IR 05000445/1992057

Insp Repts 50-445/92-57 & 50-446/92-57 on 921116-19.No Violations Noted.Major Areas Inspected:Plant Emergency, Abnormal & Alarm Procedures

ML20125D717
Person / Time
Site:	Comanche Peak
Issue date:	12/07/1992
From:	Stetka T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20125D711	List: IR 05000445/1992057 ML20125D709
References
50-445-92-57, 50-446-92-57, NUDOCS 9212160004
Download: ML20125D717 (10)

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APPENDIX U.S.-NUCLEAR REGULATORY COMMISSION

REGION IV

Inspection Report: 50-445/92-S7 50-446/92-57 Operating License: NPF-87 Construction Permit: CPPR-127 Licensee: TV Electric Skyway Tower 400 North Olive Street, L.B. 81 Dallas, Texas 75201 Facility Name: Comanche Peak Steam Electric Station (CPSES), Units 1 and 2 Inspection At: CPSES, Glen Rose, Somervell Ccunty, Texas Inspection Conducted: November 16-19, 1992 Inspector: J. E. Whittemore, Reactor Inspector, Plant Support Division of Reactor Safety cW L -

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Approved: \/ W M T. F.' St'etk~a, Chief, Maintenance Section Date Division of Reactor Safety inspection Summary Areas Inspected (Unit 11: No inspection of Unit 1. activities was performe Areas Inspected (Unit 2): Routine, announced inspection of the plant emergency, abnormal, and alarm procedures, and the related programs.

j Results:

! o The established procedure program scope and content for abnormal and alarm procedures were considered good. The guidance that addressed procedure development, review, approval, and revision had functioned to ensure high-quality alarm and abnormal procedures-(Sections 2.1 and L 2.2).

o The documentation that justified the deviation of CPSES Emergency l Response Guidelines (ERG) from the Westinghouse Owners Group ERGS (generic technical guidolines)_was considered to be superior. Also considered noteworthy, was the licensee's effort at reducing the number l

of-common setpoints that were different between the unit PDR ADOCK 05000445 G FDR '

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H Additionally, the. licensee had developed the basis for each ERG step,- 1

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caution, and note, and placed this basis document at the end of each guideline (Section 2.3).

o The prt> cesses for verification and validation of CPSES ERGS in some .

' instances lacked structure and did not provide a clear separation between the verification and validation efforts. In addition, the use of unapproved checklists was observed. The licensee was informed of these observations for consideration during their ongoing verification and validation process (Section 2.3).

Att.achments:

o . Attachment 1 - Personnel Contacted and Exit Meeting o Attachment 2 - Documents Reviewed

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DETAILS 1 PLANT STATUS During this inspection period, Comanche Peak Steam Electric Station-(CPSES)

Unit 2, was undergoing the final phase of integrated plant testing required for initial licensing and fuel loa EMERGENCY PROCEDURES (42452)

The inspection involved a review of emergency, abnormal, and alarm procedures, and related programs to ascertain whether the procedures had been developed in accordance with regulatory requirements, and whether procedure changes were reviewed, approved, and controlled in accordance with the administrative controls requirements and current draft license requirements. Further, the inspection involved the review of selected procedures for adequacy of technical content and verification of procedurc conformance to format requirements.

l 2.1 Abnormal Conditions Procedures (ABN)

The inspector reviewed a sample of abnormal conditions procedures for compliance with required format and for technical adequacy. Additionally the appropriate program administrative control procedures were reviewed to assess the integrated processes for preparation, review, approval, and change or revisio CPSES had a total of 68 abnormal conditions procedure The licensee had plans to have a total of 70 iBNs prior to licensing the unit. Of the 70,'45'

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would be common to both units, 13 would be specific to Unit 1, and 12 specific-to Unit 2. An example of procedures that were different or specific-to one unit included procedures for turbine . plant cooling water malfunctions, fires in various structures or spaces, loss of plant computer, and nionitoring of loose part monitoring system alarms. These separate or unit specific-procedures reflected differences between the units. The unit differences had been identified in &ccordance with Administrative-Procedure STA-820,

