IR 05000445/1989061

Insp Repts 50-445/89-61 & 50-446/89-61 on 890724-0804.No Violations or Deviations Noted.Major Areas Inspected:Final Reconciliation Process to Assure That Pending Open Items Addressed within Corrective Action Repts

ML20247K944
Person / Time
Site:	Comanche Peak
Issue date:	09/14/1989
From:	Ashe F, Latta R, Livermore H, Norkin D Office of Nuclear Reactor Regulation
To:
Shared Package
ML20247K939	List: IR 05000445/1989061 ML20247K935
References
50-445-89-61, 50-446-89-61, NUDOCS 8909220074
Download: ML20247K944 (48)

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• ' * 'U.- NUCLEAR-REGULATORY COMMISSION

. OFFICE OF NUCLEAR REACTOR REGULATION Permits: CPPR-126-

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NRC' Inspection Report: 50-445/89-61 y,

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50-446/89-61 CPPR-127

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Dockets: 50-445 1 50-446  !

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Construction Permit' q Expiration Dates: ;

Unit 1: August 1, 199 j Unit'2:. August 1,'1992 j l

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. Applicant: TU Electric-Skyway _ Towe !

400 North Olive Street l Lock Box 81 i Dallas, Texas 75201 j l Facility'Name:. Comanche Peak Steam Electric Station (CPSES), j

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Units 1 & 2 TInspection'At: Comanche Peak Site, Glen Rose, Texas j

Inspection Conducted: July 24 through August 4, 1989 i

Team Leader: -

h 2.___ I~#Y~i Mon _ P. Norkin, Senior Reactor Engineer Date j

/ (System interaction, HVAC systems, 1 steam generator or' upper lateral .)

. restraints, and mechanical.)

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Inspector: 'Yb $I' l $ ~ 9 I i

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F. Ashe, Electrical Engineer Date

.(Instrumentation and Controls [I&C))

Inspector: / ~

R. M. Latta,. Resident Inspector

/[6- pq Date (Electrical and I&C)

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Consultants: J. Birmingham, RTS (Quality Assurance) Chen, Parameter (Piping and pipe supports)

J. Dale, EG&G (HVAC)

W. Richins, Parameter (Civil structural)

P. Stanish - Parameter (Conduit and conduit supports)

Reviewed by: - MAL 7 - N-4 H. H. Livermore, Lead Senior Inspector Date Inspection Summary:

Inspection Conducted: July 24, 1989 through August 4, 1989 (Report 50-445/89-61; 50-446/S9-61)

Areas Inspected: Team inspection of the final reconciliation process which assures that pending open items, calculations with confirmation required, interacting inputs, and applicable deficiency documentation, such as nonconformance reports (NCRs), corrective action reports (CARS), and significant deficiency analysis reports (SDARs) are appropriately addressed within the Corrective Action Program (CAP). This objective was accomplished by inspections of selected design validation packages (DVPs) to verify that design input such as loadings, equipment data, and environmental factors are resolved to assure that as-built conditions are acceptable and consistent with design. The team performed reviews of analyses and calculations including nuclear steam supply system (NSSS) interface design criteria and the impact of NCRs, design change authorizations (DCAs), and design changes on calculations. Further, the team evaluated the TU Electric audit program for assurance that resolution of Technical Audit Program (TAP) findings and Engineering Functional Evaluation (EFE) action items were satisfactor Results: Within the areas inspected, no violations or deviations were identified. One open item with two examples was identified (paragraph 8, pages 25 and 26). The item concerned whether operational impact was considered in the calculation proces The NRC team found that the final reconciliation program was well conceived and implemented. Confirmation required items were clearly identified on the calculations and the documentation for subsequent resolutions was concis In areas such as piping / supports, conduit / supports and HVAC duct / supports, there are well defined

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systems in place to ensure proper interfaces, resolve confirmation required items, and assess the impact of design criteria revisions, NCRs, DCAs, etc., on existing calculations. The NRC team found that these systems were well execute Finally,-the NRC team found that internal audits performed by TAP and EFE provided an effective means for reviewing the CAP, including design validation, PCHVP and final reconciliation. We found that the audit reports were based on thorough review plans, the conclusions were supported by audit detail and issues identified in the audits were resolved in a technically acceptable manne l i

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i Details

] Persons Contacted l

• D. A. Barry, Sr., Manager,. Engineering, Stone and Webster 1

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Engineering Corporation (SWEC) .

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• J. W. Beck, Vice President, Nuclear Engineering, TU Electric

• R. A. Berry, Licensing Manager, Consolidated Engineering and Construction Organization (CECO)

• H. D. Bruner, Senior'Vice President, TU Electric

• W. J. Cahill, Executive Vice President, Nuclear, TU Electric T. C. Chen, IMPELL

'*C. Y. Chiou, Ebasco

• C. B. Corbin,, Licensing, TU Electric

• J. T. Conly, APE-Licensing, SWEC H. Crockett, HVAC Supervisor, TU Electric G. Dean, Lead Engineer Civil / Structural, SWEC Q. B. DuBois, EFE Completions Supervisor-Quality Assurance (QA)

• D. R. Ferguson, Operational Readiness

• C. A. Fonseca, Deputy Director, CECO N. Goldstein, Lead Engineer Technical' Issues, SWEC

• W. G. Guldemond, Manager of Site Licensing, TU Electric

• J. C. Hicks, Licensing Compliance Manager, TU Electric

• T. A.. Hope, Licensing Supervisor, TU Electric

• D. M. McAfee, Manager, QA, TU' Electric

• S. G. McBee, NRC Interface, TU Electric

• A. Meinershagan, Ebasco

• J. W. Muffett, Manager of Engineering, TU Electric

• J. Nandi, Engineering Supervisor Pipe Supports, TU Electric F. Ogden, Assistant Project Engineer Technical Issues, SWEC G. P. Phillipi,. Supervisor Licensing Closure Group, SWEC

• H. Rains, Cable Tray Hangers, CECO

• D. L. Ranstrom, Supervisor, Technical Audits, TU Electric

• P. Raysircar, Deputy Director / Senior Engineer Manager, CECO

• S. L. Stamm, Technical Advisor, CECO

• J. H. Wawrzeniak, Deputy Director, CECO

• A. C. Wong, Supervisor, Conduit Supports, Ebasco The NRC inspectors also interviewed other applicant employees during this inspection perio * Denotes personnel present at the July 28, 1989, exit meetin . Introduction and Summary Background The CPSES Project Status Report for each discipline of the CAP has a section titled " Final Reconciliation," which provides the basis for the below discussion. The purpose of final reconciliation is to consolidate the design

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, validation analyses, hardware modifications and L inspection documentation to assure design as-built consistency. Final reconciliation incorporates Post-l Construction Hardware Validation Program (PCHVP) results and resolution of hardware related Comanche Peak Response Team (CPRT) and external issues. Final reconciliation also includes confirmation that the interfacing organizations have verified that interfacing design results are compatible with'their validated desig In addition, open items, observations, and deviations related to the CAP that were identified by TAP and EFE or in SDARs are resolved as part of final reconciliation.

