IR 05000344/1988023
| ML20151A389 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 06/30/1988 |
| From: | Hopkins J, Lennartz J, Love R, Suh G, Wright G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20151A358 | List: |
| References | |
| 50-344-88-23, NUDOCS 8807190356 | |
| Download: ML20151A389 (13) | |
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U. S. NUCLEAR REGULATORY.COMISSION
REGION V
Report No. 50-344/88-23 Docket No. 50-344 License No. NPF-1 Licensee:
Portland General Electric Company 121 S.W. Salmon Street Portland, OR 97204 Facility Name:
Trojan Nuclear Plant Inspected At: Trojan Nuclear Plant Site, Rainier, Oregon Inspection Conducted:
May 10-26, 1988.
Inspectors:
bC~bN F aG--
6-30-PS R. S. Love, Reactor Inspector, Region III Date Signed Team Leader St* % 2 [ cott (,-lo-88 G. Suh, Resident Inspector, Region V Date Signed M
FoC-6 ~10-98 J. Hopkins, License Examiner, Region III Date Signed SM Fe(L b - 30- 28 J. Lennartz, License Examiner, Region III Date Signed Consultants:
J. Sears, Consultant (COMEX)
8. Glickstein, Consultant (SAIC)
Approved By:
btk F00--
l,-30-86 Geoffrey C. Wright, Chief Date Signed Operations Branch, Division of Reactor Safety, Region III 8907190356 880630
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PDR ADOCK 05000344 O
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Inspection Summa g Inspection conducted on May 10-26, 1988 (Report No. 50-344/88-23)
Areas Inspected:
Special announced safety inspection to verify that the Trojan emergency operating procedures (EOPs) are technically correct; that their'
specifications.can be meaningfully accomplished using existing equipment,-
controls, and instrumentation; and that the available procedures are sufficiently usable to provide operators an effective accident recovery tool.
The inspection was conducted in accordance with Temporary Instruction (TI) 2515/92.
Results:
Of the areas inspected, no violations or deviations were identified.-
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DETAILS'
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1.
Persons Contacted Portland General Electric Company (PGE)
- D. W. Cockfield, Vice President Nuclear
- C. A. Olmstead, General Manager
- R. P. Schmitt, Manager, Operations and Maintenance
- R. H. Budzeck, Assistant Operations Supervisor
- R.-L. Russell, Operations Supervisor
- M. A. Hoy, Operations Engineer
- D. L. Nordstrom,-Compliance Engineer
- J. Taylor, Shif t Supervisor
- C.sH. Brown, QA Operations Branch Manager
- D. J. Modeen, Senior Nuclear Engineer B. Hunt, Auxiliary Operator (RO)
W. Nicholson', Assistant Shift Supervisor (SRO)
K. Oberloh, Assistant Control Operator (RO)
M. Richard, Assistant Control Operator (RO)
P. Leahy, Control Operator (SRO)
R. Monsive, Control Operator (RO)
J. Conklin, Training Specialist III W. Reeve, Assistant Shift Supervisor (SRO)
T. Andone, Shift Supervisor (SRO)
K. Heese, Assistant Shift Supervisor (SRO)
T. Losinski, Auxiliary Operator D. Foster, Auxiliary Operator J. McGuire, Control Operator (RO)
K. Hanson, Shift Supervisor (SRO)
U. S. Nuclear Regulatory Commission (NRC)
- W. H. Regan, Jr., Chief, Human Factors Assessment Branch, NRR
- A. Chaffee, Deputy Division Director, DRS&P, Region V
- T. Burdick, Cnief, Operator Licensing Section, Region III 2.
Emergency Operating Procedures (25592)
a.
Background Emergency Operating Procedures (EOPs) have undergone significant changes due to the 1979 accident at the Three Mile Island (TMI)
facility.
The post-THI E0Ps are function-otlented rather than event-oriented.
Function-oriented E0Ps provide the oparator guidance on how to verify the adequacy of critical safety functions and how to restore and maintain these functions when they are degraded.
Function-oriented E0Ps are written in a way that the operator need not diagnose a specific accident event to maintain the plant in a safe condition.