" Reporting:and Evaluating Unit' Differences," Revision 1. The. inspector focused mainly on assessing the technical content and differences for those unit specific procedures addressing a common conditio The procedure review did not reveal any concerns for adequacy of technical-content. Those procedures reviewed contained the properly sequenced steps, cautions, notes, and transitions to address the event or condition of-concer Each ABN was entered by the identification of symptoms which were classified as either alarms or plant indications. Automatically occurring action _ had -

been correctly identified, where-appropriate. Steps within procedures that needed to be addressed initially were identified by a circle around the step number.

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-4-The preparation process for-newly developed ABNs was- conducted in accordance '

with Procedure ODA-203, " Preparation Of Abnormal Conditions Procedures,"

Revision 8. This document se*ved as a. writers guide and prescribed procedure format and layout. The licensee had-decided to place _a majority of the ABNs-in a two column format that emulated the ERGS. Approximately 21=ABNs were not being changed as they had been judged to be better adapted to the original forma The review progress for ABNs was conducted in accordance with ODA-207,

" Guidelines For Preparation And Review Of Operations Procedures," Revision 5, and STA-205, " Changes.To Procedures," Revision 15. The minimum review of ABNs _

required a technical review within the operation support group (peer review),

followed by quality assurance (QA) and radiological protection (RP) revie Problems arising from -these reviews could result in an engineering revie The inspector noted that licens arsonnel used desktop procedures and '

checklists to supplement the de; .pment and review process. _0ne of these ctacklists appeared in the form of a walkdown checklist which appeared to validate local action steps. The inspector noted that this checklist form'was not part of any of the controlling procedures used.in ABN development or revision, and, therefore, had not been officially approved. All ABNs designated as safety-related required a screening _ effort to determine'if an-unresolved safety question existed. Routinely, those procedures designated-as safety-related were reviewed by the Site Operations Review Committee (SORC).

The inspector concluded that the ABNs for Unit 2 would fully support unit operation. The procedures were technically adequate and usable for the-mitigation of abnormal conditions. The review process was judged adequate to assure a high quality product at the end of the process, 2.2 Alarm Response Procedurer (ALM)

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The inspector reviewed-a sample of alarm response procedures for compliance with required format and for technical adequacy. Additionally the appropriate program administrative control procedures were reviewed to assess the integrated- processes for preparation, review, approval, and change or-revisio The licensee had developed Unit 2 alarm procedures in accordance with Procedure ODA-205, " Preparation Of Alarm Procedures,f' Revision 7. This procedure allowed procedures to be developed in two different formats. : A separate and unique format was used for-alarms originating in the Digital-Radiation Monitoring System. ALM procedure revisions were performed.in accordance with Procedures ODA-207, " Guidelines For-Preparation ' And Review Of Operations Procedures," Revision 5, and STA-205, " Changes To' Procedures,"

Ravision_15. Revisions to alarm procedures originated primarily from design modification, identified unit differences, or changes to the unit setpoint documen Review and approval of ALMS was identical to the process used for ABN . . _ _ - ~. _ _ . _ _ _ _ _ _ _ _ . _ . _ _ . . _ . . . . . _ _ _ , _ , _ . _

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-5-The inspector did not identify any technical concerns with the large sample of ALMS reviewed. All procedures reviewed were in the correct format and technically correct. Further, the integrated preparation, review, revision, and approval process had functioned to provide satisfactory alarm procedure However, the licensee had used the same unapproved validation checklist that had been used in the review process for the ABNs and noted in Section 2.1 of this repor The inspector determined that the Unit 2 alarm response procedures would fully support operatio .3 Emergent _y Response Guidelines (ERG)

During this inspection, the licensee was in the final stages of approving the CPSES specific ERG Procedure development was complete and comments developed during the local review process had been implemented or otherwise ,

dispositioned. The licensee was awaiting the results of a final vendor review before approving and implementing the Unit 2 procedures. To avoid confusion, this report will refer to the CPSES ERGS as the emergency operating procedures (E0Ps) throughout. This is being done to prevent the reader from confusing the CPSES ERGS with the Westinghouse Owner's Group (WOG) ERGS, Revision IB, which were used as the generic technical guidelines for the development of the CPSES ERGS (EOPs).