l l b. Method of Review Final reconciliation is a process which ensures by the consolidation of analysis, hardware modification, and inspection documentation that the design and hardware are complete and consistent. As such, reconciliation requires i completing open items such as confirming design inputs and

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assumptions and resolving open items resulting from inspections by TAP, EFE, CPRT, and the NRC. Ensuring consistency involves matching as-built data obtained by the PCHVP with design inputs and reviewing NCRs, DCAs, and l

design criteria changes to assess their impact on design calculation This NRC team inspection covered most of the CAP disciplines, including work by SWEC, Ebasco, and IMPELL.

i Team members reviewing SWEC and Ebasco Systems Interaction l

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reviewed examples of resolving " Confirmation Required" statements in calculations. Generally, these statements involved design inputs and assumptions that required confirmatio For the Ebasco and IMPELL areas of

, conduits / supports, cable trays / supports, and heating /

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Ventilating / air-conditioning (HVAC) duct / supports, the NRC team reviewed the impact upon calculations due to criteria changes on clamps, changes to cable fill and thermolag, NCRs and DCAs. For the Ebasco HVAC systems work, the team reviewed the consistency of the design basis documents (DBD) with the implementing calculations for three HVAC system In addition to the above areas, the team reviewed the TAP and EFE program implementations with special emphasis upon the closecut process for inspection finding I c. Summary of Results The NRC team found that the final reconciliation program I was well conceived and implemented. Confirmation required l

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items were clearly identified on the calculations and the documentation for subsequent resolutions was concise. In areas such as piping / supports, conduit / supports and HVAC duct / supports, there are well defined systems in place to ensure proper interfaces, resolve confirmation required items, and assess the impact of design criteria revisions, NCRs, DCAs, etc., on existing calculations. The NRC team found that these systems were well execute Finally, the NRC team found that internal audits performed by TAP and EFE provided an effective means for reviewing the CAP, including design validation, PCHVP and final reconciliation. We found that the audit reports were based on thorough review plans, the conclusions were supported by audit detail and issues identified in the audits were resolved in a technically acceptable manne Within the scope of the inspection, no violations or deviations were identifie . Mechanical (50075, 50055, 37051, and 37055)

The following are calculations reviewed by the NRC team in order to evaluate the resolution of confirmation required item ME(B)-053, Revision 0, " Auxiliary Feedwater System Performance" This calculation uses a hydraulic model to analyze various configurations of auxiliary feedwater system alignment to ensure the system meets its design basis. The confirmation required the validation of three auxiliary feedwater system flow diagrams. These drawings have been validated and the NRC team agrees with the closecut of the confirmatio ME(B)-063, Revision 0, " Auxiliary Feedwater Automatic Controller Setpoints" The purpose of this calculation is to determine the optimum setting for the instruments that switch the pressure control valves to automatic control and the instrument that controls the valves once they are in the automatic mode. Confirmation was required for the " Conclusions" for the calculation which indicates the'setpoints for four instrument groups. The setpoint determinations are based upon input from calculation 16345-ME(B)-053 (above) which determined maximum pump flow to the steam generators and minimum flow in the event of a line break. The calculation file indicated that there had been no changes to the calculation since the confirmation requirement was identified, and that the " Conclusions" did not actually require confirmation. The NRC team agreed with this position.

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16345-ME(B)-187 Revision 0, "Condensato Storage Tank Capacity Requirements" This calculation established the quantity of water to be sent from the hot well to the condensate storage tank (CST) due to condensate rejection, maximum surge, and maintenance. The calculation also established the maximum temperature of water transferred to the CST, setpoints for demineralized water makeup and a level control valve, and whether the hot well inventory meets design basis capacity requirement Confirmation was required that the 3.33 minutes supply of hot-well inventory is satisfactory. Revision 1 to the calculation recalculated the hot-well inventory and determined that it provided a 5.12 minutes supply which exceeded the minimum requirements stipulated in the DBD (5 minutes) and the condenser specification (4.94 minutes). The NRC team agreed that this resolved the confirmation ite Revision 0 of the calculation indicated that the present setpoint for 1-LB-2478B (855') did not provide sufficient volume within the CST to accommodate hot-well rejection or surge without overflowing the tank. The calculation determined that a setpoint of 853' would resolve this matter. SWEC plans to amend the calculation to indicate that this confirmation item has been resolved based on resetting 1-LB-2478B to 853'.

The NRC team agrees with this actio ME(B)-073, Revision 2, "CCW Surge Tank Volume" The calculation verifies that the volume of the component cooling water (CCW) surge tank is adequate for system operation and considers the partitioned volumes for Trains A and B. The confirmation items are various references such as the specification for the spent fuel pool heat exchanger, CCW system flow diagrams, and several calculations. Confirmation actions included component validation for the spent fuel pool heat exchanger and validation of the above references (including subsequent revision numbers). The NRC team agreed that the items were resolve ME(B)-130, Revision 2, "CCW Surge Tank pressure" The calculation determined the vent line capacity of the surge tank to ensure that the tank does not exceed design pressure during worst case system inleakage to the tank. It also evaluates the tank vacuum developed due to worst case system outleakage. Revision 2 requires confirmation of the assumption l that all vents from the tank have identical piping layouts; I e.g., the inlets and outlets to all the vents are at the same elevation and subjected to equal static pressure. The Revision 2 conclusions indicated that additional relief capabilities were needed to prevent overpressurization of the

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1-CCW surge tan To accomplish this, DCA 49129 identified modifications to install two 4-inch vent lines and one 2-inch vent line on the surge tank. The effect of DCA 49129 was to supersede _the above vent line assumption; therefore, the NRC team agreed with action to delete the confirmation requiremen ME(B)-181, Revision 1, "CCW Heat Loads and Temperature l

