IR 05000280/1991023
| ML18153C732 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 08/29/1991 |
| From: | Crlenjak R, Mellen L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18153C731 | List: |
| References | |
| 50-280-91-23, 50-281-91-23, NUDOCS 9109160023 | |
| Download: ML18153C732 (13) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTA STREET, ATLANTA, GEORGIA 30323 Report Nos.:
50-280/91-23 and 50-281/91-23 Licensee: Virginia Electric and Power Company 5000 Dominion Boulevard Glen Allen, VA 20360 Docket Nos.:
50-280 and 50-281 DPR-37 License Nos.: DPR-32 and Facility Name: Surry Inspection Conducted:
July 29 - August 2, 1991 Inspector:
Approved Scope:
()
M&
(\\(\\
..,. *iv~--
~- Mellen, Team Leader Date Signed Team Members:
L. King E. Lea Accompanying Personnel:
Daie signed SUMMARY This was a special announced inspection in the area of EOP followu The inspection reviewed the resolution of comments and the resulting EOP changes.*
Results:
The team found that the licensee had well documented and well prepared responses to previous inspection finding Additionally, the team concluded that the EOPs were adequate to mitigate the broad spectrum of accidents listed in the Westinghouse Owners Group Emergency Response Guideline A previous concern identified the omission of a procedure to mitigate the consequences of the unavailability of the reactor vessel level indication 9109160023 910903 PDR ADDCK 05000280 G
- I I
system during natural circulation cooldown with steam voids presen The omission of this procedure is addressed in paragrap Other specific technical and human factors comments are contained in Appendix One inspector followupwas identified for the remaining items from Inspection Report 280,281/90-09 EOP inspection (paragraph 2).
- ** *
REPORT DETAILS Persons Contacted
- R. Bilaeu, Licensing Engineer
- R. Blount, Supervisor-Station Procedures
- D. Christian, Assistant Station Manager M. Gabriel, Senior Reactor Operator R. Mabe, Procedure Writer
- H. Mccallum, Operations Training Supervisor R. Morgan, staff Quality Specialist/Audit Coordinator
- R. Saunders, Assistant Vice President - Nuclear Operations
- E. Smith, Manager Quality Assurance A. Swander, EOP V/V Coordinator-Nuclear Procedures Specialist Other licensee employees contacted included engineers, mechanics, technicians, operators, personne NRC Representatives M. Branch, Senior Resident Inspector s. Tingen, Resident Inspector
- J. York, Resident Inspector
- Attended Exit Interview instructors, and office A listing of abbreviations used in this report is contained in Appendix Operation*s Without Reactor Vessel Level Instrumentation System The team compared the Surry EOPs with the NRC approved ERGs and found that the licensee still had not written an EOP corresponding to ES-0. 4, Natural Circulation Cooldown With Steam Void in the Vessel Without Reactor Vessel Level Instrumentation Syste The licensee responded to the original comment that the Plant Specific Technical Guidelines Deviation was inadequate for not including operations without Reactor Vessel Level Instrumentation System in "NRC EOP Revision 1A Audit, Surry Power Station Responses to Appendix Items B, c, D, E, ".
The response stated "No procedure is require ES-0. 4 is designed for plant conf.iguration without Reactor Vessel Level Instrumentation Syste The Step Deviation Evaluation adequately stated the justification for not having a procedure for natural circulation without Reactor Vessel Level Instrumentation Syste Without at least one natural circulation and one forced circulation range of Reactor Vessel Level Instrumentation System operable, Technical Specifications require the plant to be placed in hot shutdown within 60 hour6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> This limiting conditions for
- operation ensures that the likelihood of an event occurring without Reactor Vessel Level Instrumentation System available is smal Therefore, guidance for natural circulation cooldown without Reactor Vessel Level Instrumentation System available is not considered necessar Consequently, the corresponding procedure is not included in the EOP set."
The team concluded that the relatively limited probability of an event does not relieve the licensee from the requirements to provide operating procedures for ERG The team discussed this with licensee management, and the licensee agreed to provide a procedure that corresponds to ES-The licensee further stated that this would be completed by the end of the second training cycle after this inspection report is issue This is considered part of IFI 50-280, 281/91-23-01: Incomplete items from Inspection Report 50-280,281/90-09 EOP inspectio.