The licensee had developed the E0Ps in accordance with procedure ODA 204,

" Preparation Of Emergency Response Guidelines," Revision 9. This document, also referred to as the E0P Writer's Guide, had been developed to include the requii aments of the WOG Technical Guidelines and the licensee's own administrative requirements. The procedure addressed all aspects of E0P development, including review and approval. The inspector performed a detailed review of this procedure and the foll.ving Unit 2 draft E0Ps and

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related development packages:

ERG E0P-0.08, " Reactor Trip Or Safety Injection" ERG E0P-3.0B, " Steam Generator Tu'oe Rupture" ERG EOS-3.2B, " Post SGTR Cooldown Using Blowdown" ERG ECA-0.08, " Loss Of All AC Power" ERG FRS-0.lB, " Response To Nuclear Power Generation /ATWT During the review of the E0Ps and their related development packages, the inspector had three observations of licensee initiatives that resulteu in improvements in the final version of the E0P The licensre had performed an analysis to determine the basis for each step in the procedures. The basis was also determined for all cautions and notes that preceded affected steps. This analysis had been assembled in a sequence identical to the procedure steps, notes, and cautions, and then provided as an attachment to the procedure. Licensee management felt that if conditions ever resulted in an impasse, an operator would be aided in the decision making process by understanding the basis of the specific note, caution, or ste The licensee's effort to provide a basis document for individual procedure steps was considered to be a noteworthy initiativ __ _ _ _ ____ _ ___ ___________________ .. ._

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-6- l-The licensee had developed a procedure data package which included i justification documentation for all deviations from the WOG ERG Justification was provided for E0P steps that did not exactly match identical steps in the WOG ERGS, ERG steps not included in the E0Ps, and E0P steps that were not referenced by the ERGS. Additionally, documentation was provided for

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any cautions or notes that the licensee had placed.in the E0Ps due to

+achnical, administrative, or draft license requirements. The basis for iation between CPSES E0Ps and the WOG ERGS was primarily a result of

..ysical differences between the generic plant used for developing the WOG ERGS and CPSES Unit 2. The data packages also referenced each setpoint to a controlled setpoint database to allow reviewers to determine the basis of-setpoints or parameter values used in the E0Ps. It was inte'nded to have these data packages signed and approved by the appropriate manager as part of-the final E0P version approval. A review of the data packages for the above listed procedures did not lead to any concerns about the observed deviations-and related documentation. The inspector considered the documentation to be-superio The licensee had identified every setpoint or parameter value that was to be included-in the E0Ps. Once the need for a setpoint/value had been identified, a value was determined through engineering analysis and calculation. The value determination had addressed the instrumentation system design inaccuracies and ambient conditions. Setpoint calculations used conservatism to ensure that the parameter being considered remained within the design envelop The inspector reviewed a sample of setpoint derivations and did not identify any anomalies. The inspector observed that the licensee had attempted to reduce the number of common setpoints that were different on each unit by establishing one setpoint if a conservative approach could be applied. Fo example, the-E0Ps contained a value for safety injection (SI) pump shutoff head, that appeared throughout the-E0Ps. Normally, . the -operator would - be required to decide subsequent action depending on the relationship between the .

current reactor coolant system (RCS) pressure and this value. - Originally the-shutoff head for Unit I-SI pumps had been de mined to be 1686 psi and Unit 2-had been 1658 psi. The licensee performed .a analysis and determined that because of the control room indication, the setpoint for both units could be conservatively changed to 1700 psi to make the common setpoints identica Conversely, two different values of RCS pressure were used as-the maximum RCS -

pressure for allowing the operator to commence residual heat removal (RHR)

system cooldown. The values -for Unit I and 2 were 300 psi and 325 psi-respectively. The licensee concluded that the value for Unit 1 could not be raised to 325 rsi because of. physical differences between the RHR system Further, it was not considered conservative to delay the . initiation of RHR cooling during ccident conditions on Unit-2 by lowering the setpoin Therefore the licensee made the decision to leave the setpoints differen The inspector concluded that the licensee's. effort to minimize the number o different setpoints between t' + units was-beneficia . - - - . . .. --- - - . - -_~. .- - - - . - . .

