IR 05000445/1989059
| ML20248D244 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 09/25/1989 |
| From: | Cummins J, Gagliardo J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20248D237 | List: |
| References | |
| 50-445-89-59, 50-446-89-59, IEB-85-003, IEB-85-3, IEIN-88-075, IEIN-88-75, NUDOCS 8910040171 | |
| Download: ML20248D244 (41) | |
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, APPENDIX
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~U.SP NUCLEAR REGULATORY COMMISSION h'. C REGION-IV p , -
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NispectionReport: 50-445/89-5 , Permits:- CPPR-126
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,. E' 50-446/89-5 CPPR-12 ' Dockets: L50-4451 50-446
? Applicant: TU Electric- ~
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400 North Olive Street, Lock Box 81
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'InspectLion' At:: ComanchePeaiSteamElectricStation
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a ~h-InspectionConducted: August 14-25,-1989
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h 26 I * 3] Inspector: h. .
# , ;unnins, Team Leader . Operational Date + ,
y Programs Section Division of Reactor Safety
. Team Members: ~ ' E. Gagliardo,~ Chief Operational Program Section, Division of Reactor Safety, Region IV B. L.-Bartlett, Senior Resident Inspector, Wolf Creek G. R. Bryan," Operations Specialist W. D. Johnson; Senior Resident Inspector, Comanche Peak K. M. Kennedi, Licensing Examiner, Division'of Reactor Safet *
Region IV . T. 0. McKernon, Reactor Inspector Division of Reactor Safety, Region IV D.'B.~ Waters, Reactor Systems Specialist ' G. M. Wilford, Human Factors Specialist-(' y( 4., ,
.f -Approved By:. .
f ) 16ff 4 il. E. pagliardo, Chief, Operational Programs Date Section, Division of Reactor Safety, Region IV
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8910040171 890929 PDR ADOCK 05000445 G PNU _ - - _ - _ _ _ -
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-2-Inspection Summary Inspection Conducted August 14-25, 1989 (Report 50-445/89-59)
Areas Inspected: This special announced inspection was conducted in the area of emergency response guidelines (ERGS) and included the implementation of the vendor owners' group ERGS in developing the plant-specific ERGS, technical and human factors evaluations of the ERGS, validation of the plant-specific ERGS by plant walkdowns, evaluation of the ERGS using simulator exercises, review of the ERG training program, and review of engineering and quality assurance activities related to the development and implementation of the ERG Results: The inspection team concluded that the CPSES ERGS, when used by trained operators, provided adequate direction to mitigate the consequences of an acciden Strengths _ identified included:
(1). Generally, adherence to the Westinghouse Owner's Group (WOG) ERGS and plant specific documents used to develop the ERG was good; (2) The operations staff appeared to be well trained and capable of effectively using the ERGS; and (3) The organization of the ERG packages appeared to be user friendl Weaknesses identified included:
(1) Plant ccmponents were not always clearly labeled; (2) The CPSES ERGS often did not include component numbers or specify which instrument to use to obtain data; (3) Tools and equipment (e.g., valve operating reach rods and ladders) necessary to accomplish ERG actions were not dedicated and were not always readily available; (4) The simulator did not have full modeling capability to provide dynamic exercise of procedures associated with inadequate core cooling and pressurizer thermal shock; (5) Engineering involvement in the review and development of the ERGS appeared to be untimely and all of the CPSES plant-specific setpoints were still awaiting engineering verification; (6) It did not appear that adequate lighting for performing emergency response actions would be available in some areas under accident conditions; and (7) The ERG verification and validation process only required a control room walkdown and did not include an in-plant walkdow y ym; - -
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One~ unresolved item'(questionable ability of certain safety injection isolation j
~ valves to close with a differential pressure;across them. Section 2.2) was i . identifie , " . .
Six open items that the', licensee committed to correct were:-
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_InadequatestagingofdedicatedERG_equipmentanditoolsl(Section2.2);-
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Resolution of engineering . identified inconsistencies before fuel load (Section 2.7); ' Incorporation'of appropriate engineering inconsistencies into' ERGS' and training on these items completed before achieving Mode 2 operations
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Correction of: control room meter markings' (Attachment A(13)(a));
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Physical verification that feedwater isolation reset could.be accomplished
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asdirectedbyprocedure(Attachment'A(13)(a));and
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Upgrade safety' parameter display screen integrity status trees to reflect correctinformation(Attachments (17)(a)) Other items that the applicant agreed to' review and/or correct are discussed i ' the repor Inspection Conducted August 14-25,'1989 (Report'50-446/89-59) Areas Inspected: No inspection of Comanche Peak,-Unit 2, was conducte . I J
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< SUMMARY OF SIGNIFICANT FINDINGS
. Strengths During simulator exercises, operators were capable of using the ERGS including the peacekeeping methods to track the current, applicable ERG. Also, the operators displayed a team approach and communications were goo Generally,> the operations staff appeared to be well trained and knowledgeable of the plant and the: ERGS and appeared to have. confidence in the ERG Organization of the ERG packages in the control room appeared to be user friendly; the procedure was on the right and the attachments on the left of the individual folders. Another enhancement was the inclusion of the bases and appropriate flow charts.in the ERG packag k'eaknesses L" The ability of certain safety injection isolation valves to close with a
. differential pressure'across them was questionabl *
Plant components were not always clearly labele * The CPSES ERGS often did not include component numbers or specify which instrument to use to obtain dat * Tools and equipment (e.g., valve operating reach rods and ladders) necessary to accomplish ERG actions were not dedicated and were not always readily availabl * The simulator did not have full modeling capability to provide dynamic exercise of procedures associated with inadequate core cooling and pressurizer thermal shoc l l
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Engineering involvement in the review and development of the ERGS appeared to be untimely and all of the CPSES plant-specific setpoints were still awaiting engineering verificatio I _ _
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It'did not appear. that adequate lighting fbr performing emergency response actions would be available=in some areas underfaccident conditions. --
*- - The' ERG $erificationand. validation (V&V)processonlyrequiredacontrol < .
room'walkdown and did not include an in-plant walkdow .
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CONCLUSIONS JL
~ ' -< The team concluded that the CPSES-ERGS, when used by trained. operators,-provid '
adequate direction to mitigate the consequences of an accident. However, the team identified 'a number of. weaknesses including those discussed above, which
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.To evaluate the_ ERGS, the, tea .
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1 Reviewed the ERGS and the documents used to develop the ERGS, LW * - Compared the ERGS with the Westinghouse Owners Group (WOG) ERG and d ' reviewed the applicant's justification of deviations from the p guidelines,.
*- Performed in-plant walkdowns of the ERGS, * '
Evaluated the ERGS during the performance of accident scenarios on
.: the site-specific simulator, and *
Performed human factors evaluation of the ERGS during all phases of the inspectio , The' tasks referred to:in the report are the inspection tasks described in
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TI 2515/92 and cover.the general areas identified below: Task 1.- -Basic' ERG / vendor. generic. technical guideline compariso . Task 2 - Independent technical adequacy review of.the emergency response L guideline ". . Task 3 - , Review of.the ERGS by control room and plaqt walkdown Task 4 - Evaluation of ERGS on.the plant specific simulato . Task 5'- Ongoing evaluation of ERGS,
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Task 6 Human factors evaluation of ERG The: licensee identified the CPSES emergency operating procedures (EOPs) as ERGS. -In this report, these two terms are used interchangeabl ~ DETAILED INSPECTION ' 2.1 Comparison of ERGS and the Westinghouse Owners Group Emergency Response Guidelines (Task 1) ..
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The team inspection compared the plant's ERGS with the WOG ERGS, < Revision IA; to ensure:that the applicant had g'enerated procedures in
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accordance with the WOG recommendations. The ERGS reviewed are listed in p ' Attachment D to this report. When deviations between these documents were identified, the team verified that the deviations were documented on a
, -plant-specific deviation form and, when required, a safety analysis report had been prepared in~~accordance with 10 CFR 50.5 ' '
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. Selected.ob'ervati_ons s by the^ team are the following:-
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LThiW6G[usersguidelimitsthe'WOGguidelinestoModes1and2. The WOG guidelines stated that in order to extend the. guidelines to'other
' modes, a detailed review must be-performed. Some of the applicant's-
-ERGS have been extended to Modes 3 and/or 4 (e.g.,tEmergency' Operating
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' =7 t, :Subprocedure (EOS) 0.0,.EOS 0.1, and Function' Restoration to
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" Pressure,(FRP)0.2). The applicant's detailed review in regard t [extendingthelapplicabilityoftheERGshadnotbeendocumentedon
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2 & plant-specific devia' don forms. The failure:to properly document the
& deviations; did not' result in any safety-significant problems., The
l applicant agreed to' resolve this. ite q. x
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)' - JDesignlifferences'betweenthe'CPSESand.theWOGRevision1Ageneric "
/ 9 plant resulted'in atPleast one example tof the applicant foilowing the
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w x + yWO,G guidelines when a: deviation would have'been more appropriate. . W ' Procedure;(EOP) 1.0,. Step 8.a stated " Spray pumps running," which!