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The purpose of this inspection was to verify that the Trojan E0Ps are technically correct, that their specified actions can be accomplished using existing equipment, controls and instrumentation, and'that the
.available procedures are sufficiently usable.
The objectives were accomplished by performing:. (A) a desk-top-review of all of Trojan's Emergency Operating Procedures (E0Ps) (25 Emergency. Instructions (EIs), six critical safety function status trees, and 18 function restoration procedures (FRs); (8) system walkdowns;of 16 E0Ps and one Off-Normal Instruction; and (C) a human factors review of the procedures, walkdown of procedures, and interview of ten users or develop 3rs of the procedures and Procedures Generating Package (PGP).
For a detailed listing of the procedures reviewed and walked down, see Appendix A.
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This inspection report only provides examples of observations noted during the inspection.
The licensee was provided detailed debriefings in which all of the team's observations were discussed.
b.
Desk-Top Review The desk-top review was accomplished by comparing the procedures identified in Appendix A with the Westinghouse 0wners Group (WOG).
Emergency Response Guidelines (ERGS), Trojan's Prccedures Generating Package (PGP) (writers guide and step verification document), plant specific instrument setpoint document, and the applicable background documents. When deviations between the ERGS, PGP, E0Ps, and the setpoint document.were identified, the inspectors verified that the deviations were documented, justified, and when required, that a safety an' ysis had been performed in accordance with 10 CFR 50.59.
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In addition, the inspectors reviewed the licensee's verification and validation (V&V) of the Trojan E0?s.
(1)
Inspection Results (a) Generic technical guidelines wer, prepared for all of the ERGS.
These generic guidelines provide a ca aplete and documented analytical basis for each of the emergency procedures.
The generic technical guidelines have been verified by the WOG.
The Trojan step verification documents and E0Ps are modeled after and closely paralleled the generic WOG High Pressure ERGS, Revision 1.
In general, Trojan E0P deviations from the WOG Guidelines, Trojan PGP, setpoint documents, and step sequencing were identified, documented and technically justified.
While many minor deviations between the E0Ps and the ERGS were identified, -5ere were no safety significant deviations identified that required a 10 CFR 50.59 safety analysis
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review.
l During a review of the "Step Verification" and "Values and Setpoints" documents, it was observed that changes to these
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documents were uncontrolled in that they were made by pen and ink.
Further investigation indicated that the
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documents themselves were not listed as "Quality Documents" and were therefore, not controlled.
The information from these two documents are directly _ translated into the Trojan E0Ps, which are quality related and controlled documents.
During:the exit interview on May 26, 1988, the licensee committed ~to making.the Step Verification and Values and Setpoints documents controlled documents and to control all changes to these documents.
Pending verification that the e
subject documents are being controlled, this item is open (50-344/88-23-01).
(b) In general, the Trojan E0Ps were found acceptable.
However, when plant specific information from the step verifir ' ion and setpoint documents were entered in the
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E0Ps, numerous typographical errors and omissions were made and not corrected.
This is indicative of a less than adequate review of the procedures by line management and quality overview groups (e.g., operating crews, training department personnel and the QA/QC organization).
The following examples are provided:
(1)
In Procedure EI-3, the third CAUTION prior to Stap 1 directs the operator te perform the actions in Procedure FR-H.3 in parallel with EI-3 if steam generator (S/G) level exceeds 95%.
The same CAUTION is in FR-H.3 but is worded differently.
The intent of the CAUTION, as worded in FR-H.3, is to remind the operator not to release steam from an overfilled S/G until the overfill condition is evaluated.
During the inspection, the licensee agreed to reword the CAUTION in EI-3.
(2) The Values and Setpoints' Document was revised to require a pressurizer level of greater than 21%
' greater than 30% for adverse containment].
This revision was made in E0P ECA-3.2, Step 10.c, Action /
Ernected Response (A/ER) column.
However, pressurizer level was not revised in E0P ECA-3.2, Step 9.b, A/ER and Response Not Obtained (RNO) columns; StepEll.d, A/ER column; Stip 12.d, A/ER column; and Step 13.c, A/ER column.
In addition, the revised pressurizer levels were not changed in the Step Verification Document.