Independent Technical Adequacy Review of EOPs During the initial EOP inspection the following human factors/Writer's Guide concerns were noted:
-
Inaccurate and inconsistent use of key words and symbols to include the words verify and should, transition terms, bullets and asterisk Cautions and notes used in conflict with Writer's Guide instruction During the followup inspection the team found that these concerns had been resolved through either modification of the Writer's Guide for Dual-Column Procedures (VPAP-0505)
or through changes to the applicable procedure step A review of revision 2 of VPAP-0505 revealed that the Writer's Guide was generally well written and complete, and would provide the necessary guidance to ensure consistently written dual-column procedures over time and personnel change The team reviewed the licensee's responses to Writer's Guide comments and observations. With the exception of the items listed below, the comments are either no longer applicable or have been satisfactorily resolved with revisions. to the Writer's Guid Appendix C item I. 4.: The original comment read, "section 6. 4. 8 ( c), Referencing, stated that specific steps of supplemental procedures should be considered for incorporation into the original procedure instead of referencin The Writer's Guide failed to provide restrictive criteria for consideration."
The licensee responded that the Writer's Guide could not be
.,
4 * restrictive and state how many steps equals lengthy, and that the decision to reference, transition, or to include steps must be made on.a case by case basi The team agreed that the Writer's Guide need not state how many steps equals lengthy; however, it was inappropriate not to provide Writer's Guide guidance to the EOP writer to aid in making the decision on when to reference or when to include step This guidance will ensure consistency of EOPs over time and between different EOP writers. The licensee concurred and will include instructions on referencing and branching in a future revision to the Writer's Guid Appendix C item II.6: This comment identified the use of punctuation/underlining in a manner that was not clearly defined in the Writer's Guid The Writer's Guide stated
"underline AND or OR in high-level steps for emphasis."
In ECA-1.2, substeps contained an underlined AN The licensee agreed to revise the Writer's Guide to state that underlining was an optional technique for emphasis on high or low level steps when these terms are used as conjunction Management Control of EOPs and Interfacing Documents The licensees Quality Assurance group performed an EOP audit in April 198 The audit was reasonably detailed and found a few of the technical problems later identified in the NRC EOP inspectio The team reviewed the completed audit report and determined the documentation of the corrective actions was adequat During the April 1990 exit for NRC Inspection Report 90-09 the licensee stated that EOP procedures would be audited on an annual basi As of August 1991, Quality Assurance had not reaudited the EOP procedure.
Follow-up on Previous Inspection Findings (91702) (Closed)
IFI 280,281/90-09-0l, Revise AP-12.01, Loss of Intake Cana Revisions to the procedure have been complete The inspectors reviewed design calculation ME 166 which included resolution of all previous comments on the upper intake canal water inventory loss during loss of all AC powe The procedure reflected the results of the engineering calculation and directed operators to take appropriate actions to maintain cooling water flow for safety related equipmen (Closed)
IFI 280,281/90-09-02, Initiate operator use of sound powered phone Installation of the sound powered phones, enclosures, and extension cords were complete *
A memorandum had been distributed to all operators to make them aware of the equipment location.