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.wa'siin'accordance withithe WOG guidance;:however, CPSES containment . >
T T 4 spray (CS)Lpumps started-automatically on any safety injection Nr i j signal. .Thetgeneric plant WOG guidelines called for_ verification v '[ N ithat^thef CS ' pumps were running' to ensure that: spray was flowingzto
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: the containment. AtlCPSES,;the discharge valves would have'to be;
$1 'l . checked open to achieve the same desired results.: The applicant'
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1 4 % <> ;connitted.t.o change the procedures where this step occurs.
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4The team' concluded that the applicant had incorporated the owners group l f guidelines adequately into a plant-specific set of procedures that will
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F m enab'le the operators to mitigate emergency events, and, in _ general', that~ ""'.. , ' ithe applicant *s' deviations from the ~WOG guidelines were technically correct g . , ' Land properly' documente .
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2.2 Technical Adequacy Review of ERGS'(Task 2)
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n h,4 4 ' ' JTheiteam: reviewed the ERGS listed in Attachment D to ensure that the L ' M il - '
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guidelines provided to the applicant by the WOG. During this review, the team considered the following:
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.The prioritization of-accident mitigation strategies in the ERGS was 4 - appropriat ~* The step sequence of the WOG guidelines was followe .
V '* Procedure entry and exit points were cleariand correct, w ,
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-Transitions between and within procedures were appropriate and well le defined.
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" ' Deviations between the CPSES ERGS and,the WOG guidelines had been L
identified and justified by the applican i
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! Deviations required by the plant-specific design had bcen incorporated as necessar * 4
Notes and caution statements were used properl * Plant-specific values and setpoints were correct, and adverse contain-ment values were provide * Decision points were clear and understandabl The ERGS followed the guidance of the plant-specific ERG writer's guid ,The-team concluded that, in general, the ERGS were technically adequate and accurately incorporated the WOG guidelines. Deviations from the WOG guidelines were documented and justified. With minor exceptions, the team's findings were positive with respect to'the items listed abov ~ During the inspection, the team identified a number of technical concern Comments pertaining to these concerns are given in Attachment A to this report. The' applicant's position and planned or completed action are also summarized in Attachment A. The following are representative of the concerns identified by the tea The team selected eight plant-specific values from the procedures for review and verification. Two were adverse containment values. Five were found
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to contain errors. Errors identified are listed in Attachment The team identified several cases of discrepancies between the ERGS and their bases and one case of a discrepancy between the safety parameter display system and the ERGS. These discrepancies, which are described in Attachment A, indicate a weakness in the applicant's process of revising the ERGS and ensuring that associated documents and displays are kept consistent with the ERG An issue involving the centrifugal charging pump (CCP) safety injection isolation valves is described in Attachment A of this report. Although Attachment 2 of Procedure E0P 2.0 calls fcr these valves to be closed against a differential pressure, the applicant's response to IE Bulletin 85-03 indicated that the maximum operating differential pressure across these valves when closing was O psig. This is an unresolved item (50-445/8959-01).
A weakness in regard to the implementation of portions of the ERGS performed outside the control room is also described in Attachment A(2)(b). Generally, tools ~, hoses, and ladders have not been staged and dedicated. The applicant
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committed to perform a review to determine all the tools and equipment required, to stage these items, and to establish a method for controlling them. These actions are to be completed before Mode 4 is entered. This is an open item (50-445/8959-02).
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-10-2.3 Human Factors-Review of ERGS (Task 6)
An integral part of the ERG inspection effort was to identify human factors considerations in the implementation of the Comanche Peak ERGS. The main
- area of concern for the human factors review related to the usability of the ERG A summary of the human factors review findings follows. A detailed list-of observations from this review is provided in Attachment B along with the applicant's responses to these observation . ERG Writer's Guide-The team performed a desk-top review of the Comanche Peak ERG Writer's Guide and the ERGS to identify deficiencies in the Writer's Guide and discrepancies in the implementation of the ERG In reviewing the Writer's Guide against the ERGS and through' control room walkthroughs, the team identified a number of deviations from the Writer's Guide. Most of these deviations stem from a lack of adherence to the guidelines in the following areas:
Instruction step length and content Level of detail Component identification
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Operator aids Specific examples of deviations-in each of these areas are provided in Attachment In general, the team found the Writer's Guide was well written and provides adequate guidance for the development of ERG . Control Room The team found specific instances of indicators not supporting procedural requirements. The specific values called out in the procedure cannot be determined exactly from the indicator. Specific instances are given
. in Attachment . Local Control Actions The team identified several concerns during the in-plant walkdown These-are discussed in Attachment . Simulator Scenarios The team observed two operating crews execute three scenarios eac No significant problems were observed. Both crews exhibited high levels of knowledge of the ERGS and the plant. Communication among
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m [- v . i-11-the crew members was very good, Peacekeeping and making the transitions t within and between procedures were executed with no problems. The team noted a,few problems with specific steps. These.are discussed in Attachment . Personnel Interviews The team interviewed plant personnel to determine their understanding of the ERGS as well as their responsibilities and required actions, g< both individually and as a team. The personnel interviewed included a shift supervisor, a unit supervisor, two reactor operators, and two y auxiliary operator T In general, all personnel were highly confident that the currently implemented procedures would successfully mitigate the consequences of the broad range of accidents the ERGS envelop Selected, concerns related to improving for specific procedures and/or plant conditions expressed during the interviews included the following: The senior reactor operators and reactor operators interviewed specifically referred to Step 10 of Procedure E0P 1.0 as a step requiring careful interpretation by the operator. This comment provided additional confinnation of what was observed during the simulator exercise * Step 17 of Procedure E0P 0.0 was also cited as a step where clearer and more specific direction was neede Auxiliary operators (A0s) expressed a desire for a plant-wide labeling system, especially when considering the newly trained A0s. This would include prominent placards with direction-of-flo arrow .4 Review of E0Ps by Control Room and Plant Walkdowns (Task 3) The team conducted in-plant and control room walkdowns of the ERGS listed in Attachment D, verified that transfers were correct, and walked down the applicable sections of the supporting procedures. With exceptions, the team found that the nomenclature was consistent with operator usage and training, although specific eauipment identification numbers were rarely used. Errors and omissions that the team evaluated as potential problems for operators acting under stress are identified in Attachments A and B; 1 items of lesser importance were discussed informally with the applican l
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.The team noted weaknesses in the areas of technical adequacy, human i factors, and V&V. Weaknesses identified by the team during the technical adequacy and human factors reviews are listed in Attachments A and '
Although specific V&V weaknesses are identified in this section, it should be noted that many of the weaknesses listed in Attachments A and B also constitute weaknesses in the V&V process, which should have identified { them.
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. c -12-Except as indicated in the attachments, the team found that the indicators, 4 annunciators, and controls referenced in the ERGS were available. Two controlled sets of procedures for emergency and abnormal conditions were maintained in the control room and were available to the operators. The team verified that all control room procedures were the latest revision, Lexcept as note Paragraph 3.3.5 of NUREG-0899 states that, after they are developed, the E0ps are to undergo a process of V&V to determine if the procedures are technically acequate, address both technical and human factors issues, and can be accurately and efficiently carried ou The initial Comanche Peak V&V program was described in the procedure generation package (PGP) that was submitted to the NRC. The program at the~ time'of the~ inspection was defined in Operaticas Department Administrative Procedure (ODA)-024. The goal of the program was to demonstrate that actions'in the ERGS could be followed by trained operators to respond effectively to emergency condition The Comanche Peak program required validation when changes of intent have been made to the ERGS. The process consisted of: : Preparation of scenarios to test the ERGS * ' Validation to determine usability (level of detail, understandability)
and operational correctness (plant compatibility) in accordance with Procedure ODA-20 Verification was required for all revisions and included the following as specified in Procedure ODA-204:
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Simulator exercises
* Control room or simulator walkthrough Desk-top review for technical content and readability The applicant sponsored three one-time ERG V8V eudits conducted by the TU Electric quality assurance (0A) organization, an independent contractor, and Westinghouse; all of these took place during 1989. The team reviewed these audits and found them to be excellen The team reviewed V&V documentation on a sample basis. Generally, the forms were complete and substantive, particularly those of the multidisciplinary review tea The team, however, noted the following weaknesses: *
In-plant walkdown of the ERGS was not required. The applicant performed a one-time walkdown of ERG attachments; however, local actinn steps in the procedures were not included. The applicant committed to include complete in-plant walkdowns when control room walkdowns are require _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ -
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V6V was not required for ERG satellite procedures (e.g.3 abnormal
' " procedures (ABNs), integrated plant operating procedures (IP0s),
system operating procedures (SOPS), and alarm response procedures).