for Various Operating Modes" This calculation determines the.CCW heat exchanger inlet and outlet temperatures and then calculates CCW temperatures downstream of equipment and CCW return line temperatures. The required confirmations involved validation of reference documents. These were accomplished by SWEC validation of Gibbs and Hill (G&H)' documents and by referring to validated revisions of documents originally identified as requiring confirmation. The NRC team agreed with this proces ME(B)-162, Revision 0, " Determination of Set Pressures and Capacity Required For Leakage Flow Relief ValveJ Being Added to the Suction Lines of Auxiliary Feedwater Pumps" The objective of this_ calculation is to determine the setpoint and capacity of the leakage flow relief valves which were added for overpressure protection of the suction piping. Revision 1 indicates that three auxiliary feedwater system flow diagrams were confirmed by means of citing validated revisions to the document Input 3 to the calculation, the set pressure for the three leakage flow relief valves (50 psig), was confirmed by validating calculation 16345-ME(B)-022. The NRC team agrees that these items have been adequately confirme ME(B)-203, Revision 2, " Set Pressure Calculation For Existing Relief Valves in the CCW System" The calculation determines whether set pressures of existing relief valves in the CCW system are sufficient for preventing overpressure of protected components. Assumption 3 indicates that the containment CCW drain tank will become an atmospheric tank. This assumption required confirmation by removing valve 1-RV4724 from the system. DCA 60658/0 provides confirmation of removing valve 1-RV4724. This modification provides an open vent for the drain tank and eliminates the possibility of tank overpressure. The calculation also required verification of field run relief valve discharge piping. This was accomplished and documented on various drawings; e.g., BRP-CC-1-RB-003, Revision CP-1. Finally, Revision 2 of the calculation resolved a confirmation item by updating and correcting maximum backpressure values using new friction loss data from calculation 16345-ME(B)-093. The NRC team agreed that the confirmation resolutions were appropriat _ _ _ - - _ - _ _ _ - _ _ - _ _ _ _ _ _ _ - - _ - _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ . . _ _ _ _ - _ _ _ _ _ _ _ _ - - - - _ _ _ _ - _ . _ _ _ _ - _ _ _ _ _ -

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The,following are EFE open items reviewed by the NRC team in order to' evaluate the resolution proces EFE Item P02 DBD-ME-003, Revision 0-A, June 24, 1987, which addresses control room habitability, does not address the FSAR commitment regarding analysis to ensure that steam from a postulated steam generator blowdown piping break in the Electrical and Control building does not enter the Control room. This action item was closed since the.DBD (Revision 2) now addresses the impact of a .

blowdown' piping break. ' Implementation of this' analysis is Ebasco's responsibility as indicated in letter SWE-156 dated July 27, 1987. ~ The NRC team agrees that the EFE action ite was. adequately resolve EFE Item PO4 Action item PO4 was written against Revision 0 of DBD-ME-232 due to_the lack of adequate design criteria associated with the following concerns: Leak detection and collection requirements for CT  ;

(Containment Spray) syste l System design provisions for the single passive failure of a CT system componen Adequate separation of redundant trains for flooding due to equipment leakag EFE closed this item based upon-the following project actions: Revision 3 of.DBD-ME-232 commits to design the CT system in accordance with the requirements of GDC-54 and 56, which cover the requirements to provide leak detection, isolation, and containment capabilities for piping systems which penetrate reactor containment. The specific requirements and criteria to meet this commitment for the CT system are provided'in Section 5.1 of DBD-ME-013.

i, Revision 3 of DBD-ME-232 requires the CT system to meet the' redundancy and power source requirements of ESF systems including the design requirement to accommodate a single passive failure in the long-term during the recirculation phas DBD-ME-007, Revision 2 and EME 2.24-02, Revision 2, adequately address normal equipment leakage among those ~

source events used to determine the transient flood conditions in performing flooding analyse I

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The NRC reviewed the above referenced DBDs and agreed that the project actions were responsive to the action ite EFE Item P14 G&H calculation 2323-232-3, Revision 1, which was validated by SWEC, had an error in calculating a head loss for header No. 1, nozzles 12-15. The project stated that the head loss error was insignificant to the pressure drop across the spray nozzles and the orifice sizing. The resolution also involved a review of calculations and their associated validation records to confirm that the status was correct regarding the superseding and supplementing of G&H calculations. The NRC team agreed with EFE action to resolve the action ite EFE Item P19 The purpose of Calculation 16345-ME(B)-169 is to determine CT system flow rates for the injection and recirculation modes, and also to calculate the net positive section head (NPSH)

available to the CT pumps during these two mode Calculation 169 did not address a steam line break or feedwater line break inside containmen The response evaluation indicated that a further review of the FSAR accident analyses and the emergency operating procedures

.shows that for a main steam line break (MSLB) or feedwater line break (FWLB) there would not be a recirculation mode of the CT system taking suction from the containment sump. The NRC team concurred in EFE's action to resolve this item on this basi EFE Item P23 This item documented various concerns with calculation 16345-ME(B)-057, Revision 0, "CT Fluid Dynamics During Startup." Calculation 16345-ME(B)-276, Revision 0, superseded -057 and was determined by EFE to address the original concerns regarding pump runout, pressure drop across the nozzles during priming, spray header elevation variations, drop in refueling water storage tank (RWST) level following a small break loss of coolant accident (LOCA), etc. The NRC team agreed with EFE's basis to close out this ite EFE Item P49 Calculation 16345-ME(B)-124, Revision 0, calculated a minimum time of 15.2 minutes (based on available water in the RWST) to switch over the CT system from injection to recirculatio This was at variance with NUREG-0797, Supplement 3, page 6-1 (SSER 3) which accepted a minimum switch-over time of a little over 18 minutes as a sufficient time for the operator to perform manual safety actions following a LOCA. EFE closed

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this-item on the basis of confirming that the project pointed out the inconsistency to TU Electric in DVP-1-llI-R-1532. The NRC inspector determined from TU. Electric that calculation-124 has been superseded by calculation 16345-ME(B)-389 which determined that the switch-over time is approximately 18.3 minutes. Based on the superseding calculation, the NRC team concludes that the inconsistency noted by EFE has been adequately resolve In summary, the NRC team's sample review of mechanical calculation confirmation items and EFE action items led to a conclusion that actions to resolve both of these were effectiv . Systems Interaction (50090, 48051, 48055, and 37051)

Open Item 27*

• An open item is any identified item or issue required to be resolved prior to closure of the System Interaction Progra Calculation CPE-DS-CA-0000-640, Revision 0, determines zones of interaction in the Containment building for nonseismic components whose failure during a seismic event could lead to damage to a Category I system or component. Documentation was required to prove the validity of references related to floor response spectra. Revision 2 to the calculation stated that amplified response spectra provided by DCA-72,518 enveloped all refined response spectra in the above references. The NRC team determined that this statement was based upon alternate calculations performed by SWEC/IMPELL which validated the original G&H spectra, except for the service water intake structure and Category I tanks. SWEC provided new spectra to Ebasco/ Systems interaction for the latter structure The NRC team agreed that this item was properly confirme Open Item 82 Calculation CPE-SI-CA-0000-732 calculates the gross volumes of  ;

the various rooms / areas in Safeguards building No. I for use in I environmental analyses. The calculation states in its assumptions section that DCAs against structural drawings will not be reviewed and will be identified as an open ite Calculation CPE-SI-CA-0000-764 calculated gross volumes in Safeguard building No. 1 based on walkdown data and compared these volumes with those calculated in calculation-732. The differences for areas / rooms were less than 10%, which was judged to have a negligible impact on the results of the environmental analysis. The NRC team agrees with the method of  ;