4 (Closed)
IFI 280,281/90-09-03, Provide emergency battery power to operations radio repeater The revisions to the procedure to provide power to operations radio repeaters had been complete (Closed) IFI 280, 281/90-09-04, Conduct IPE analysis of loss of intake canal resulting from loss of all AC power to a uni An individual plant examination was being conducted to address the canal inventory questio The IPE was scheduled to be submitted for NRC review on August 30, 199 (Closed)
IFI 280,281/90-09-05, Revise 18 steps in 15 procedures to address technical deficiencie The specific technical deficiencies listed in this IFI were adequately dispositione (Closed)
IFI 280, 281/90-09-06, Revise one step in AP-22. 01, Loss of refueling cavity level, prior to refueling outag The inspectors reviewed this Abnormal Operating Procedure and determined that it had been appropriately revise (Closed)
IFI 280,281/90-09-07, Corrective actions for all technical and human factors comment The inspectors reviewed all of the comments from the previous inspection and determined that with few exceptions the items had been satisfactorily dispositione The remaining deficiencies are listed in Appendix This is considered part of IFI
- 50-280,281/91-23-0l: Incomplete items from Inspection Report 50-280,281/90-09 EOP inspectio (Closed)
IFI 280,281/90-09-08, Corrective actions for all Writer's Guide comments. The previous concerns had been resolved through either modification of the Writer's Guide for Dual-Column Procedures (VPAP-0505)
or through changes to the applicable procedure step (Closed)
IFI 280,281/90-09-09, Corrective actions for all nomenclature and labelling i tern An ongoing program for correcting nomenclature and labelling i terns has been established and implemente The inspectors reviewed samples of revised labelling and did not identify any discrepancie Exit Interview The inspection scope and findings were summarized on August 2, 1991, with those persons indicated in paragraph The NRC described the areas inspected and discussed in detail the inspection finding No proprietary material is
- contained in this repor received from the license No dissenting comments were 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 IFI 280,281/90-09-07 IFI 280,281/90-09-08 IFI 280,281/90-09-09 IFI 280,281/91-23-01 Description (Closed) Revise AP-12.01, Loss of Intake Cana (Closed) Initiate operator use of sound powered phone (Closed) Provide emergency battery power to operations radio repeater (Closed) Conduct IPE analysis of loss of intake canal resulting from loss of all AC power to a uni (Closed) Revise 18 steps in 15 procedures to address technical deficiencie (Closed)
Revise one step in AP-22.01, Loss of refueling cavity level, prior to refueling outag (Closed)
Corrective actions for all technical and human factors comment (Closed)
Corrective actions for all Writer's Guide comment (Closed)
Corrective actions for all nomenclature and labelling item (Open)
Incomplete items from
Inspection Report
280,281/90-09
EOP inspection (paragraph 2).
- E-0
E-1
E-2
E-3
ES-ES-ES-ES-ES-ES-ES-ES-1. 4
ES-ES-ES-ES-ECA-ECA-ECA-ECA-ECA-ECA-ECA-ECA-ECA-F-0
F-1
F-2
F-3
F-4
F-5
F-6
FR-FR-FR-FR-FR-FR-FR-APPENDIX A
EOPs REVIEWED
Reactor Trip 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
Void in Reactor Vessel
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 Using Blowdown
Post-SGTR Cooldown Using Steam Dump
Loss of all AC Power
Loss of all AC Power Recovery without SI
Required
Loss Of All AC Power With SI Required
Loss Of Emergency Coolant Recirculation
LOCA Outside Containment
Uncontrolled Depressurization of All Steam
Generators
SGTR with Loss of Reactor Coolant -
Subcooled Recovery Desired
SGTR with Loss of Reactor Coolant
Saturated Recovery
SGTR Without Pressurizer Pressure Control
Critical Safety Function 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 of Secondary Heat Sink
Response to Steam Generator Overpressure
Response to Steam Generator High Level
Response to Loss of Normal Steam Release
REV 6
REV 5
REV 4
REV 5
REV 2
REV 5
REV 4
REV 3
REV 5
REV 5
REV 4
REV 2
REV 2
REV 4
REV 4
REV 4
REV 5
REV 5
REV 3
REV 4
REV 3
REV 5
REV 5
REV 5
REV 4
REV 2
REV 2
REV 2
REV 2
REV 2
REV 2
REV 2
REV 4
REV 4
REV 4
REV 4
REV 2
REV 2
'
Appendix A
Capabilities
REV 2
FR-Response to Steam Generator Low Level
REV 2
FR-Response To Imminent Pressurized Thermal
REV 4
Shock Condition
FR-Response to Anticipated Pressurized Thermal
REV 2
Shock Condition
FR-Response to Nuclear Power Generation/ATWS
REV 4
FR-Response to Loss of Core Shutdown
REV 2
FR-Response to Containment High Pressure
REV 4
FR-Response to Containment Flooding
REV 3
FR-Response To Containment High Radiation Level REV 2
FR-Response to Containment Positive Pressure
REV 4
APPENDIX B
TECHNICAL AND HUMAN FACTORS COMMENTS
The comments/responses for EOPs listed in appendix A that are not
listed below the have been reviewed and have been satisfactorily
resolved: ES-1.2 Post LOCA Cooldown and Depressurization step 16e: This step required the operator to read 25
gallons per minute or greater on the charging flow
indicato The scale at the bottom of the indicator is
compressed; thus making it difficult to determine this
flow rat The licensee stated in its response that
replacement of the meter scale has been deferred under
the Five Year Capital Improvement Plan (Virginia Power
Letter 90-312A).