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-7-The E0P Writer's Guide contained instructions for the licensee to verify and validate the E0P The procedure differentiated between the two. processes and loosely defined verification as a technical review and validation as a performance revie However, in a review of the data that documented the processes, there was not a clear indication whether the two processes had occurred separately or independently. For example, a 7 item checklist entitled " Verification Checklist," from the E0P Writer's Guide, hed been used to document validation of the E0Ps on the simulator. Only 4 of the 7 items had been addressed by the validation team. A licensee representative stated that the remainder of the checklist items were to be addressed by the procedure writer. The inspector reviewed the three remaining items and determined that they represented verification efforts, (e.g., comparison with technical guidelines, writer's guide, and ERG flow charts). This practice allowed the procedure writer to complete the verification checklist, however, this effort should have been a separate independent revie For validation of local action steps, the licensee used the identical unapproved checklist that had been used for walkdown review of ABNs and ALM However, since this checklist was totally performance oriented according to the available documentation, it appeared as if the local action steps had not been verified. Unapproved instructions were also used to address the multi-discipline team review and use of the controlled setpoint document. Also two non-proceduralized checkl4sts were used to ensure the completeness of E0P development er revisio The inspector walked down the local actwn steps of attachments 1 and 2 to ERG E0P-3,08, " Steam Generator Tube Rupture." These attachments addressed isolating the affected steam generator and re-establishing reactor coolant pump (RCP) seal water flo These procedures were technically correct and useable. The inspector informed the licensee of problems associated with page numbering and locations. The most significant obser;ation was that the licensee was not consistent in identifying, within the procedures, those steps which were to be performed locally. However, licensee personnel involved in the procedure walkdowns were readily able to identify local action. The word local was used sporadically in steps where local action was intended, and there were instances where the location was not pecified for steps that were to be performed locally. The inspector informed a licensee representative, that based on the inconsistencies identified during the review of the two attachments, it appeared that a thorough human factors review of local action steps had not been performe The licensee's verification and validation processes uid not fully address local environmental conditions for action steps that were performed locall The local action walkdown validation checklist only addressed local lighting conditions and equipment access. Local action environmental conditions related to ambient temperature and personnel radiation exposure had been addressed outside of the validation and verification proces Coping studies for total loss of AC power (blackout) had analyzed and determined local temperatures during the most severe conditions. An administrative procedure to address operator heat stress had been implemented to address overheating

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-8-concern Additionally, in order to assess local radiological conditions, tht licensee had performed an analysis to determine post-accident vital area mission radiation doses that would be received by personnel during the performance of required local action steps, under design basis accident conditions. This analysis had provided nodal dose rates for all identified tasks, which ircluded exposure during transit to and from a common starting point. Transit times had been determined by conservative estimates of average travel rates through the particular distances and spaces in feet per second, dependent on whetner the path section was a floor, stairwell, or went through a fire or security doo The inspector asked the licensee to provide bases for the times that had been designated as exposure time for personnel to perform a given mission. These mission performance times had been taken from the analysis performed for Unit 1 and were listed in the Final Safety Analysis Report (FSAR), as the tasks were identical for the two units. F'- 2 of the 17 identified missions, the licensee could not provide the bases ) the times designated to perform the missions or task The other performance times had been verified by survey or validated by walkdown. At the conclusion of the inspection, the licensee was still conducting a search for documentation that would provide the bases for the identified mission performance time During the week following the inspection, the licenset provided documentation to the inspector that validated the designated mission performance times. The licensee's apprc:ch had been to perform timed walkdowns of the tasks for all 17 missions. Where appropriate, multiple operator surveys were performed to