A one-time verification of these procedures was completed, and a
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one-time validation was scheduled for the next revision issued on a per-procedure basis. The applicant comitted to apply the ERG V&V process.to'the satellite procedure Verification by simulator walkdown was an approved but marginal substitute for verification by control room walkdown since some equipment items were missing and configuration control of the simulator generally lags behind that of the control room. The applicant committed to revise the procedure to limit walkdown
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verification to the. control room, not the simulato * One of the recent audits showed that human factors review of the
' Writer's Guide and the "as-labeled" plant was deficient. The team concurred. Human factors input appeared to be limited to an ODA checklist from the human factors member of the multidisciplinary team. The applicant committed to incorporate additional human '
factors. input to the ERG development proces . Inconsistent use of warnings between identical steps (e.g., Function
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Restoration to Core Cooling (FRC) 0.1, Steps 11.a and 14.a. and Emergency Contingency Action (ECA) 1.1, Step 25.b call for the dumping of steam at the maximum rate; only the ECA step has a warning about main steam isolation on high steam flow).
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The control room copics of the colored critical safety function trees were the current revision, but contained no revision number or dat The integrity tree was missing a connection logic line. An obsolete version of Procedure FRC 0.1 and the current version were both present in the shift supervisor's controlled file. These problems were resolved prior to the end of the inspectio *- Many procedures required the operator to determine normality of parameters being monitored. Control board green and yellow demarcations, which show the normal operating range were incorrect in many case The team concluded that the applicant had recognized a historical weakness
.in the ERG V&V program and compensated by conducting three specific external audits and upgrading the ERG V&V program requirements. Although the team had difficulty evaluating the impact of the one-time audits as distinct from the' ongoing program, it concluded that the end result of the V&Y process as applied to ERG, Revision 4, was satisfactory except as noted abov y -
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- - -14-2.5 ERG Evaluation Using the Plant-Specific Simulator'(Task 4)
The team conducted simulator exercises using the ERGS. The applicant provided two operating crews to assist in the exercise responses and enabled a dynamic evaluation of the capability of the ERGS to successfully mitigate
- the consequences of an accident. The objectives of the exercises included'
assessments of the peacekeeping method for making transitions between procedures and for maintaining an awareness of the steps completed within procedures, adequacy of training on procedures, and procedure flow, and actions taken. -Furthermore, the exercises ascertained physical limitations or hindrances in completing procedures and the adequacy of ti.e. information for completing the correct procedural ste Each operating crew consisted of the minimum shift complement of four licensed operators and'one shift technical advisor. Each crew ran three scenarios for the evaluation of the ERGS. The first scenario for each
. crew was a postulated spurious reactor trip and trip recovery to enable the NRC team to review the crew's' style of response and interaction. The
first crew then performed the second and third scenarios, while the second crew performed the fifth and sixth scenario The second scenario was a postulated intermediate-break loss-of-coolant accident caused by a stuck-open pressurizer safety ' valve that was of sufficient magnitude to depressurize the reactor and require shutoff of the reactor coolant pumps. The objective of the scenario was to exercise successive transitions through four procedures and arrive at a position requiring cooldown via natural circulation. During the event, the unit supervisor misinterpreted the requirements of a step in Procedure E0P The unit supervisor made the transition to a different recovery procedure instead of going back to the beginning of Procedure E0P 1.0 and making the transition out of the expected recovery procedure. The applicant was evalaating a procedural enhancement to reduce the likelihood of misinter-preting the step during a real even The third scenario was a postulated reactor trip caused by a turbine trip resulting in an anticipated transient without trip (ATWT). The ATWT was terminated by a loss of all alternating current (AC) power, compounded by a' failure of the turbine-driven auxiliary feedwater pump (AFWP), which resulted in a loss of tne secondary heat sink. The objective of the scenario was to exercise a large number of procedures and to test peacekeeping; communication; attention to notes, cautions, and foldout requirements; and the use of function restoration guidelines. The scenario was terminated once the feed-and-bleed operation was established. Operator response to this complex scenario was very smooth with no apparent problem The fifth scenario was a postulated moderate-sized loss-of-coolant accident followed by a less of all AC power, resulting in inadequate core coolin The objective of the scenario was to arrive at an inadequate core cooling situation that had to be addressed once AC power was restored. In addition, a further test of the operator response to a potential challenge to vessel
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; integrity.was . included. . Operator responses to the conditions of the scenario were satisfactory. However, the simulator experienced difficulty in. producing;the expected conditions of inadequate core cooling and . .'
. pressurized thermal shock. The event was terminated before the expected i
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end results were achieved. The applicant =had-initiated steps to upgrade the simulator models for use'in testing'the ERGS as completely as possibl The sixth scenario was a postulated steam generator tube rupture in
- conjunction with a faulted steam generator. The tube rupture was large enough to cause the operator to exercise contingency actions for a combined tube rupture and. loss-of-coolant accident. The operators diagnosed and handled the event satisfactorily, and made the transitions between procedures..'
used the procedure foldouts, and interpreted the^ procedural requirements without di.fficulty.- F Overall, the operators' demonstrated good communications and conveyed infomation on the procedures in effect to other members of the crew at transition points. Adequate training on the' ERGS was apparent, and no significant problems with the use of the procedures in mitigating accident events were note '2.6- ' ERG Training; The team assessed.the adequacy of the operator training on the utility-specific ERGS by reviewing training conducted on the. procedures during various stages L of development over the past year and the app 1' cent's plan for ERG training
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for newly licensed operators. The licensed operators received training - during two different periods. The first period'.was a 1-week session oriented toward validation of the then-current version of the ERGS on the simulator; the secono period was a 2-week session within a 5-week special training program for.all licensed shift operators. The training in each period was divided equally between classroom presentation and simulator-exercises, affording each operator a minimum of 112 hours'of ERG trainin In' addition, the applicant stated that licensed operators normally received
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4'to 8 hours of ERG training during requalification training, and had recently attended a 1-hour workshop on procedural changes incorporated in the current ~ revision of the ERGS. The team reviewed the training materials
:and the exercises conducted on the simulator and noted that, whereas . formal' lesson plans were not prepared, the training provided appeared to be sufficient to enable the operators to be adequately knowledgeable of the procedures and the background and basis for the procedural steps. The applicant was preparing lesson plans and training modules for future , initial licensed operator training programs. ' These programs will provide-28 hours of classroom training on the WOG ERG' background and 5 weeks of . classroom / simulator.. training on the ERGS. The program appeared to be satisfactory and closely linked to past ERG trainin .The training for the A0s consisted of a 4-hour introduction to the ERGS in the, fall of 1988 and a 4-hour training session for each group of A0s that was initiated at the end of July 1989 and was ongoing at the time of the . . -
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; inspection;';In' addition, the'A0s conducted plant.walkdowns'of< ERG attach-m d '
s ments and' local, action. steps to ensure consistency.with plant nomenclature and familiarity with' required actions. The training:provided to'the A0s
% : appeared-to be adequate for: familiarizing them'with the actions required
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'in,their. areas of, responsibility during accident situation .
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f2.7E Ongoing Evaluation of ERGS * (25592) (Task 5):
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m 7 , Section 6.2.3.of=NUREG-0899 requires that licensee's establish a program
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' for the ongoing evaluation of EOPs '(ERGS). NUREG-0899 further requims
?thatLthe ongoing evaluation program include 'the evaluation of the technical-n?,
? g Ladequacy of the EOPs.on the basis of operational experience and use, training Q< . experience, simulator exercises, and control room,walkthrough w
- N .The. team noted~thattin'.'a memorandum dated June 8, 1989, the applicant" >'
directed all' operations department employees;to notify the appropriate ~ personnel-of any procedural changes that are considered to be necessar '
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Procedure ODA-207,-" Guidelines for the Preparation and Review of Operations
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Procedures," Revision-2,-August 18, 1987, required (Step 6.1.6) that all
, operations procedures'(including the ERGS) be revised against the design basisiduring the development of a revision or~every 2 years. : All of the
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< active: ERGS-had been revised in 1989 and had been scheduled forLthis next - v Lbienniall review. Procedure ODA-207 provided a form (Figure 7.1) that was
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to be:used as*a checklist for' verifying'the adequacy of a new or' revised procedur Section!6.3 of Procedure ODA-207; provided ghi_delines and a form (Figure 7.2) g to be used for recommending changes, or. im The
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team reviewed selected forms (Figure'7.2)provements in' procedure on which individuals.had recom- " , ' ' mended changes.to the ERGS.c The recommendations had been' entered into the
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n ,; applicant's tracking l system,~and printouts were routed to the managers and s' J supervisors ofsthe operations department. : Applicant representatives
; interviewed stated that the Figure'7.2 form was not always returned to the '
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lindividualsubmittingtherecommendation. They noted that if the recommen ,
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'dation was. incorporated into the' procedure, it would appear in the next
'7' revision, but' if .the. recommendation was not approved, the individual i% ; making thei recommendation was1usually notified. Some of the individuals
%5 interviewed'noted that they-had not yet' received feedback on recommendations submitted..~An~ applicant' representative stated that the status of an -
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*:. ^ : employee recommendation was shown on the printout of the tracking system,
-but it was apparent that:some of the supervisors receiving the printouts
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%i were'not sharing them with all of their employees. Measures were needed
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' ;to ensure that:the originators were made aware in a timely manner of the
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, ~ys resolution of recommendations. Failure to do so can be a disincentive to
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' individuals continuing to submit recommendations, m,- . D The' applicant had also issued Procedure ODA-204, " Preparation of Emergency Response Guidelines," Revision 7, on July 18, 1989. In Sections 6.2 and
< 6.3 of Procedure ODA-204, guidelines were provided for submitting
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: recommendations' for changing the ERGS using Procedure Form ODA-204k 2. The (team noted, however, that for all of the. recommendations ' reviewed the form"
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o , ys "in Procedure ODA-207, figure.7.2, was used and not Procedure Form ODA-204- The applicant.needs to' resolve this issue and define which form is to be
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'used:>for making recommendations for changing ~ the ERG j Thefte r 1'ewed)theapplicant'sprocessforreviewingtheoriginalERGs 7 ,4 n ;and changes.theret Recentirevisions.of Procedure ODA-204 required a- '
i9 ; multidisciplinary review of ERGS.and changes to the ERGS before the fall
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of.1988; however.1the engineering department-had little or no' involvement [&1 W J < in the: review of the ERGS. At that time, the applicant developed a process
/for-reviewing all of the ERGsito determine actual'or~ potential differences wN J'
between the ERGS and the applicable design-basis. documents. . Engineering 4 0 ' ". personne1' identified a total of,298' inconsistencies that required further' review and' evaluations _ Att the time of this inspection approximately half-of the11 inconsistencies'had been resolved., The applicant committed to resolve .