ECA-Loss of All AC Power Recovery With SI Required step 3 caution: This step stated the energized emergency
bus load should not exceed 1200 amp The Main Control
Room emergency bus current indicator reads o - 800 amp The licensee revised the caution statemen The change
to the caution statement did not eliminate the 1200 amps
limitation, and the o -
800 current meter was still in
us.
ECA-1.1 Loss of Emergency Coolant Recirculation step 14 RNO: This step failed to state the method to be
used to throttle the Safety Injection valves. The Safety
Injection valves were not designed to be throttled. The
licensee failed to provide adequate instructions to
accomplish throttling of Safety Injection flo.
ECA-2.1 Uncontrolled Depressurization of All Steam Generators step 3 note: The recorders for Reactor Coolant Pump
leakoff flow 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 The licensee stated labelling
concerns will be resolved by the end of first quarter
1993.
- Appendix B
2 Attachment 1, Step 4b: This comment identified that
steam generator blowdown permission key switches had no
position indication labelin The licensee stated
labelling concerns will be resolved by the end of first
quarter 199 Step 14a RNO: This step did not direct the operator to
restore intake canal level by going to AP 12. 0 AP 12. 01 ensures that the licensee stay within the
design envelope for loss of upper intake cana The
instructions provided in this step did not contain the
requirements for the design envelope for loss of intake
cana.
ECA-3. 2 SGTR with Loss of Reactor Coolant-saturated Recovery Step 21d RNO:
This step stated isolate I The
licensee had two separate IA systems, outside IA and
containment I This * step require isolation of
containment IA, but did not identify which IA system
was to be isolate.
ECA-3.3 SGTR without Pressurizer Pressure Control Step 3 o note: This note was only applicable when
depressurization occurred as a result of an open PORV
or use of the spray syste The licensee stated the
ERG basis document was in error and the PORVs were
assumed to be unavailable during this procedur Vendor emergency procedure development groups have not
provided supplementary documented guidance concerning
this mechanism and the subsequent deletion of this
not This note will be revised when appropriate
guidance is availabl The comments/responses for Abnormal Operating Procedures not
listed below have been satisfactorily resolve.
AP-10.00 Station Blackout Step 7c: This step required that the instrument air
compressor be powered from the #3 Emergency * Diesel
Generator during emergency conditions.. The bullet did
not specify the air compressors could be powered from
alternate power source step 7f: This step required the turbine building
instrument air system be crossed tied by opening I-IA-
44 and 2-IA-4 These valves were not labeled in the
fiel Appendix B
3 Step 23: This step required the 34. 5 KV bus #5 be
energized by offsite powe The step did not provide
instructions for accomplishing this tas.
AP-16.00 Excessive RCS Leakage General Comment: This procedure directed the operators
to perform a safety injection prior to a reactor tri These instructions were inappropriate in that Safety
Injection isolates feedwater and could result in an
ATWS if the reactor failed to tri The licensee
stated the procedure would be revised to require a
reactor trip prior to a safety injectio.
AP-22.00 Fuel Handling Abnormal Conditions Step 5b:
This procedure step required valves VS-103A
and VS-103B be placed in the closed positio These
valves do not have position indicatio The licensee
responded by revising the procedure to require the
closure of alternate valve However, step 7d RNO
still requires the closure of VS-103A and *
IFI
KV
NRC
RNO
APPENDIX C
Abbreviations
Emergency Operating Procedures
Inspector Followup Item
Kilovolt
Nuclear Regulatory Commission
Power Operated Relief Valve
Response Not Obtained