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Open Item 170 Calculation CPE-SI-CA-0000-730 determines the worst flooding discharges that occur during normal operation from a single high energy line break in each area of the safeguards buildin The open item indicates that the maximum mass flow rate used in the calculation has to be confirmed. The confirmation process involved ensuring that the breaks providing the greatest flow assumed in particular areas were valid.- Reconciliation involved making revisions where final breaks had greater flow than that assumed. The NRC temn agreed with this proces Open Item 26 Calculation CPE-DS-CA-0000-610 defines the pipe movement and pipe whip restraint loading and acceptability resulting from postulated high energy pipe ruptures for residual heat removal (RHR) lines inside the containment (problems 1-26A and B). The open item indicates that the original calculation of the hinge point location is to be superseded by the method described in DBD ME-007. The hinge point location has been calculated in a new calculation, CPE-SI-CA-0000-733, Revision 1, in accordance '

with DBD ME-007. The NRC team considers that this action resolves the open ite Open Item 171 Calculation CPE-SI-CA-0000-699 determines the pressure and temperature transients in the various areas of the electrical 4

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and control building following a high energy line break in room 113. Ebasco needed to satisfy environmental qualification parameters while simultaneously minimizing hardware changes; e.g., sealing doors, introducing isolation dampers in HVAC ,

I ducts, and changing door positions (open or closed). Open item 171 concerned assumptions related to these potential changes. Subsequently, Ebasco revised its approach and superseded this calculation. This new approach uses HVAC isolation dampers introduced by design modification request / construction 89-1-013 and changes introduced by DCA 67131; e.g., 100 ft. roll-up door leading to the turbine building is modified to blow out at .2 psi differential and doors to stairwell EC-2 are sealed. The NRC team agrees with closure of open item 171 based on the above action Open Item 257 Calculation CPE-SI-CA-0000-733 determines the locations of plastic hinges along the RHR lines inside containment due to postulated pipe breaks 12-RC-1-007-2501R-1 and 12RC-1-069-2501R-1. The open item concerns confirmation of the forcing function coefficient used in the calculation;, i.e., a supporting calculation file for the coefficient was not i

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available. Revision 1 to calculation 733 calculated the

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forcing function. The NRC team considers that this resolved l the open ite Open Item 259 Calculation CPE-SI-CA-0000-737 is a pipe rupture analysis for a steam ganerator blowdown line (restrained). It is categorized as Stress Problems 1-79E and F and determines pipe end movements and the acceptability of pipe whip restraints. Open item 257 for Revision 0 of the calculation concerns the need to refine and rerun the RELAP model due to excessive strain in U-Bar restraint SB-1-60-903-D37W. The NRC team reviewed Revision 2 to the calculation which reflected revised calculations and indicated that calculated strain was within allowables for the restraint. Therefore, the open item resolution is acceptabl Open Item 329 The open item was intended to provide design validation of attributes in the high energy line break source lis Calculation CPE-SI-CA-0000-914, Revision 0, identified jet types and obtained stagnation pressures and temperatures governing the break flows for all high energy line breaks on Unit 1 and Common. The NRC team determined that the latest version of the source list, Revision 8, reflected an interface with CPE-SI-CA-0000-914, Revision 2. The team considers that the open item was correctly resolve Open Item 336 Calculations CPE-SI-CA-0000-662, 663, 664, and 693 are building flooding calculations and all make the assumption that leakage from threaded piping fittings is enveloped by moderate energy line break events on the same size piping. The open item required confirmation of this assumption. The open item was resolved by IMPELL report 11-0210-0007, Revision 0, " Fire Protection Threaded Pipe Failure Evaluation Report," which stated that the quantities of water involved were "far short of the amount to constitute an internal flooding hazard." The NRC team considers that this action resolved the concern with leakage from threaded fittings as a potential input to the flooding analysi . Civil / Structural (45051, 46055, 46071, 48051, 48055, 37055, and 37051)

Steam Generator Upper Lateral Restraint NUREG-0797 Supplement No. 17, Section 4.1.2.3, documents the staff's review of design validation for the steam generator

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upper lateral restraint beam. On page 4-20, the discussion of structural loading states that the calculation books indicated that the seismic' loads and pipe reactions under thermal loads generated by postulated breaks require confirmation. The loads were being finalized by Westinghouse taking into account the revised support stiffnesses provided by SWEC and could not be reviewed by the staff during the time frame of the above revie During the current inspection, the NRC team followed up on the load confirmation matter. The team reviewed SWEC letter SWW-0148 dated June 6, 1988, to Westinghouse which transmitted revised information concerning the local stiffness / flexibility values of the supporting concrete structure at reactor coolant system equipment support locations. SWEC letter SWW-0298 dated October 13, 1988, transmitted the asymmetric pressure loads acting on the steam generator. Westinghouse letter WPT-11556 dated July 7, 1989, to SWEC details the loads imposed on the building structure by the reactor vessel, reactor coolant pump, and steam generator support The Westinghouse letter incorporates a SWEC revision (July 11, 1989) in subcompartment pressurization which was necessitated by an error identified by TAP audit ATP-89-144. The error involved calculation 16345-EM(B)-091, Revision 1, in which pressures on the lower portions of the steam generator were applied at one elevation segment higher than they should have bee The NRC team reviewed SWEC calculations 16345-CS(B)-036, 121, and 262 which incorporate the most current structural interface loads for NSSS components. The SWEC calculations apply to the I steam generator compartment concrete structures, the upper I

lateral restraint beams and the concrete embedments, respectively. Calculation-036 concluded that the steam generator compartments are adequate to withstand the design environment as given in DBD-CS-083. Calculation-121 concluded that the upper lateral support beam structural system was adequate to sustain all design loads. The box girder and end beam connection were shown to meet the requirements of l

DBD CS-08 Calculation 262 concluded that the embedment l anchorages and concrete corbels of the upper lateral support beams are structurally adequate to withstand all design loads and loading combinations as specified in DBD CS-08 In addition, the NRC team reviewed change notices to the above l calculations which factored in revised Westinghouse loads resulting from the above TAP identified error. In each case, the change notice documented that the structures were acceptabl In summary, the NRC team determined that, for the steam generator compartments, the various Comanche Peak design organizations interfaced in an acceptable manner to reconcile f

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seismic loads and pipe reactions under thermal loads generated by postulated break Design Validation Package 09A, " Containment Concrete" The NRC inspector reviewed the confirmations required for calculation CSC-131, Revision 2, " Analysis and Design Check of the Equipment Hatch Area of the Containment Building, Unit 1."