In that operator training and system interlocks would prevent energizing the pressurizer heaters if the heaters were not adequately covered with water, the incorrect pressurizer level values were not considered to be safety significant.
(3) Another example of setpoint changes not being reflected in the E0Ps was the reactor coolant pump (RCP) seal injection flow.
The revised flow rate was 6.6 to 13 gpm.
In Procedure ES-0.3, Step 8, A/ER column, the RCP seal injection flow rate was listed as 6.6 to 9 gpm.
The RCP seal injection flow rates
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were missing throughout the rest of the E0Ps
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(Examples:
FR-C.1, Step 5;zFR-C.2, Step 5;.FR-C.3,
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Step 24; FR-P.1, Step 10; ES-0.2, Step 19; ES-0.4, Step 7; and ES-1.1, Steps 5 and 19).
In'that all:
operators queried were aware of the correct RCP seal injection flow-rate, the missing and incorrect. flow rates were not considered to'be safety significant.
The licensee was informed that all like CAUTIONS, NOTES, and steps should be worded consistently throughout^all of the E0Ps. 'The licensee was also informed that positive steps must be taken to ensure that all changes (setpoint values, steps, etc.) be reflected.in all the applicable E0Ps and steps within the procedure.
At the present time, one individual is responsible for researching the potential procedure revision and doing a manual search of all the E0Ps to ensure that the change is reflected in-all the E0Ps and steps within the procedure.
During the exit interview on May 26, 1988, the licensee committed to establishing a system to ensure that all changes are reflected in all the applicable procedures. At the time of.the exit interview, the steps being considered were:
(1) assigning additional personnel to the E0P maintenance effort; (2) modifying an existing computer system; or (3) purchasing a new computer system capable of identifying all procedures / steps where a given setpoint, CAUTION, NOTE, or step is utilized.
Pendir.g verification that pos" 've action has been taken
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and steps are in place to er that all changes are reflected in all the applicau E0Ps and steps within a given E0P, this item is ietn (50-344/88-23-02).
c.
Plant Walkdown
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Plant walkdowns of select Emergency Operating Procedures (EOPs) and one Off-Normal Instruction (ONI) were performed during the inspection.
See Appendix A for a list of procedures walked down.
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.The ONI walked down was in draft form, and was undergoing the Trojan I
Procedure Validation and Verification process.
Three walkdown teams l
were utilized, with each team consisting of two NRC personnel and a
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licensed Reactor Operator (RO) or a licensed Senior Reactor Operator (SRO).
The walkdowns were performed to verify that the E0P specified actions could be accomplished by the operators using existing equipment, controls, and instrumentation.
(1)
Inspection Results In general, the operators were able to implement the Trojan E0Ps as written.
However, some concerns were identified during the walkdown of the E0Ps.
(a) Availability of Aids (1)
In many of the E0Ps (e.g., ES-0.1) the operator is directed to re-establish letdown flow.
The E0P
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directs the operator to. adjust the letdown pressure controller tc 61 psig before opening the letdown orifice isolation valves.
However,t.ith the orifice, isolation valves closed, the operator would not be'
able to read 60 psig on-any available instrumentation.
This step should be reworded in-all applicaoie E0Ps to provide the operator with useful directions.
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(2) While executing the Loss of All AC procedure (ECA 0.0), the procedure directs the operator to fill
the Diesel Fire Pump Fuel Tank using barrels filled from the Start-up Boiler Storage Tank.
This method would be very time consuming and difficult to perform.
The Inspection Team recommended that a better method be developed.
(3) Small hand tools are available to the Auxiliary Operator at his workstation or in the plant tool crib.
Keys for the tool crib as well as hand held radios are available in the Control' Room.
Special tools and equipment to support the EOPs were available.in Emergency Lockers throughout the plant.
In addition to the available tools,.the Inspection Team recommended placing a dedicated screwdriver in the Emergency Locker by the Plant Air Compressors (A/C).
A screwdriver.is needed to hook up emergency cooling to the "B" Joy A/C as directed by many of the E0Ps (e.g., ECA 0.0), if needed.
(4) Many of the E0Ps (e.g., EI-0) direct the operator to verify that all containment Phase A isolation valves have closed if a PF:se A automatic isolation signal is iresent.