, -allJof the differences or provide justification for the inconsistencies ~ :beforeifue111oad.o'This is an open item that will be reinspected before. ,
fuelload(445/8959-03). The' applicant also committed to have those 'k,
' , inconsistencies.that require revision of the ERGS incorporated into the - , : applicable ERG.and the appropriate training completed before achieving g Mode 2 operations. . This is an open item:(445/446/8959-04). .,
TheteamfoundthhtQApersonnelhadbeeninvolvedinthereviewofthe
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w ' 4 i ERGS since 1984.7 In the 1984 reviews, the-QA reviewer had used a checklist
' Lbased 'on NUREG-2005 that was generally completed with the exception of M ithat'portionf of the checklist requiring,a control room /in-plant walkdow The! team found that QA' personnel had been performing only table-top reviews L of the' ERGS and had not performed in-planttor control room walldowns. The u u- team also noted that'for the reviews performed in 1985 and 1989, the F ~NUREG-2005 checklist had not been used. The QA reviewer interviewed W stated that he used Procedure ODA-204 as a guideline for the review. The ' :4 team,also foundlthat the' resolution of the QA reviewer's comments had not: . been: forwarded to him~and he was not aware of their resolution. The team . considers:a: review without walking down at least portions of the procedure '
to.be.only minimally acceptabl The te'am also' reviewed the QA audit and surveillance activities relating
* to the E0Ps. Theapplicant.hadperformed~anaudit(TUG 89-12)oftheERGs Land' controlling documents in June 1989, but the ' audit did not include a walkdown of the ERGS.. The audit did include a review of the operator <
training program related to the ERGS, but not the monitoring of actual O , training ont the ERGS. _The team found that no QA surveillance had been
, -performed that monitored the training of the operators on the ERG "' . Applicant representatives comitted to include an annual audit of the ERGS in their master audit plans and to perform a semiannual surveillance of p simulator training on the ERGS. They also committed to walk down future . revisions to the ERG .
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2.81 Exit Meeting and Persons Contacted
- On August 25, 1989, the team and other NRC representatives held an exit meeting with; applicant personnel-and discussed the scope end findings of-the inspection. Persons contacted by the' team and attendees at the exit *
meeting are identified in Attachment . The applicant did not identify as proprietary any of the material provided to or reviewed by the team during this inspection.- t
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. #r TECHNICAL ADEQUACY REVIEW Ic +1 (The following comments stem from the technical reviews and walkdowns of the '
Comanche ~ Peak emergency response guidelines (ERGS) and supporting procedure The applicant agreed that these comments were valid and comitted to resolve all items. ' Specific commitments accompany selected item :
(1)LECA0.0: Loss"of All AC Power , Attachment 1, Step 3.B.1: .This step directed the operator to take specific . ,
llocal actions. 'In particular, the step directed DC load shed for Train "B" by transferring EXD2-1 power from Source I to Source 2 on Panel XED2-Is
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by use of a key lock switch. However, during a plant walkdown, the team noted that the Source =1 light was illuminated when the key lock switch was positioned to Source 2. .The team told the applicant that this condition was: incongruent with.the intended: function of;the key lock switch. The
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applicant stated that this-item was being tracked.under Test Deficiency Report 8702 M would be corrected before Unit 1 fuel loa (2) ECA 1.1: Lossof Emergency Coolant Recirculation-(a)' Step 8.c: In this step, the adverse containment value_ for reactor
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coolant system'(RCS) pressure was 600 psig. . The value calculated by the inspector was 615.psig. The applicant stated that its value was because of rounding in a nonconservative direction and agreed to correct the erro (b) Step 9: This step provided actions to verif from the refueling water. storage tank (RWST)y to the that there was containment no backflow sum In certain situations,-the running residual heat removal (RHR) pump-could be taking suction from the containment sump and the empty RWS If the operator was unable to isolate the RHR pump from the RWST, the s pump should be stoppen to prevent it from being damaged. The response not obtained (RNO) statement for Step 9.b did not contain this instruction. ' The applicant agreed f o add the instruction to stop the
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running RHR pum (3) ECA 3.2: SGTR With Loss of Reactor Coolant - Saturated Recovery Desired
.(a) Attachment 2: In this attachment, action was specified to fill the spent fuel pool with demineralized water to maintain level while the water normally.in the pool was being used to refill the RWST. The operator was directed to contact chemistry personnel to sample the pool for boron concentration and " add boric acid as necessary." The A0 contacted expected that chemistry personnel had the responsibility for adding boric acid. When contacted, the chemistry personnel had no procedure or knowledge of any requirement for them to add boric acid. The applicant indicated that the operators would have the
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, ' responsibility for this action and would implement clarification of and training on the procedure. Additional procedural guidance'to transfer the boric acid to the spent fuel pool area and dedication of a source of boric acid would also be provide .(b) Attachment 5: t Steps were specified to close or verify closed valves which will achieve condensate polishing isolation. The team noted that Valve 100-403, which was the bypass valve for 100-004, was located about 20 feet above the floor. . The A0 stated that an electric lift
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specifically set aside for operations use would be utilized to access the manual valve for repositioning or verifying its position. At the time of the walkdown, the electric lift was not readily availabl Also, no procedures to ensure its capability to operate, if required, (such as monthly battery checks) were evident. Additionally, in Attachment 2 to the same procedure, valves requiring reach-rod operators were not specified. Tools required to operate the valves were not dedicated, the hose to provide demineralized water supply to the spent fuel pool was not in pce, methods to deploy the fire hose used in refilling the spent fuel pool were not specified, and it was not apparent that ladders required to access equipment in overhead areas were dedicated. The applicant had committed to perform a comprehensive study to determine the tools and equipment required to be dedicated and specify how the availability of tools and equipment will be controlled. This is identified as an open item in Section of the report (445/8959-02).
(4) Procedure E0P 0.0: Reactor Trip or Safety Injection (a) The team noted the following parameter errors and disagreements in
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meter markings.- Parameters should be verified, corresponding surveillance procedures reviewed for conformance, and the individual instrument-face markings verified to be correc Entry conditions:
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Pressurizer pressure: The setpoint was 188 Instrument markings j were still set at about the obsolete value of 191 i
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Turbine trip oil pressure: The ERG listed the setpoint at 59 psi .The alarm procedure listed it at 72.3 psig (in accordance with i DCA 80907, Revision 2). The ERG value appeared to be incorrec * Any safety injection signal: Alarm Procedure 1ALB 6C of ALM-0063A i listed 1829 psig for pressurizer lo lo pressure. Technical { Specifications listed 1820 psi j
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Steam generator NR level instruments: These instruments were redlined at 44 percent. The current setpoint is 28 percen , I i
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The 2/3 steamline pressure safety injection: Initiation was at 605 psig. For readability, the emergency operating procedure (EOP) correctly uses 610 psig. The yellow field on the 16 instruments
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extended to 590 psi Attachment 2:
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Step 2.d (and other procedures using the reactor coolant pump (RCP) start criteria attachment): This step required that the operator verify that the RCP component cooling water return flows were normal. The eight meters for bearing coolers and thermal barriers were ,not marked with green and yellow band Miscellaneous:
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Step 3: Buses 1EA1 and IEA2 volt meters were outlined in blue stripes-to draw the operator's attention during a post-trip walkdown. Since the meters were not marked with green and yellow bands, the operator had to pause and read the meter, then compare it with a norm received from training or from the ERG This is an open item (445/8959-05) pending NRC review of the applicants actions to correct the above meter marking x (b) Step 3.a RNO section: This step directed the operator to transfer to Procedure ECA 0.0. Step 5.c of Procedure ECA 0.0 directed the operator to return to Procedure E0P 0.0 if power had been restored to at least one emergency AC bus at which time the operator verified that the emergency safeguards loads had automatically sequenced onto the safeguards bus. The RNO section directed the operator to manually start the safeguards pumps if the pumps did not automatically sequenc .
