Denotes those present at the Exit Meeting on June 3, 1988.
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10.2 Procedures Reviewed _
F-EOP-1, "E0P Cautions," Revision 3
F-EOP-2, "RPV Control (Boron Injection Not Required)," Revision 1
F-EOP-3, "RPV Control (Boron Injection Required)," Revision 1
F-EOP-4, "Primary Containment Control," Revision 1
F-EOP-5, "Secondary Containment Control," Revision 1
F-EOP-6, "Radioactivity Release Control," Revision 1
F-E0P-7, "RPV Flooding," Revision 1
F-AP-1.4, "Control of Plant Procedures," Revision 7
F-A0P-15, "Recovery from an isolation," Revision 9
F-A0P-33, "Alternate Shutdown Cooling," Revision 0
F-A0P-34, "Alternate Control Rod Insertion," Revision 0
F-A0P-35, "Post Accident Venting of the Primary Containment,"
Revision 0
F-A0P-36, "Stuck Open Relief Valve," Revision 3
F-A0P-37, "Boron Injection Using the CRD System," Revision 0
F-A0P-38, "E0P Isolation / Interlock Overrides," Revision 3
F-A0P-43, "Plant Snutdown from Outside the Control Room,"
Revision 8
F-AP-1.2, "Plant Operating Review Committee", Revision 4
F-AP-1.4, "Control of Plant Procedures", Revision 4
F-AP-2.2, "Procedure for Emergency Operating Procedures,"
Revision 5
F-SP-2, "Post LOCA Venting cf Containment & Operation of the Main
Steam Leakage Collection System," Revision 8
JAFNPP Records Retention / Turnover Schedule, Revision 3.1
JAfhPP Emergency Procedure Guide, Revision 3
0050-2, "Operating Principles and Philosophy," Revision 3
0050-4, "Posting of Operator Aids," Revision 1
P50-4, "Quality Assurance & Plant Operating Records," Revision 3
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