If the-isolation valves fail to close automatically, and cannot be closed manually from the Main Control Board, then the Auxiliary Operator would have to close the valves locally.
A list of valves that the Auxiliary Operator would have to close would.
have to be generated from the status lights in the f
Control Room.
The Inspection Team recommended that a Phase A isolation valve checklist be developed and made available in the Control Room or incorporated into the applicable procedures.
This would provide the operators with a readily available list if local isolation of these valves was necessary.
(b)
Labels and Tags The labeling / tagging of equipment in support of the Els was very good.
(c) Equipment Accessibility
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(1) Automatic and manual controls for safety systems are l
installed in the Control Room.
In the event of a
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malfunction of one or more of these control functions, 7the ' Auxiliary Operator would be directed by the Control Room to locally operate various equipment, e.g., open/close valves; start /stop pumps, motors, diesel generators,'etc.
In general, equipment accessibility was good.
Many valves had permanent ladders and/or platforms installed.
However, in one-case a ladder is required to reach the nitrogen supply manual isolation valves (8965 A/B) to the Safety Injection (SI) Accumulators.
Since thesel valves have low priority ^when addressed in the E0Ps, inaccessibility of-these is not considered safety significant.
The Inspection Team recommended that this issue be addressed by the licensee in the near future.
(2) During the inspection, it was noted that Figures 12.3-20 through 22 of the Trojan Final Safety Analysis Report indicate that the post-accident radiation surrounding certain valves may oe greater than 100 Rem /hr.
Examples of these include the RHR Pump Discharge to Safety Injection or Charging Pump Suction valves (8804 A/B), Boron Injection Tank Inlet (BIT) valves (8803 A/B), and BIT Outlet valves (8801 A/B).
This would create a safety hazard if these valves had to be locally operated in this environment.
To limit personnel exposure to radiation in the Auxiliary Building during an accident, the licensee has inserted a CAUTION in applicable E0Ps.
This CAUTION requires reactor coolant system (RCS) gross gamma activity to oe verified-below 700 uCi/ gram before letdown can be established.
If RCS gross gamma activity was greater than 700 uCi/ gram and letdown was established, this.would simit p?rsonnel entry into various areas of the Auxiliary BL11a1ng.
3.
Human Factors Review The objectives of the human factors review was to c.sure that the E0Ps followed the guidance in the Trojan Nuclear Plant Procedures Generation Package (PGP) for Emergency Operating Proceduras (E0Ps), A0-4-7, Revision 3, and NUREG 0899, Guidelines for the Preparation of Emergency Operating Procedures; and to ensure that the E0Ps can be physicully and effectively carried out.
To achieve these objectives, the human factors evaluator performed a desk-top review of the E0Ps, participated in walkdowns of select E0Ps, and interviewed selected users and developers of the E0Ps.
a.
Desk-Top Review Yhe E0Ps reviewed are listed in Appendix A, and, in general, comply with the PGP and NUREG 0899.
The following exceptions were noted:
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t (1) Steps are wordy or too much detail is provided:
This concern included (1) unnecessary Cautions and Notes (Examples
.first Caution prior to Step 1 in ES-0.2, and the Note at the end of
'ECA-1,1, Step 24);.(2) parenthetical comments contained in substeps which should be Notes or additional steps (Examples -
'ES-1.4, Step 1.c); and-(3) steps which are not worded in a simple and clear structure.(Examples --EI-0, Step 6 a RN0; FR-c.2, Step 10.a, RN0; ES-0.1, Step 5.a. RNO).
(2) Regarding Cautions and Notes, the following two exceptions were identified: (1) Cautions and Notes which contain action statements need to be re-written to remove the action statement.
(Examples - Caution prior to Step 4 in ES-1.3, Caution prior to Step 45 in ES-1.1); (2) Cautions and Notes are placed in the middle of Steps and they should precede the affected step.
(Examples - ES-3.2, Step 9; ECA-0.0, Step 16, ECA-2,1, Step 1; FR-H.1, Step 2c).
(3) Ambiguous phrases such as "as necessary" should be made more specific.
For example, in ES 1.1, Step Sd, the operator must adjust seal injection flow "as necessary" using HFK-182.