Confirmations required were (1) the design pressure within the containment generated by the design basis accident, (2) temperatures inside and outside the containment used for load combinations, (3) temperature distribution in the steel barrel used for the equipment hatch, (4) seismic acceleration input values, and (5) various input forces, stresses, et Items 1 and 2 above were confirmed by reference to Design Basis Document (DBD)-CS-073, Revision 2, " Concrete Containment Structure." Item 3 was confirmed by Calculation Change Notice (CCN)-002 which provides an alternate calculation of the temperature gradients and barrel stresses originally computed by G& Item 4 was confirmed by reference to DBD-CS-081, Revision 4, " General Structural Design Criteria." Item 5 was superseded by calculations and data provided in Revision 2 of calculation CSC-131 with no confirmations require The NRC inspector reviewed DBD-CS-073 and DBD-CS-081, and verified that items 1, 2 and 4 were adequately addressed and confirmed. The NRC inspector reviewed the calculation change provided in CCN-002 and verified that item 3 was confirmed by the alternate calculation. In addition, the NRC inspector verified that Revision 2 of calculation CSC-131 supersedes item 5 above. The NRC inspector concluded that the required confirmations were correctly closed for calculation CSC-13 Desien Validation Package 09C, " Fuel Building" The NRC inspector reviewed the confirmation required for calculation CSB-207, Revision 0, " Fuel Building Beam B-24 Analysis." One confirmation was required regarding a seismic acceleration value used in the analysis. This acceleration value was taken from G&H calculation DFB-1C, Revision 0, which required confirmation. DBD-CS-081, Revision 4, provides a lower acceleration value. The applicant closed the required confirmation in an interoffice correspondence by stating that the beam was qualified in calculation CSB-207 using a higher acceleration value than presently required by DBD-CS-08 The NRC inspector reviewed calculation CSB-207 nnd verified that the seismic acceleration value used was greater than that required by DBD-CS-081. The NRC inspector concluded that the required confirmation was adequately close . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _

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Design Validation' Package 09D, " Safeguards Building" The NRC inspector reviewed the confirmations required for calculation CSC-074, Revision 2, " Floor Slab Design EL 810'6",

safeguards Building." Confirmations required were (1) the adequacy of the floor slab using three load combinations not considered-by calculation CSC-074, (2) several equipment loads, (3) review of vendor shop drawings gpecifying rebar installation, and (4) an assumed 50 temperature differentia These required confirmations were closed in change notice CCN-002 to calculation CSC-07 The three load combinations were addressed in CCN-002 with the exception of four terms: (1) pipe reactions under thermal conditions due to the postulated pipe break (R ),

(2) equivalent static load on the structure du8 to reaction on the broken high-energy pipe (Y , (3) jet impingement on the structure due to the postulated) break (Y ), and (4) missile impactonthestructureduetothepostukatedbreak(Y).

CCN-002 states that the effect of R , Y Y and Y 03 the Safeguardsbuildingwillbeaddress8dby,cakc,ulatioECSC-145,

" Evaluation of Y , Y , Y , R,T PWR, M Safeguards Building," sched61ed)for*comhletfo,n by laEe, September 198 In addition, the applicant has committed (Commitment Data Form Register 18222) to provide final design jet impingement loads and inputs.for evaluating structures for impacts due to pipe breaks in a revision to DBD-CS-081, " General Structural Design Criteria."

The applicant confirmed the equipment loads and rebar installation by reference to vendor documents and by conservative evaluations in ghange notice CCN-002 to calculation CSC-074. The 50 temperature differential was confirmed by reference to DBD-CS-081, Revision 4. The maximum temperature differential of 52.9 F was between LOCA accident temperature and ambient const5u ti n temperature. CCN-002 determined that the use of 50 F as the temperature differential was confirmed and that the increase of 2.9 F does not effect the results of calculation CSC-07 The NRC inspector reviewed the above documentation and concluded that the required confirmations were correctly and adequately documented. In addition, the NRC inspector concluded that the effect of the terms not yet evaluated for the load combinations is adequately tracke The NRC inspector reviewed the confirmations required for calculation CSS-143, Revision 1, " Evaluate the Effect of Postulated Rebar Cuts in Safeguards Building as Defined in NCRs 87-5982, 87-5983, 87-5984, and 87-5986." Confirmations required were (1) confirmation of G&H calculations SSB-ll8c,

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set 4; SMI-105C, set 3; and SSB-106C3, set 4; and (2) confirmation of a load tabl Calculation change notice CCN-001 to CSS-143 states that the data used from G&H calculations SSB-106C3, set 4, and SMI-105C, set 3 (required reinforcement steel area) in calculation CSS-143 are not required as the rebar cuts specifically addressed are analyzed in SWEC calculations CSC-074 and (

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CSC-082. Since the latter two calculations will provide the qualification of the rebar cuts, these portions of calculation CSS-143 are no longer required and the required confirmation is close Change notice CCN-001 to calculation CSS-143 states that the safeguards building secondary walls (G&H calculation SSB-118C, set 4, also addresses these walls) have been analyzed in SWEC calculation CSC-090 for all but three load combination CCN-001 provides confirmation of the remaining three load combinations and thus removes the required confirmation of G&H calculation SSB-118C, set The load table was confirmed as all required load combinations specified in the DBD-CS-084 are addressed in CCN-001. The G&H calculations were validated or voided with SWEC calculations referenced to provide confirmation of calculation CSS-14 The NRC inspector reviewed the above documentation and concluded that the required confirmations and new calculations were correct and adequate to close the confirmation Design Validation Package 09K, " Containment Liner and Penetrations" The NRC inspector reviewed the confirmation required for calculation CSB-026, Revision 2, " Reactor Containment Building Liner Insert Plates." Confirmations required were (1) loads for analyzing the rotating platform and (2) attachment loads from structural attachment loading schedules. These required confirmations were closed in an interoffice correspondence dated June 29, 1989, and included with the calculatio The rotating platform loads were provided in confirmed calculations CSB-008, Revision 3, and CSB-027, Revision O. The containment liner insert plate for the rotating platform support was reanalyzed in Revision 2 of calculation CSB-026 thus providing the required confirmatio Calculation CSB-026 provides lists of acceptable loads for a i l

variety of containment liner insert plates. Individual liner plates with attachments are qualified by comparing the specific loads for each plate with the generically allowable loads provided in calculation CSB-02 If individual loads exceed I

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the loads listed in calculation CSB-026, they will be resolved !

on a' case by case basis either by refining the loading .

estimates or by' analyzing the specific plate / loading situatio i

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The required confirmation for specific attachment loads was unnecessary as calculation CSB-026 only provides generic loading qualification The NRC inspector reviewed the above documentation and discussed the use of the generic loads qualified in calculation CSB-026 with the applicant. The NRC inspector concluded that the required confirmations were correctly and adequately close Design Validation Package 09T, " Seismic Analysis Verification" The NRC inspector reviewed the confirmations required for calculation CSC-113, Revision 0, " Validation of Acceleration Profiles." One confirmation was required regarding a referenced calculation which had not been signed and issue Change notice CCN-001 for calculation CSC-113 removed the confirmation requirement and states that the referenced calculation was filed (signed off and issued).

The NRC inspector requested additional information from the applicant regarding the extent of review required for CCN-001 for calculation CSC-113. CCN-001 was signed off by the same reviewer as the independent reviewer. This is allowed by Procedure STP 11.5-2. Procedure PP-009 states, in part, "If confirmed inputs and/or assumptions do not adversely affect the calculation results then a calculation revision is not needed."