~ Caution 6-C-2 of Procedure ECA 0.0-directed a safety injection reset (if actuated during the procedure) to permit manual loading of equipment on an AC safeguards bu The team asked the applicant if a similar caution would be appropriate in Procedure E0P 0.0 before Step 5 (i.e., a caution requiring operators to place the specific safeguards equipment in a pull-to-lock position before attempting to manually start the pumps). The basis for this concern'was NRC IE Information Notice (IEN) 88-75 and Supplement 1, dated September 16, 1988, and April 17, 1989, respectively. In subsequent discussions with the applicant, it was learned that an applicability review in accordance with IEN 88-75 had been completed by_ September 1, 1989. The applicant could find no similar antipump circuitry problems to those discussed in IEN 88-7 (c) Step 4: This step directed the operator to transfer to Emergency Operating Subprocedure (EOS) 0.1. If Procedure E0P 0.0 was entered in Mode 3, the transfer was inappropriate because Subprocedure EOS was only applicable in Modes 1 and i
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, (d)' During the'in-plant walkdown, the team observed that the covers for the three postaccident sampling system remote modules located in th ~
810 control building switchgear room were cumbersome to remove and install and could damage the equipment. located under the covers (CPI-PSMEPS-08A,etc.)
(5)'ProcedureE0P1.0: Loss of Reactor or Secondary Coolant
' .(a) The value of 385'F for hot-leg temperature differed from.the value <
of 430'F given in-the basis for Step 15.a. The applicant had agreed
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to issue a procedure change notice to correct-this error and to review the other. ERGS-for similar errors.
< (b): . Step 14.a: In this step, the value for RWST level was 40 percen The applicant's calculation showed a calculated value of 43.5 percent, which was then rounded.down to 40 percent. The inspector's calculation '.. came to 40.6 percent. The final value in the procedure remained
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unchanged, but the calculation contained error (c) Step 15.a:~ In this. step, the value for RCS temperature was 385* In the applicant's calculation, the correct initial accumulator pressure'
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was not used. The correct RCS temperature should have been 387.5* F (rounded to a conservative value that could be read on the control room-instruments).
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. (6) Procedure E0P 2.0: Faulted Steam Generator Isolation Attachment-2f The basis-for this attachment indicated that the main steam isolation-bypass valves received a clos'e signal, but at Comanche Peak these were manual valves. . The applicant agreed to correct this error when the procedure'was next revised. The applicant further agreed to revise Procedure 0DA-204 to' include specific requirements to' review the bases whenithe associated procedure was revised. The applicant had committed to issue this revision by November 21. 198 <(7)' Subproce' dure EOS 0.1: Reactor Trip Response ' (a) Step 1: . This step required the operator to check RCS average temperatur The instrument should be specified. (The Comanche Peak Tav computer senses Tc and nitrogen-16 (N-16). Post-trip N-16 will decay rapidly, -and this Tav will equal Tc.)
L(b) Paragraph b, applicability: In the event of a Mode 3 reactor trip with no safety injection, Procedure E0P 0.0, Step 4 directed the
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operator to transfer to Subprocedure EOS 0.1. Subprocedure EOS was not applicable in Mode 3, nor had an alternative procedure been
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(c) Step 4: . The procedure was used to check pressurizer (PRZR) level greater
.# than 20 percen+.. The number required was plant-specific letdown isolation or 17 percent. The applicant committed to revise this and other identical procedural steps to change 20 percent to 17 percen (d) Step 4.b: This step should be restructured to allow the operator to select the appropriate RNO respons It was possible to have an RNO condition with both letdown and charging out-of-service or with either out-of-service. In the latter case, a mandatory sequence to-accomplish an already complete action was inappropriate and time consumin (e) Step 5, RNO b.1: Contrary to the WOG guidelines, this step manually initiated safety injection when pressure was less than normal and decreasing. There was no deviation for manual safety injection initiation, although an existing deviation did note that the procedure provided a direct transition to Procedure E0P 0.0 "because safety injection is inevitable." A direct transition to Procedure E0P was superfluous, since safety injection initiation was an entry
. condition for. Procedure E0P 0.0 and was also addressed in the introductory caution of Subprocedure EOS 0.1. The applicant committed to revise the procedure to remove manual safety injection initiation or provide a substantive. deviation justifying its retentio (f) Step 5.b. RN0: Since the two cases for this step were on separate
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pages and the word "0R" did not separate the substeps, the presentation , was initially confusing. The applicant committed to revise the' t procedure to place all of Step 5.b RNO on the same pag (8) Subprocedure EOS 0.2: Natural Circulation Cooldown
. (a) Operations Test Procedure OPT-301, Section 9.1.6: In this section the Station Curve Book was referenced. The proper reference was the Technical Data Manua (b) Note before Step 13 through Step 15: The procedure did not include a !
check to determine whether secondary makeup inventory was sufficient ! to remain in Subprocedure EOS 0.2 or whether transfer to t Subprocedure EOS 0.3 or Subprocedure EOS 0.4 was require i (c) Step 15: The bases document discussed a 27-hour wait period before cooling below 350 F if the control rod drive mechanism fans were not running. The wait period was not incorporated in the procedur '(d) Step 21: An RNO statement was not present requiring the 88-hour wait period discussed in the bases document if the fans were not availabl '
(see (9)(b) below) (9) Subprocedure EOS 0.3: Natural Circulation Cooldown With Steam Void in Vessel (With Reactor and Vessel Level Instrumentation <
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4 committed to contact WOG for information on this item and to evaluate K thefitem on a.plantes pecific basis.
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(b)l{ Step 11:: The' applican't committed to add an' RNO statement implementing (% '
:the 88-hour wait period discussed in the Comanche Peak bases document-
; .and 5he WOG ERG background-document if the fans are not availabl ..
' (: (10)'SubprocedurcEOS4.0: ' Natural Circulation Cooldown With Steam Void in Y
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(a)i Step,.1.a.in Subprocedures-E05 0.3 and EOS 0.4 and elsewhere in other r
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proceduresi The intent of the step was to establish the prerequisites o
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since-additional un,ique: coolant pump start start criteria werebut contained not to startinthe pump, subsequent
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steps. Since the prerequisites and actual start criteria were both G" contained in the same. paragraph of the attachment, the operator would probably start the pumplin Step 1.a without establishing the unique
. start criteria. : The applicant committed to revise the attachment ' ~to~ separate. prerequisites from pump start criteria,
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.(b) Steps 3.d and 5.d: These steps required thatupressurizer level be , ,
maintained using charging and letdown. . Letdown was not included in
^ > > .the step .'(c) Step 4.a: This step directed the operator to check letdown in service. The RNO statement was used to_ depressurize using a POR " .No RNO statement existed to establish letdown, although Step 6 . presumes letdown was in' servic (d) Step 21: An RNO statement was not present implementing.the 88-hour wait period discussed in the CPSES bases' document and the WOG ERG background document when the fans were not available. (see(9)(b)-
above) l f .(11) Subprocedure EOS 1.1: Safety Injection Termination P :(a) Step 9: The deviation document for Step 9 indicated that a contingency [" action had been added to locally close the centrifugal charging .4
' pump (CCP) safety injection valves if they failed to close from the '
control room. The team considered that these motor-operated valves should be capable of remote operation against the maximum operational J differential pressure. The applicant's responses to NRC IE 1
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Bulletin 85-03 indicated that the design and maximum operating { differential pressures in the closing direction for Valves 1-8801A and 1-8801B were both 0 psid. An explanatory note indicated that there was no safety-related requirement for these valves to be closed against a differential pressure. Because Subprocedure EOS 1.1 calls
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further review.:' Function Restoration to Pressure' (FRP) 0.1, Step 13' O also callsifor these valves to be closed with a CCP in service and'
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notes that'the valves may be'in a high-radiation area. This is t
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C ,# identified as.an unresolved item (50-445/8959-01) in Section 2.2 of r ,1 -this repor ~ c (b) Step 30i This step directed the operator to Integrated Plant
, 3' Operating Procedure (IPO)-007A,~" Maintaining Hot. Standby," for subsequent actions outside the ERG procedure set. A problem was4 ' "~ ' ' i noted in'that Step 30 could be attained without RCPs in service, t -which was not considered in Procedure IP0-007A. The applicant M > ' concurred in +Se team's findings and committed to' add the necessary - ' procedural step (12) Subprocedure EOS 1.2: 'PostLossofCoolantAccident(LOCA)Cooldownand a> Depressurization '
b Step 22.c: !This t step identified 15*F (45'F for adverse containment conditions) as the subcooling' margin for stopping manual'depressurization
.of;the RCS. The applicant failed to add 10 degrees to its site-specific value as, required by the WOG' guidelines. The correct value should have-been.25" F (55* F for adverse containment conditions). The team's review of the nonconservative' error. verified that it did not cause a safety-
.significant problem.-
'(13) Function Restoration to Heat Sink (FRH) 0.1: Response-to Loss'of Secondary Heat Sink
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-(a)' In Step 5.c.1: 'This step directed the operator to a-series of actions to reset.feedwater isolation if a safety injection actuation had 4 < -
occurred. Specifically, the operator was directed to pull. Universal
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Logic Card A213 in'both trains of the solid state protection ~ system logic cabinets'and then replace them to' clear the' safety injection input to feedwater isolation. The team questioned the applicant on whether'this action had been physically verified during hot functional testing conducted to date. The. applicant determined that this' physical ' verification had not yet been performed and committed to perform it before Mode'4 was entered. This is an open item (445/8959-06).