This step could be n;ade clearer by identifying the range (6.6 gpm to 13 gpm) within which seal injection flow must be adjusted.
(4) Steps which are continued on additional pages are not identified as being continued.
The lack of identification makes it more difficult for the operator to keep his place withi1 the procedure.
(Examples - ES-0.2, Step 8; EI-3, Step 4; ECA-3.3, Step 4; FR-P.1, Step 8).
(5) Logic statements are not worded so that the conli';on is-presented first and then the action to be taken.
' Examples -
ECA-0.0, Step 16b.3 RN0; FR-I.3, Step 170.2 RNO).
(6) The same step is worded differently in different pricedures.
(Examples - FR-C.1, Step 12c versus FR-C.2, Step 13a; FR-H.2, Step 2a RNO versus FR-H.3, Step 2c RN0).
Consistent wording should be maintained when steps are the same.
b.
Walkdown of E0Ps The inspectiol team observed that operators were familiar with the plant and labeling of plant equipment was very good.
However, the following concerns were identified durinp the walkdowns:
(1) The following meters for reading containment pressure do not have units indicated on them:
PI-2080, PI-2001, PI-2082, PI-2083.
During the exit interview, the licensee stated the
"psig" will be identified on the meters prior to reactor startup.
The concern was identified in EI-0, Step 11a.
(2) Local in plant actions are incorrectly identified as ma.iual control-room actions.
During the walkdown of FR-S.1, Step 7.b, RNO, the operator was instructed to "manually isolate [ primary
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makeup water] dilution flow paths".
Because the step was worded o
as a manual action, the operator was unclear as to what the step was telling him to do.
The operator later determined that this
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was a local action and could not be accomplished from the control room.
(Examples - E-0, Step 5e; ECA-0.0, Step 15; FR-Z.1, Step 6 RN0; FR-C.2, Step 1),
(3) Plant-specific procedures need to be referenced in the E0Ps.
Procedures such as starting containment chillers and adding makeup to the reactor coolant' system from the holdup. tanks or spent fuel pool, are abnormal operations which may not be remembered by the operator.
(Examples - ECA-3.1, Step 34; ECA-1.1, Steps 16b ar.d 16c; ECA-0.0, Step 17).
(4) During the walkdown of FR-H.1,.the need for an instrument and control (I&C) procedure was identified.
In Step 5.b, RNO, of the-procedure, the operator is instructed to have I&C open the
.feedwater isolation valves by lifting leads or jumpering contacts.
There was no I&C procedure to do.this action.
The operator pointed out.that this action may take a long time to perform because the I&C technician may not be familiar with the
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(5) Valves which are needed to restore steam generator blowdown and sampling are not identified in the procedure.
These valves are:
M0-6716, M0-6717, MO-6718 and MO-6719.
This concern was identified in EI-3, Step 44.
(6) Several E0Ps require extensive action by the Auxiliary Operator.
A spare copy of the E0P or plant-specific procedure appears to be needed in the auxiliary operator field office.
The Auxiliary Operator currently has to go to the control room and make a copy of the procedure each time it is needed.
This issue is of greater concern during a loss of all AC power when the copy machine will not be available.
(Example - ECA-0.0, Step 7, RNO).
c.
Interviews A total of ten personnel interviews were conducted.
Persons interviewed included:
two Shift Supervisors (SSs); two Ae.sistant Shift Supervisors (ASSs); two Control Operators (COs); two Assistant Control Operators (ACOs); and two Auxiliary Operators (A0s).
In general, the operators expressed a positive attitude towards the procedures.
A concern with place keeping methods was identified.
During the tabletop review it was noted that numerous transitions are made in the procedures.
For example, in EI-1, there are approximately seven transition points out of the procedure.
The operators interviewed stated that they currently use yellow tabs, sbaets of paper, or paper clips to keep their place, and that these methods are not always effectiv...
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Discussion with the licensee and walkdowns of the procedures identified the availability of log sheets to document when transitions are made, and grease pencils to check off completed steps in the procedure.
The licensee was already aware of this concern and has identified it as a training issue.
Overall, operators felt that the current procedures are'a big improvement over the previous revision and supported their use.
4.