A revision of calculation CSC-113 was not generated due to CCN-001. In addition, the applicant stated that a review of the effect of any changes to the referenced calculation on the conclusions reached in calculation CSC-113 was completed for CCN-001. The NRC inspector concluded that CCN-001 to calculation CSC-ll3 adequately satisfied the required confirmatio The NRC inspector reviewed the confirmations required for calculation CSC-007, Revision 0, " Development of Dynamic Model for Seismic Profiles for the Reactor Building." Confirmations required were (1) the use of the Comanche Peak FSAR as a reference, (2) the weight of the polar crane as specified in the FSAR, (3) specifications for a roll-away missile shield as specified in the FSAR, and (4) damping values as specified in the FSA Change notice CCN-01 for calculation CSC-007 removed the confirmation requirements by reference to a vendor arawing identifying the weight of the polar crane and by stating that information in the FSAR does not require confirmatio The NRC inspector discussed CCN-01 to calculation CSC-007 with the engineer who reviewed and signed CCN-01. The engineer

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stated that calculation CSC-007 was issued prior to final acceptance ofLthe Comanche Peak FSAR by SWE A confirmation of information taken from the'FSAR'was'actually unnecessary but was identified conservatively by the engineer who prepared the i calculation. The current FSAR for the date CCN-01 was issued was. compared by the applicant with the FSAR used for" calculation CSC-007. No changes were identified and the 'I required confirmations were removed. The NRC inspector l

. concluded that CCN-01 to calculation CSC-007 adequately l satisfied the required confirmation . HVAC Systems (50100, 37051, 35061, and 50075) j i

The NRC team evaluated the consistency between DBDs and  ;

implementing calculations for three HVAC systems: control room )

air-conditioning,. diesel generator air ventilation, and uninterruptible power supply area air conditioning. Such consistency-is indicative of the reconciliation process, and the NRC team focused.on DBD parameters that affect calculation Control Room Air-Conditioning System (CRACS)

The NRC team-reviewed DBD ME-304, Revision 2, " Control Room Air Conditioning System," calculation X-EB-304-1, Revision 5,

" Control Room Space' Heat Gains and Maximum Space Temperature" I and FSAR Figure 9.4-1, " Flow Diagram Control Room Air  ;

Conditioning." The review addressed functiongl requirements to i l

maintain the Cgntrol Room complex (CRC) at 75 Fi 5 F Dry Bulb (DB) and 35-50 relative humidity during all plant condition The mechanical equipgent rooms serving the CRC arg required to be maintained at 104 F DB maximum-(summer) and 40 F DD minimum (winter) during all plant condition The upper limit of the i design temperature in areas located.within the CRC, but which l are not essential to habitability or which do not congain I safety related equipment (e.g., kitchen), shall be 80 F, and  !

under maximum cooling load conditions temperatures in such areas may be permitted to exceed slightly the 80 F limi The aboveindoorenvironmentalconditgonsshaglbemaintainedunder outdoor extremegconditions of 110 F DB/80 F Wet Bulb (WB)

(summer) and 20 F DB (winter).

During the normal operation mode, two of the four air-conditioning units, 25,845 cubic feet per minute (CFM) l capacity each (total airflow capacity of'51,690 CFM) are required to be in operation. One of the two makeup air supply fans introduces 3,000 CFM of outdoor air for ventilation which mixes with 48,690 CFM of return air. The air quantity of 51,690 CFM is distributed through duct work to the various spaces. One of two 1,250 CFM capacity kitchen and toilet exhaust fans ventilates the kitchen, toilet, janitor closet and locker room. One of two 1,750 CFM capacity control room

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exhaust fans exhausts between 950 and 1,750 CFM of air from the H CR The motorized damper at the intake of the control room exhaust fan is modulated to achieve this variation. The' makeup air flow rate of 3,000 CFM exceeds the total mechanical exhaust from the CRC-(1250 CFM plus 950 CFM) by 800 CFM maximum in order to maintain the CRC at a slightly positive pressure of 0.125 inch Water Gauge (WG) with respect to its surrounding During the emergency recirculation mode, the operating makeup air supply fans, the operating control room exhaust fan and the kitchen / toilet exhaust fan all shut dow One of two emergency pressurization units draws a maximum of 800 CFM of outdoor air to maintain the above positive pressur The NRC team determined that'the capacities for major equipment and components described in Section 10 of the DBD, based upon equipment specifications, vendor drawings, and instruction manuals, are consistent with the above functional requirement In addition, the NRC-team determined that the above DBD functional. requirements were consistently and correctly implemented in calculation X-EB-304- Diesel Generator Area Ventilation System The NRC team reviewed.DBD ME-302A, Revision 2, " Diesel Generator Area Ventilation System," and calculation 1-EB-302A-1, Revision 4, "As-Built HVAC Calculation Diesel Generator Area - Space Heat Gains, Space Heat Losses and Maximum and Minimum Temperatures - Unit 1." The review addressed functional requirement During normal plant operation, the diesel generator area ventilation systemgis required to maintain an igdoor temperature of 122 F DB maximum (summer) and40gDBmingmum (winter) under gutdoor extreme conditions of 110 F DB/80 F WB (summer) and 20 F DB (winter). During emergency operation, as well as when testing the diesel generator dgring normal operation, maximum roomgtemperatures of 129 F DB (HVAC equipment room) and 122 F DB (all other rooms) shall not be exceeded under the above summer extreme conditions. During emgrgency operation (non-lE heaters unavailable) in winter at 20 F DB outdoor temperature the diesel generatgr, day tgnk and HVAC eguipment rooms shall be maintained at 40 F DB, 20 F DB, and 30 F DB, respectivel The DBD system description and operation section indicated that the diesel generator room ventilation subsystem diesel generator area ventilation system consists of two independent trains, each containing four 25 percent capacity, 50,000 CFM each exhaust fans. The day tank room ventilation subsystem consists of one independent train for each of the two diesel {

generator areas (DGA). Each train contains one 100 percent capacity 3,350 CFM exhaust fan exhausting 400 CFM from the day l

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tank room and 2950 CFM from the ventilation equipment roo 'Each DGA is provided with a 5 kw capacity electric unit heate These permanent unit heaters are supplemented by portable unit heater During normal plant conditions in summer when the diesel generator is not operating, only one of the diesel generator l