(b) Step 7.a.3: This step directed the operator to use euxiliary spray to
, .depressurize the,RCS to less than 1910 psig. If a safety injection , signal had been' received, normal charging would not be available until sthe charging was aligned. . Auxiliary spray would, therefore, not be available, requiring depressurization using the pressurizer PORV. A charging flow path was not established in the procedure until Step 27, after the feed-and-bleed operation had been effective and secondary cooling had beer, reestablished. The applicant committed to evaluate what. procedural changes were required in.conjur.ction with the WOG, ,
since the WOG guidelines specify the same action , . _._._.___ _ _ _
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t ( 3wg j '(b)JStep'5,'secondbullet:;(Theapplicantcommittedtorevisethe_'stepto: 6 s
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Jresolveanomenclature_ problem'(bottomversus'+11 inches)andallow
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' train inoperable and;one bad thermocouple'at'+11- inches on the other
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- if lights /were 11tyat'a higher level).,
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[1 ~(c) Step 5 .RNO,'andiStep'14', RNO: Dur'ing walkdowns, the operators _ were<
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was greater.than 15' ' H . H f f(d);' Step'10.b: Under certain accident sequences, this_ step will introduce nitrogen gas' (N ) to the RCS.since it incorrectly assumed that the " accumulator;inj$ction valves were closed. ' For example, if' natural
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' circulation cooldown using Subprocedure EOS 0.2 was being conducted b atithe-plant and an' accumulator ~could not be isolated at Step 16.b.
e -the accumulator would be'left, vented'with-the injection valve ope ; '
'If thereafter, the integrity tree went.to orange or red, the operator would transfer to FRP 0.1. At Step-10.b_of Procedure FRP.O.1, the
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. operator'could inadvertently add nitrogen to the RCS through the -
. supply < valve past the' open vent and'throughithe open accumulator
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injection valv l L (e) Step 13. RNO: ..During the walkdown, th'e team concluded that' radiation .T ' levels in'the 810-foot-safeguards penetration room would exceed 25 R per hour under' failed fuel accident conditions and challenge the feasibility g of local operation of the CCP safety injection isolation valves. The applicant had committed to revise the procedure to ensure as low as
, reasonably, achievable concerns were addresse .(f)_ Step 17: This step neglected the pressurizer heaters.
9 (g) During the walkdown, the team noted there were no emergency lights in Room 81 (the letdown and seal water valve room) in the 810-foot
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safeguards _ building. -(Although all valves either could be operated with reach rods or were mounted motor-operated or air-operated valves, emergency lighting was required for access and egress in accordance with-Appendix R of 10 CFR Part 50).
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1 (15)hRPid.2: . Response to Anticipated Pressur$. . d i Thermal; Shock Conditio , 1 ,
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iStep l'.b,IRN0: Although not 11sted inLth'e WOG guidelines, closing $the '
~ - - .MSIVs 'a.nd the MSIV bypasses was aJlogical:alternstiie if; the ' steam dumps; (could not be" closed..:This deviationsfrom the guidelines was documented in " the: FRP 0.2 deviation' document. However.Lthe document did not indicate if ym s -. *
cthe effect of tthe deviation on.other; procedures' had been evalu~ated. Since'
~c : the%xi.t from FRPl0.2 called for a' return .to the procedu're and step in
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" 'effect; the applicant' committed to ensure'that;all other procedures to '
Lwhich this exit might' apply'have been evaluated,with respect to the assume Op { position'of the MSIVs and the.MSIV bypasses, r
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g y Step 1: During walkdown', the operator indicated that leak identification'
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Rinvolved system-by-system isolation and monitoring for a change in sump
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level.3 The_intentLof this step was to identify the leakage' source by
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Observation of operating system parameters such as pressures, flow rate Land tank levels. The applicant committed to revise the step to clarify
< >that intent.-
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- (17)' General: Comments'
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* (a)~ During the contro1froom walkthrough..the' team noted that_the safety
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parameter display' system screen for the integrity status tree had an t
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, outdated pressure-temperature 11mit curve as compared with that in 4 the approved procedures. The applicant indicated that the computer c
- ; display would be revised to correspond with the-procedures'before
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< Mode 4 operations were started. This is an open item (445/8959-07). 1
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. (b) ;In various places in all the ERGS, the applicant used an RNO statement
; to direct local operation of: equipment. The team found that the use
.of this statement was inconsistent. For example, the applicant _,
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Appropriately used this'RNO statement in Subprocedure EOS 1.3, Steps Land 8.b but-failed to use it in others, such as Procedure ECA 0.0, a Steps.5-and 9. Procedure ECA 0.1, Step!'7.a and 2.bi-and Subprocedure EOS The applicant committed to add the RNO statement to'Subprocedure.EOS 1.4 and to ensure consistency throughout the ERGS.
'l L(c) _All of the WOG ERGS were applicable to Modes 1 and WOG conducted
'a-detailed review of mode applicability. The results were contained ' ,
in the' users guide. Table 1 of the guide listed extended mode applicability by procedure and defined the conditions that must be-
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1 met for extended applicability. In the applicability paragraphs of
;the Comanche Peak procedures, the' applicant had listed Modes:1 and 2 plus the extended modes and had incorrectly applied'the extended-mode * . conditions as prerequisites for Modes 1 and 2 as well as the extended ,, modes. The applicant committed to revise' mode applicability and "
the supporting bases and deviations as necessary to resolve this proble ,
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i f(d)ProcedureODA-204,Section'6.2.3,indicatedthatall!affected
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iprocedures would be listed in the generic procedure comparison package. This section was blank in the current version of th < ,
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generic procedure comparison package. The applicant stated that the .s affected procedures would be listed-in this section by October 21.- '. ~
:1989.. The . team agreed that this was acceptable, since the primary '
use of this section involves the long-term maintenance of the E0Ps - +
-(e) Proc'edure FRP 0.1, Procedure FRP 0.2, and. Procedure (RZ.0.2 indicated
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'that they were applicable in Modes 1 through;4. : Entry to these procedures was from Procedure E0P-'0.0 via the. critical safety function
.? integrity tree.- Procedure E0P 0.0 did not apply in Mode 4 ,and usage rules' prohibited' initiation of,the trees until' Procedure E0P 0.0,
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. Step 128, or!following an exit from Procedure E0P 0.0 to another E0 : Therefore, Procedure FRP 0.1, Procedure FRP.0.2 and Procedure FRZ'O.21 could not be implemented from Mode 4.
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s HUMAN FACTORS REVIEW
- . The following comments stem from the human factors review of the ERG Detailed comments"are grouped.by area of. concern. Comments are accompanied by , :the applicant's response and/or disposition of the concer ~
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'. T1)'. Writer'sGuide
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U Instruction Step Length and Content (a) ~ Procedure ECA 0.0, Step 9: This step required the operator to check
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if the condensate storage tank was isolated from the hotwel '
.However, the valves listed'could not be manipulated from the control-
- boards. 'The applicant committed to revise the step so that local
~ verification by the operator was indicated and had initiated a '
procedure change notic (b)J1 Procedure ECA 0.1, Step 2.b: This step directed the operator to
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establish instrument air and nitrogen to the containment. However,.
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on a loss.of all; AC power, the Train "B" air compressor must be
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started locally. The applicant committed to change this step when the procedure was next revise (c) Procedure'ECA 0.2 Step.4: This step directed the operator to manually load specified safeguards equipment on the AC safeguards bus..'The list included the containment spray pump (CSP). However, a
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review of the generic document indicated that the CSP should be placed in standby rather than' loaded onto the' bus. The applicant stated that a change will be made to the procedural step requiring
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pla' cement' of the CSP in standby. The change will become effective in
- the next revision of the procedure (d) Procedure E0P 0.0, Steps 17a and 17c: During simulator scenarios there appeared to be some confusion in regard to terminology. The term "CCP/SI" appeared in both steps. The team suggested using
. *CCP/SI" in Step _17.a and " SIP 1/ SIP 2 dischsrge flow" in Step 1 ;
'The applicant committed to evaluate this comment and submit it to
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the training personnel for further evaluatio (e) Procedure E0P-0.1, Step 10: This step required the operator to check
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if the reactor coolant system (RCS) pressure was stable or increasing and if the steam generator pressure was stable or decreasing. If
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either was different, the response not obtained (RNO) statement directed the operator to return to Step 1. The RNO statement implied that a ' secondary rupture or fault had suddenly disappeared or that a primary fault, such as a stuck-open pressurizer valve, had been correcte The inspector asked if the RNO statement should be more specific in i
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ensuring-the ' direction in which the individual pressures were going to enable the operator to make'a more deliberate decision as to whether
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.or not to return to Step 1 or to continue with the procedure. LThe applicant committed to review this step.and make changes as necessar (f) Subprocedure EOS11.1, Step 18.a. and Procedure.ECA_0.1, Step-11.a:, '
Thesestepsdirectedthe'operatortoensurecomponentcoolingwater'(CCW) flow tc the seal water heat exchanger (1-ALB-3B 1.16). The steps should also direct the operator to check the annunciator "NOT LIT."' The applicant comitted.to add "NOT LIT" to this step and had issued ia change' recommendatio (g) Subprocedure EOS:1.1,- Attachment 2, Step 1: This step directed the
. operator'to maintain seal injection water temperature below 130 ' * ' ' The operator.used the "VCT OUT TEMP" indicator (1-T1-116) to determine this1 temperature. This indicator should have been called out in the < . procedure. The applicant comitted to make this chang (h) Subprocedure EOS-1.1. Attachment 2, Step 1, 6th bullet: This :,tep ' ._did not: provide adequate direction on how to determine if the reactor aa coolant' pump motor had stopped. The applicant comitted to rewrite .this, step to include a time period the operator must wait before attempting-a restart. A change recommendation was issue " (1) Procedure FRC-0.1, Step 8: The words "in dry air" should have been ' deleted because they did not apply. The applicant agreed to reference the appropriate meter that the operator'should use and issued a change . recommendatio (j) Function Restoration to Subcriticality (FRS) 0.1, Step 4.b, RNO, .