, verify accuracy of the performance time Finally, the walkdown and survey times were compared to the times that nad been utilized in the FSAR. This total effort served to prove that those times specified in the FSAR were conservativ .4 Conclusions  :

The inspector concluded that the licensee's Unit 2 ABNs, ALMS, and E0Ps were adequate to support operation. It was also noted, however, that unapproved checklists were being utilized in the verification and validation proces This observation was discussed with licensee representatives for consideration with the ongoing validation and verification activities, e

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ATTACHMENT 1 ,

1 PERSONS CONTACTED

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1.1 Licensee Personnel

• B. Adams, Instrument and Control Engineering Supervisor C. Bearden, Procedure Writer fR. Braddy, Assistant Project Manager, Unit 2

• J. Brau, Supervisor, Operations Support D. Browning, Alarm Procedure Coordinator

• R. Carter, Maintenance Engineer

• J. Donohue, Manager, Operations

• E. Dyas Auditor, Nuclear Overview G. Gerniin, Senior Engineer

• J. liitzfeld, Technical Programs Project Management Lead

• J. Kelley, Vice President, Nucl;ar Operations

• D. Kross, Unit 2 Operations

• E. Luengas, ISEG, Nuclear Overview

• J. Martin, Senior Engineer, ISEG

• D. McAfee, Manager, Quality Assurance

'G. Merka, Licensing Engineer

• D. Moore, Manager, Maintenance

• S. Palmers, Stipulation Manager

• D. Pendleton, Unit 2 Projects

• Phillipi, Supervisor, Quality Engineering

• C. Rau, Project Manager, Unit 2 A. Shediosky, Procedure B. Smith, Emergency Response Guideline Coordinator

• D. Wilken, Unit 2 Maintenance

• L. Wojcik, Supervisor, Nuclear and Mechanical Engineering Analysis 1.2 CASE ,

• 0. Thero, Consultant 1.3 NRC Personnel

• Constable, Chief, F nt Support Section

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• D. Graves, Senior Resident Inspector l *B. Holian, Senior Project _ Manager

! *C Johnson, Reactor inspectrr l *D. Skeen, Reactor Inspector l

• Denotes personnel that attenued the exit meetin In addition to the perrilel listed above, the inspectors contacted other personnel during this ins uct'on perio EX11 MEETING An exit' meeting was conducted on November 19, 199 During this meeting, the inspector re'ziewed the scope and findings of the inspection. The licensee did not (dentify as proprietary, any information provided to, or reviewed by the irispector.

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ATTACHMENT 2 DOCUMENTS REVIEWED Abnormal Conditions Procedure (ABN)-706, "Presselrizer Level Instrumentation Halfunction," Revision 3 ABN-710. " Steam Generator Level Instrumentation Malfunction," Revision 6 ABN-808A, " Response To fire in Service Water Intake Structure," Revision 2 ABN-808B, " Response To Fire In Service Water Intake Structure," Revision 0 ABN-906A, " Loss of P2500 Or ERF Computer," Revision 2 ABN-910A, " Loose Parts Monitoring Alarms, Revision 0 ABN-9108, " Loose Parts Monitoring Alarms, Revision 0 CPSES Emergency Response Guideline (ERG) E0P 0.08, " Reactor Trip Or Safety injection," Revision 0 CPSES ERG E0P 3.0, " Steam Generator Tube Rupture," Revision 0 CPSES ERG ECA 0.08, " Loss Of All AC Power," Revision 0 CPSES ERG FRS 0.1B, " Response To Nuclear Power Generation /ATWT," Revision 0 Westinghouse Owners Group (WOG) Technical Guidelines (TG) for ERG E-0,

" Reactor Trip Or Safety injection," Revision IB WOG TG For ERG E-3, " Steam Generator Tube Rupture," Revision IB WOG TG For ERG ECA-0.0, " Loss Of All AC Power," Revision IB WOG TG for ERG FRS-0.1, " Response To Nuclear Power Generation /ATWT,"

Revision IB

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