Open Items Open items are matters which have been discussed with the licensee which will be reviewed fu'rther by the inspector, and which involves some action on the part of the NRC or licensee or both.
The open items disclosed during the inspection are-discussed in Paragraphs 2.b(1)(a) and 2.b(1)(b)
of this report.
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Exit Interview (30703)
The inspectors met with the licensee representatives (denoted in Paragraph 1) at the conclusion of the inspection on May 26, 1988.
The-inspectors summarized the purpose, scope, and-findings of the inspection and the likaly informational content of the report.
The-licensee acknowledged this information and did not identify any proprietary information.
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APPENDIX A-Listing of Emergency Instructions
- EI-0, Revision 13 Reactor Trip, Safety Injection, and Diagnosis ES-0.0, Revision 1 Rediagnosis ES-0.1, Revision 5 Reactor Trip. Response.
-ES-0.2, Revision 3 Nat. ural Circulation.Cooldown
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- ES-0.3, R3 vision 1 Natural Circulation Cooldown With Steam Void in Vessel (W/RVLIS)
ES-0.4, Revision 1 Natural Circulation Cooldown With Steam Void in Vessel (W/0-RVLIS)
EI-1,. Revision 19 Loss of Reactor or Secondary Coolant
- ES-1.3, Revision 6 Transfer;to Cold Leg Recirculation ES-1.4, Revision 5 Transfer to Hot-Leg Recirculation
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EI-2, Revision ~13 Faulted. Steam Generator Isolation
- EI-3, Revision 12
_ Steam Generator Tube Rupture ES-3.1,~ Revision 1 Post = Steam. Generator Tube Rupture Cooldown Using Backfill ES-3.2, Revision 2 Post Steam Generator Tube Rupture Cooldown
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Using Steam Generator Blowdown ES-3.3, Revision 2 Post Steam Generator Tube Rupture Cooldown Using Steam Dump
- ECA-0.0,' Revision 3 Loss of All AC Power ECA-0.1, Revision 2 Recovery From Loss of All AC Power Without SI Required
- ECA-1.1, Revision 2 Loss of Emergency Coolant Recirculation ECA-1.2, Revision 1 LOCA a,tside Containment ECA-2.1, Revision'2_
Unco-olled Depressurization of All Steam Geneiators
- ECA-3.1, Revision 2 SGTR With Loss of Reactor Coolant - Subcooled Recovery Desired ECA-3.2, Revision 2 SGTR With Loss of Reactor Coolant - Saturated Recovery Desired
- ECA-3.3, Revision 2 SGTR Without Pressurizer Pressure Control FR-0, Revision 4 Critical Safety Function Status Trees (CSFST)
- FR-S.1, Revision 4 Response to Nuclear Power Generation (ATWS)
FR-S.2, Revision 2 Response to Loss of Core Shutdown l
- FR-C.1, Revision 6 Response to Inadequate Core Cooling l
FR-C.2, Revision 2 Response to Degraded Core Cooling
~FR-C.3, Revision 3
?esponse to-Saturated Core Cooling
- FR-H.1, Revision 6 Recoonse to Loss of Secondary Heat Sink FR-H.2, Revision 3
. Response to Steam Generator Overpressure FR-H.3, Revision 3 Response to Steam Generator High Level FR-H.4, Revision 3 Response to Loss of Normal Steam Release Capability FR-H.5, Revision 2 Response to Steam Generator Low Level
- FR-P.1, Revision 4 Response to Imminent Pressurized Thermal Shock Condition
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FR-P.1, Revision 2
. Response to Anticipated Pressurized Thermal Shock Condition
- FR-Z.1, Revision 2 Response to High-Containment Pressure FR-Z.2, Revision 2 Response to High' Containment Sump Level
~FR-Z.3, Revision 3-Response to High Containment Radiation Levels
- FR-I.1, Revision 3 Response to High Pressurizer Level FR-I.1, Revision 3 Response to Low Pressurizer Level FR-I.3, Revision 6 Response to Voids in Reactor Vessel
- 0NI-17, ORAFT Control Room Inaccessibility NOTES:
1.
A desk-top review was performed on all Emergency Instructions and the-Off-Normal Instruction.
2.
- Indicates the instructions that were walked-down.
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