Substep 5 for charging flow and Substep 2 for P0 charging ficw: Both substeps directed the operator to establish charging flow but did not provide flow criteria. The applicant comitted to change the 1 procedural step by adding an acceptance criterion of more than 30 gallons per minute. A change recommendation was issue (k) Procedure FRS 0.1, Step 12:- This step provided direction for isolating the faulted steam generators. The step stated, " Place turbine driven auxiliary feedwater (TDAFW) pump steam supply valves in pull-to-lock."
This statement was inconsistent with similar steps in Procedure E0P In response to this item, the applicant comitted to revise Procedure FRS 0.1 for consistency when it was next revise Level of Detail (a) Procedure ECA 0.0, Step 5, RN0: This step required emergency start of the diesel generator. This action was performed from the control board. However, no contingency actions for local emergency start of I
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. (b);l Procedure E0P 3.0, Step 30.b.2:'i This. step directed the operator to-
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ensurecomponentcoolingLwater.nonsafeguardsiloopflow.- -To' accomplish'~ _'this, the operator had to check the reactor coolant pump (RCP) thennal' l' , P
' ' barrier _ cooler component sealt return flow (RCP THBR CLR'_CCS RET FLO) ' > ' , .y ' , , ' indicators;.-The1 applicant committedit'ofinclude'this terminologyfand 3 - ', eagreed to initiate la change ~ recommendation.
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;(c)oSubprocedureEOSlil,; Attachment 2, te'p~2.g:. Ensured that the'
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, . 'overcurrent trip; selector switch for the RCPs was-in the COLD LOOP i ' position was.a< local: action a'nd should:have?been ps'ecified as suc The applicant committed to make this ch'nge:and a had initiated a .
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$ j. . ' 4 (d); Subprocedure EOS 1.2 2n$ note beforef Step _9: This step required the
' ' Doperatorf to determine if pressurizer pressure decreased to less than ' ' '1960 psig. . The operator used permissive control and interlock panels ~; ~
M . (PCIPs) 3.8 and 4.8 to make this determination. Both should have pt *
.been-referenced in this step. The applicant committed to evaluate
', ' this comment and make'the necessary changes.;
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. Component Identification' ' 3 1' .
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(a)-ProcedureLECA1.1,Stepsd.eand4.f: These steps required the operator P ,"to open'water chilled_nonsafeguards water containment'
cooling loopisolation isolationvalves and valves, component respectivel ~ '
-Theseistepsrequiremultiplevalvemanipulations,butdonotlistthe ' .A _' appropriate valves. 'The applicant committed to incorporate th e appropriate . valves and had initiated-a change. recommendatio H -(b)"ProcedureFRH"O.1, Step 7.d.2: This step directed the operator to ,
establish condensate feed flow. Valve numbers should have been-incorporated'into this step fer the feedwater pump recirculation m > valves,> condensate pump recirculation valve,'and feedwater heater bypass valve. The applicant committed to include these valve numbers and had initiated a change recommendatio (b) Subprocedure E05 1.1, Attachment 2, Step 2.a: This step directed the _ operator to verify that the' oil reservoir alarms were clear; however,
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:itLdid not list the alarms to be checked. The applicant committed i E*.c to include' the appropriate alarms and had initiated a change recommendation.
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. e' (c) Subprocedure EOS 1.1, Attachment 2. Step 2.f: This step directed the operator to verify that the seal water standpipe low level alarms were clear;.however, it did not include the alarms to be checke The. applicant comitted to include the appropriate alarms and had
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'_ initiated a change recommendatio >
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the operator was confused as to which miniflow valves were referenced.,
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v : Valve numbers should have been added. -The' applicant comitted to-
.g submit'this discrepancy.to training personnel for-further evaluation.
b f he)? E Procedure E0P 'O.0, Step i 38.a,. RNO,:12th' bullet: The procedure stated
- "W' ;"SFP cooling pump," while.the; equipment placard reads "SFPCW Pump 1-and-2."- The applicant comitted to.make the procedure match the yg designation in the plant and had' initiated.a-change recommendation.-
?(f) Procedure E0P 0.0, Attachment 2, page 3: :The labels-for. motor control center (MCC) XB1-3, Fans 17 end.18, and MCC XB4-4, Fans 23 and 24, were-m : handwritten on: tape. .The applicantLcomitted to request plant
' labels for;these components.:
. Operator: Aids -
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. (a) Subprocedure EOS-1.1,' Step 16: ;Th'is step required the operator to have M ,
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' makeup set for-greater than RCS boron concentration. An operator aid- - for ' specific point settings was suggested. ,The applicant comitted to-evaluate the aid.on the simulator. If an operator aid was not sincorportted, point settings'would be added to.the procedur m , '(b) Subprocedure EOS 1.1. Attachment:3, last bullet: The saturation curve' for 'detemining: RCS cold-leg.. temperature was not provided in ' *' ' , . the control room; The' applicant committed to determine if an operator aid or a table in the procedure was necessar Miscellaneous (a) Procedure E0P. 1.0,' Step 2.b, RNO:,.The words "IF NOT, THEN" appeared- , .in the-RNO statement. The Writer's~ Guide stated that the use of a . dual-column format implied that if'an RNO statement was not int the . action ~ expected response (AER)' column, to go to the RNO column. The applicant committed to delete "IF NOT, THEN" in the next revision .; -
of the procedur (b)' Subprocedure EOS 0.2. Step 3.a: The AER column should be structured as an "IF, THEN" statement.. The applicant comitted to make this 1 change and had issued a change recommendatio ,
' - .f (c) FunctibnRestoration1toPressure(FRP)0.1, Steps 23and24,RNO: . Since the "IF" statement'11sted two and-gated conditions, the concluding."IF NOT THEN" was not clear. The applicant comitted- ~ ' * ,
to revise these step '
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E> -5-(2) Control Room (a) Procedure ECA 1.1, Step-26, and Procedure FRC 0.2, Step 10: These steps directed the operator to check that steam generator pressures-
> were less than 185 psig. The indicator for steam generator pressure-was marked in 20-degree graduations; therefore, the operator could not accurately read this value. The applicant committed to change the procedure an'J substitute an appropriate value that could be read from the indicato (b)' Procedure ECA 3.1. Attachment 2: Units of measure were not provided on the control room indicator for containment sump level. The a'pplicant had comitted to evaluate this discrepancy and had submitted a human engineering discrepancy repor (c) The containment pressure narrow range digital readout, which had a maximum readout of 2.5 psig, had no. indication of range. Operators , .needed to be reminded via procedure or operator aid of this limitation in ' order to use other indicators to determine containment pressur The applicant committed to'suDmit a change recommendation for inclusion of an operator aid for these indicator (d). For excess letdown divert valve (XS LTDN DIVERT VLV) 1E D2-1 a control / display mismatch existed. The switch position and indicator lights should have been aligned. The applicant was aware of this problem and had submitted a human engineering discrepancy repor (3) Local Control Actions (a) Procedure E0P 0.0, Step 2, RN0: This step directed the operator to trip the turbine valve via the local trip valve located at the hydraulic control rack. l.ocally, the equipment had two labels and no operator aid to assist the operator in accomplishing the trip. The applicant committed to remove the unnecessary label and to add an operator ai (b) ' Procedure E0P 0.0, Attachment 4, page 5: This portion of the procedure directed the operator to access the postaccident sampling system panel. The operator had difficulty. locating the panel because the nomenclature on the panel was " PASS CNTNT ISOLATI0ll VALVE CONTROL PANEL." The applicant committed to make the procedure match the panel and had issued a change recommendatio (c) Procedure E0P 3.0, Attachment 4: This procedure directed the operator to access Valves IVD-403 and IVD-409. The labeling tags for these .
valves could not be read from the floor. The applicant was upgrading the i labeling system to a new high visibility labeling system, which should remedy the problem. The applicant committed to review these valves once the new system had been implemented and to evaluate the need for tools and operator aid i m.m__ _ _ .