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area ventilation system 50,000 CFM exhaust fans is required to be in operation. During summer testing of each diesel generator, all four fans per train start automatically with the diesel generator start signal. Outdoor air quantity of 200,000 CFM is drawn in and exhausted. For the day tank room ventilation subsystem, during normal operation (summer and winter) both fans (one per train) are required to be in operation. outdoor air quantity of 3350 CFM is drawn in and 400 CFM and 2950 CFM are exhausted via the day tank room and ventilation equipment room, respectivel The NRC team reviewed calculation lEB-302-A-1 and determined that the above functional and system operation requirements-were implemented in a manner consistent with the DB Uninterruptible Power Supply Area Air-conditioning System The NRC team reviewed DBD-ME-313, Revision 1, "Uninterruptible Power Supply (UPS) Area Air Conditioning System," and calculation X-EB-313-1, "UPS Area Space Heat Gains and Maximum Temperatures," and focused upon the functional requirements and system operatio The uninterruptible power supply area air-conditioning system is reggired to maintain an indoor temperature within a maximum of 104 F in both the UPS distribution rooms and AC equipment room during normal plant operation. During either upset, emergencyorfaultedmodesofoperatgon,indoortemperatures are required to be maintained at 122 F maximum. The uninterruptible power supply area air-conditioning system is required to remove the following quantities of heat (in BTUH),

in addition to that generated from fan motors:

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Number of Emergency Room- Rooms Normal > Mode Mode

'UPS Dist. Rm.- 2.(1 per 53,945 each .53,945 each

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. Train: A unit)

UPS Dist. Rm.- 2 (1 per 64,630 each 64,630 each Train B unit)

AC Equip. 2 ,565 17,530 N AC Equip. ; R ,485* 1020*

N * Room sensible heat gain when equipment in the room is on the standby mod During all modes of plant operation, an AC unit (Train.A or B)

is in operation. The total supply and return system air flow-

-rate is 11,200 CF The NRC team reviewed calculation X-EB-313-1, Revision 3, and determined that the above functional and system operation requirements were implemented in a manner consistent with the DB . Heating, Ventilation, and Air-conditioning (50100, 37051, 35061, and 50075)

For final reconciliation in this area, Ebasco developed Procedure SAG CP-40, Revision 0, "HVAC Structural Design Reconciliation." This procedure provides the guidelines for design-reconciliation and applies these guidelines to the as-built hardware installed in the field using appropriate analysis methods. This analysis results in either acceptance of the hardware or identification of required modification The validation program was designed in a manner that allowed a preliminary design validation to be completed prior to required field rework or QC verification of as-built data. A final reconciliation was then required to assure that the final installed-QC verified hardware was in accordance with design criteria and properly considered in the final calculation Under SAG-40, a technical review will be performed and identified as " Screen 1." Depending on the outcome of this technical review, a second, more in depth technical review termed " screen 2" may be performe _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ .

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l The technical review for Screen 1 consists of an analysis of: Technical open items that are a result of audits, criteria changes, or other items which impact the calculatio System impact items which are a result of system hardware or configuration changes which impact the calculation.

l L Interaction ratios (IR). Interaction ratios are used as an indication of whether the impact of the change on the analysis systems (strings) should be considere These consisted of reviews for screen 1 and screen 2 consisting of one or more of the following:

. Hangers anchored to walls for correct orientatio . Anchor bolts not Code checked by STRUDL (structural analysis program) including verification of algebraic sign and magnitude of force . Member bridge welds (weld gaps).

. Changes in inaccessible attributes worst case assumption . Changes in Richmond insert allowable . Correctness of algebraic sign and magnitude of design forces for welds with interaction ratios greater than .8 . Addition or replacement of volume damper, isolation damper, and gravity dampe . Deleted /added support . Addition / replacement / relocation of flex connectio . Changes in duct insulation type . Addition / replacement of fire damper . Changes in duct construction typ . Determining the highest member interaction ratio (IR)

from STRUDL Code . Determining the highest anchor bolt IR from the STRUDL Code . Determining the highest weld IR for the hange __- _ _ - _ _ _ - _ _ _ _ _ _ _ - - _ _ _ _ - -

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. Determining the highest duct I . Determining the highest duct to hanger I Under screen 2, the impact of specific technical concerns relative to any given calculation is evaluated and a determination made as to whether an " abbreviated reconciliation" or a " full reconciliation" will be performe The major consideration for determining which reconciliation will be performed is the IR. If the IR is > .85, a full reconciliation will be performed; if the IR is < -85, . an abbreviated reconciliation may be performed. The material presented to the NRC inspector identified a total population of 176 strings with 1140 supports. A total of 22 strings and 127 supports.have been completed. The four strings identified below contained a total of 60 IR points with 14 irs above .8 The NRC inspector reviewed all of the irs and related calculations and agreed with the conclusions reached by the TU Electric engineer The NRC inspector reviewed the packages, NCRs, DCAs and calculations associated with the following string String N No. of Supports No. of NCRs No. of DCAs 31 6 4 6 32 5 2 11 46Y 5 1 12 46Z 4 1 11 The NRC inspector identified no assumptions which required confirmation. Documentation reviews were performed to ensure that work required by DCAs had in fact been implemented as required and that all necessary QC inspections were satisfactorily completed. NCRs were checked for completeness and correct QC sign-offs. Calculations relevant to the DCAs, NCRs, etc., were checked for accuracy and methodology. Based on the satisfactory review of the above actions, the NRC inspector determined that the final reconciliation process was properly implemented for the four HVAC strings inspecte . _ Instrumentation and Control (52055, 37055, 51055, 51065, 51061)

l The following are calculations that the NRC team reviewed to assess resolution of confirmatory items.

l Calculation IC(B)-002 The calculation determines the required air accumulator capacity for the auxiliary feedwater valves and the auxiliary feedwater recirculation valves. An input used for this calculation and requiring confirmation was that air

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.25 accumulators require sufficient capacity.to either regulate flow or isolate-a steam generator.within a period of 30 minutes or 5' operations after loss of instrument air. Revision 3 for-this calculation removed'the confirmation required by indicating that the related confirmed design basis document requires air accumulators with sufficient. capacity to either regulate flow or isolate a steam generator within a period of 30 minutes or 5 operations after loss.of-instrument air. The NRC-team concludes that this input for the calculation 11s consistent with the confirmed system requirement and, as such, the item has been adequately confirme Calculation IC(B)-004 The calculation determines the adjusted setpoint and reset point for pressure bistable switch 1-PB-2453. When low pressure is detected, the pressure. bistable' switch sends a signal to the discharge valve logic that transfers the valve control system into the automatic mode if an automatic start signal is present for the motor driven auxiliary feedwater pumps. The normal process operating range of greater than 1472 psig required confirmation. Sufficient margin (to prevent spurious system operation) between the normal operating pressure and the calculated setpoint value of 1449 psig also required confirmation, Revision 1 of this calculation removed the confirmation required items by. noting that the normal operating pressure of 1438 psig was obtained from mechanical calculation 16345-ME(B)-053, Revision 1, " Auxiliary Feedwater System Performance." The revised calculation also indicated that it is acceptable for the setpoint to be higher than the normal operating pressure since (1) during automatic start conditions the control valves would transfer to automatic as required and (2) during normal plant conditions (no automatic-start signal) the control valves can be placed in manual control.