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. - '(d) Procedure E0P 3.0, Attachment 4: During the in-plant walkdown o this attachment, the team found that emergency lighting was-not - available for all ERG-related equipment. .The applicant's response was that an emergency lighting review had been completed recently and that problem areas were being evaluated to determine corrective action " ' (e) During the plant wa'lkdowns, the team found that a" method for identifying' ERG-related components:(c'omponents specifically ,
referenced in the ERGS) was lacking. .The applicant comitted to . 4 evaluate the identification'of these components-in a distinctive '* -
.. manner. . ._ .
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ATTA'CHMENT C [ PERSONS CONTACTED AND EXIT MEETING ATTENDEES R, I Name, Affiliation (
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C.1 Alexander Reactor Operator-F :*R. L.-Ashley TU Licensing 1*J. M. Ayers ' Quality Program Manager
?T.-Bain", ,
' Shif t . Supervisor
*D. P. Barry ' Consolidated.EngineeringContractorOrganization.(CECO)
r ~ C. Bates ' Reactor Operator
"*J.;W. Bec . Vice President, Nuclear Engineering
.L .*M. R. Blevins Nuclear.0perations Support-
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T. Broughton Unit Supervisor
* J.' Buck'
L- 1 AG ..
- S. Burnett-
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Unit Supervisor t .D.' Butler" Reactor Operator-
- *W..J. Cahill Manager, Site Licensing
C. Davis- Reactor Operator ' y *D. ET Deviney Deputy Director Quality Assurance
' * Donohue Manager, OPS T. Evans; CECO - *R. Flores- ..
Operations Support Manager
*B. Garde! CAS D. Green .
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Reactor Operato *W. G.'Guldemon Manager, Site Licensing R. Guyer Contractor, Simulator Training Specialist'
* G. Hartshorn Operations'Surve111ance Quality' Assurance Supervisor , *J. C. Hicks- Licensing- *C. B. Hogg Engineering =,
- *T. A. Hope Site Licensing D x
- T. Jank Unit Supervisor
' t *J. J. Kelley Plant. Manager
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D._ Knox Auxiliary Operator T.~ Marsh Shift Supervisor R. Martinez Unit Supervisor J. McInvale- Shift Technical Advisor J.'McMahon Manager. Nuclear Training
* I. Melton Executive Assistant to Vice President, Nuclear Operations J. Metzler Auxiliary Operator M. Oliver: Training Specialist
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*E. Ottney ' CASE'
*D. L. Paris Operations' Support Engineering T. Park ' Auxiliary Operator
.M. Riggs
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Electrical Engineer, CECO
*W. Ross Senior Engineer, Operations Technical Support
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*A. H. Saunders Quality Surveillance Manager E. Skelton Reactor Operator
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* B.; Smith- '
dperations 4 *A'. B. Scott Vice President, Nuclear Operations:-
~ .*J. F.'Streeter -
Deputy Director, Quality Assurance ,w L '*W. R.'Stacha- - Management. System G. Taylo . Reactor. Operator
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G. Thatche Auxiliary Operator-
- '*M.iThero
: CASE .D. Thompson ~ Auxiliary Operator T P. Uselton Reactor. Operator *J'- E. Walke .- Nuclear Training - *D.l Walling ~ Technical' Support , *B. W.:, Wells Director, Quality Assurance *D.:R. Woodlan Licensing ' -NRC Personnel *S. D. Bitter .
Resident Inspector-0perations, CPPD
*R. J.'Eckenrode NRR/DLPG, Section Chief.HFAB *J. E. Lyons; . CPPD-AD for Technical Programs , ' *R. F. Warnick-- CPPD '*J. S. Wiebe- CPPD The team also contacted other members of the applicant's staff during the inspection period to discuss identified issue ' * Denotes *those personnel in attendance at the exit meeting held on August 25,:198 'I J + ,
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re ATTACHMENT D
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wmg 3 - DOCUMENTS REVIEWED
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[Nyg ' Revisiont Title o -
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'Resi.dualHeatRemovalSystemMalfunctio ,
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E S ABN-601A' _ 37 1 Response to "A" 138/345KV System Malfunction-
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IALM-00644- 3- 1ALB-6D (3.9; Turbine Trip 011 Pressure Low'
' , IAOR.CE1.LP[ LTraining Lesson Plan, A0RT/ Current Events'
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.. _ DBD-ME1261 11 : Design Basis Document Safety. Injection System
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00022 t li Simulator Exercise Guide, Steam Generator Tube
% Pupture T D005;.. 2 ' Simulator Exercise Guide, ATWT ' 'W .DWG El-0031 CP-4 '
Sheets 27, 27A, 28, "6.9KV Switchgear Bus IEa2, '
, Component Cooling Water PP12-Tag CP1-CCAPCC-02 -
:BKR 1APCC2 Schematic Diagram"
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4 LECA.0.0' '3 . Loss of All.AC' Power , ECA' ;3' ' Loss'of All AC Power Recovery Without Safety
#_ Injection Required ECA 0.2c ' -3 Lossiof~All AC Power Recovery With Safety
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.' Injection Required ECA-1.1: 3 Loss of Emergency Coolant Recirculation ECA-1.-21 2 LOCA Outside Containment lECA 2.1' '4 Uncontrolled Depressurization of All Steam
. - Generators
.LECA' .SGTR With Loss of. Reactor Coolant - Subcooled
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.c w- Recovery Desired '
_ _ ECA L 3., SGTR With Loss of Reactor Coolant - Saturated l Recovery Desired ECA SGTR Without Pressurizer Pressure Control M E0P. ~4 Reactor Trip or Safety Injection EOP ' Loss of Reactor.or Secondary Coolant-E0P 2.0' Faulted Steam Generator Isolation E0P Steam Generator Tube Rupture a -EOS 0.0- 2 Rediagnosis EOS Reactor Trip Response EOS . Natural Circulation Cooldown EOS Natural Circulation Cooldown With Steam Void in Vessel (With RVLIS)
* . -EOS ; 4 Natural Circulation Cooldown With Steam Void in
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Transfer to Cold Leg Recirculation"
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Transfer'to Hot Leg Recirculation H, , ,;EOS 3;15 4, Post-SGTR Cooldown Using Backfill;
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Post-SGTR Cooldown Using Blowdown '
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= Post'-SGTR.Cooldown'UsingSteam. Dump [
%7 .'s 1FRC ,4? iResponse to: Inadequate Core Cooling =
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x JFRHt 0.21 ' =3 , l Response to Steam Generator:0 overpressure w ' FRH'O.3L ,
'4- V EResponse.to Steam Generator High Level'
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_ : Response to High Pressurizer Level
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: Shock Condition 3 ; FRS.0.11 4 <
Response to Nuclear: Power Generation /ATWT FRS:0.2, 3 ~ Response-to Loss of Core Shutodwn
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Actions For Loss of Decay Heat. Removal-for ,
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, Letter 88-17. Loss of Decay Heat Removal (DHR),"
E February 10, 1989 , e T l4 .I",'
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Draft Training Lesson Plans for Initial Licansed ,
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Training Schedule and Notes for Licensed Operator ^
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, ABBREVI.ATIONE AND ACRONYMS- ' ~w ..
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m 'l i ABNi 'abnormalIprocedure ?AC alternating current
.. EAERJ, action expected response ,d AFWP'- auxiliary .feedwater pum t -
AO' auxiliary operator: ,. AORT > auxiliary' operator requalification training
~ATWT- anticipated transients without trip'
CASE . citizens? alliance for safe energy' ' Centrifugal _ charging pum '
>CCP ' . CCS ;compunent cooling' seal ~ ,
iCECO { consolidated engineering ~ contractor organization
, IFR , cCode' of Federal Regul.ations o -CLR: 3 coole .
. , , ~
CPSES' iComanche Peak' Steam Electric Station CS . c ~
=containmentisprayl : CSP; : containment (spray pum DHR; decay. heat removal < ,
y' DWG ' ' drawing-ECA,, emergency. contingency action E0P - - emergency operating procedure EOS < emergency operating subprocedure ERG emergency response guideline- <l > FRC function restoration to core cooling-FRH -function restoration to. heat sink
'FRI function restoration to RCS inventory FRP- function restoration to pressure (RCS) .FRS function restoration to subcriticality " .FRZ function restoration to containment conditions IE- ' Office.of Inspection and Enforcement IEN IE information notice INPO ' institute of' nuclear power operations IPO integrated plant operating procedure LOCA' loss of coolant accident 'LORTi licensed operator requalification training LTDN' letdown MCC motor control center 'MSIV main steam isolation valve MSL main steamline '0DA' operations department administrative procedure , OPT operations test procedure PCIP permissive control and interlock panel PGP procedure generation package PORY power operated relief valve-0 quality assurance RCS- reactor coolant system -
RCP . reactor coolant pum ' RET retur RHR residual heat removal l3
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