L The inspector found from damage as a result of a runout

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condition that while the calculation adequately protects the pump, it does not adequately cover operational consideration For example, with the set points adjusted as described, the discharge valve would be transferred to automatic control whenever an automatic start signal occurred. The operators would, therefore, not be able to take manual control of the valve to. control steam generator level and cool-down rate. The inspector found that Revision 2 of this calculation changed the set point.to 1210 psig which allows the operator to regain manual control of the discharge value. The inspector is concerned, however, that these operational considerations weren't considered.in the process of reviewing and approving Revision 1 of this calculation. This item is considered open pending applicant demonstration that operational considerations have been considered in all calculations (445/8961-0-01).

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Calculation IC(B)-008

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The calculation determines the adjusted setpoint and reset point for flow bistable switches 1-FB-2456B1 and 1-FB-2456B When high flow rates are detected in the discharge pipe for the auxiliary feedwater pump, comparators B1 and B2 send signals to the pump discharge. control valve logic that will automatically l limit the pump flow to a value that will prevent pump runou The normal process operating range of less than 525 gpm required confirmation. Confirmation that the calculated setpoint values of 544.36 gpm and 589.46 gpm have sufficient margin from the normal operating region of 525 gpm to prevent spurious system operation was also required. Revision 1 of this calculation removed tne confirmation required items by noting worst case conditions, resulting from this calculation, may result in the flow bistable being unable to reset after actuation. This condition is acceptable because the actuation of the bistable at high flow coincident with low pressure will transfer the system to automatic pressure control which.is the safe condition for auxiliary feedwater pump runout protectio The inspector found that while the calculation adequately l protects the pump from damage as a result of a runout condition, it does not adequately cover operational considerations. For example, with the setpoints adjusted as described, once the discharge valve is transferred to automatic control as a result of an actual or transient condition, the l bistable would not reset. The operators would, therefore, not be able to take manual control of the valve to control steam generator level and cool-down rate. The inspector found that Revision 2 of this calculation changed the setpoint to 800 gpm which allows the bistable to roset and allows the operator to regain manual control of the discharge valve. This is an additional example of open item 445/8961-0-0 Calculation IC(B)-019 The calculation determines the adjusted setpoint and reset point for flow bistable switch 1-FB-4773-1A. A flow transmitter senses pump discharge flow and sends a signal by way of the loop power supply and square root extractor to the signal comparator. When low flow is detected for the attendant containment spray pump, the comparator sends a signal to open the recirculation line flow valve for that pump. Transmitter uncertainty calculations due to effects of temperature and total integrated dose at the transmitter locatgon were performed using values of 120 F and 1.61 x 10 Rads. These numerical values required confirmation. Revision 2 of this calculation revised the transmitter uncertainty due to temperature effects because it was determined from environmental drawing MI-3000, Sheet 02, Revision CP-16 that the maximum temperature at the transmitter location is 122 _ _ - _ _ _ _ - _ - - - _ - - - _ - - _ _ _ _ _ - _ _ _ _ _ _ - _ _ - _ _ - _ _ _ _ . _ _ _ - - _ __- _ _ _ _ _ - _ _ .

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27 Revision 2 of_the calculation also revised the radiation

. exposure effect'since calculation NU(S)-160 determined that the total intggrated dose at the transmitter location was 2.06 x 10 Rads. The NRC team determined that the identified items requiring confirmation were adequately considered and appropriately revised and incorporated into the subsequent calculatio Calculation IC(B)-060 The calculation determines the adjusted setpoint and reset point for flow bistable switch 1-FB-2457A. The flow transmitter senses motor driven auxiliary feedwater pump 02 discharge flow and sends a signal by way of the loop power supply and square root extractor to the signal comparato When low flow is detected, the comparator. sends a signal to open a valve in the recirculation line for the pump. Numerical values for temperature and total integrated dose at the transmitter location (used.to calculate related uncertainties)

required confirmation. The NRC team verified that revised numerical values for temperature and total integrated dose were obtained from appropriate sources and correctly incorporated into Revision 1 for the calculation. The NRC team finds that the confirmation items have been adequately addresse Calculation IC(B)-016 The calculation determines the adjusted setpoint and reset point for component cooling water surge tank level bistable The calculation addresses two channels used to monitor and control the hi, lo-lo, and empty levels in the component cooling water surge tank. Numerical values used in the calculation for the normal process operating range (obtainable from the related mechanical calculation), temperature and total integrated dosage at the transmitter location, and the calibrated range of the transmitter required confirmatio Revision 2 of the calculation used the numerical values for the normal process operating range from mechanical calculation ME(B)-073. Numerical values for the remaining three items were obtained from appropriate reference sources and correctly incorporated into Revision 2 of the calculation. The NRC team agrees that the items requiring confirmation have been appropriately and adequately addresse Calculation IC(B)-065 The calculation established the adjusted setpoint and reset i point for pressure switch 1-PS-4518. This pressure switch  !

senses the Train B component cooling water pump discharge '

I pressure and when low pressure is detected the switch sends a signal to the Train A component cooling water pump start logi .

Revision 1 for the calculation noted that the process safety

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s limit off50 psig for the switch required confirmation <

Validated. mechanical calculation'ME(B)-311 clearly determined the process safety limit to be 41 psig. Revision for the'

calculation appropriately incorporated the revised process limit and correctly revised the calculation. The NRC team agrees that the confirmation item has been adequately-addressed, Calculation IC(B)-106 The calculation determines the setpoint and-reset point for flow bistable switches. Flow transmitters sense component cooling water discharge flow at the reactor coolant pump thermal barrier coolers and send their signals by way of a loop power supply and square root extractor to signal comparator When high flow is detected in any loop, the corresponding comparator sends a signal-to automatically stop the flow from all thermal barriers by closing valves. Numerical values for the process limit and normal ~ process operating range, maximum expected component cooling water temperature after a reactor coolant pump thermal barrier tube failure, flow' orifice data, transmitter seismic error, total integrated dosage at transmitter location,.and expected error from cable leakage due-to' severe environment required confirmation. The team determined that numerical values for each of these items were obtained from appropriate reference documents and adequately incorporated into Revision 1 of the calculation. The NRC team agrees'that the items requiring confirmation have been appropriately considered and adequately addresse The following are EFE concerns that the NRC team reviewed to assers their resolutio Action Item No. CO2 Design Basis Document (DBD)-EE-037, Section 5.1, states "The instruments are assumed to be calibgated in situ with an ambient temperature range of 40-10g F for field-mounted outsidg containment equipment, 122 F for containment equipment and 75 F for control room equipment. The setpoint methodology calculations conservatively assume the worst case normal environmental temperature effect based on maximum environmental temperature change for the specific sensor location." EFE noted that the two statements appeared to be in conflict when the worst case normal environmental temperature change is less than the assumption stated above.

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The project response noted that the conflicting statements in l- DBD-EE-037, Revision 0, apparently came from the existing G&H calculation methodology at the time when the revalidation DBD was being prepared. Further, DBDs were reviewed as part of the design validation program and discrepancies identified and

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29 p j K corrected. Revised DBD-EE-037, Revision 1-A was issued in E October 19, 1987.- SWEC instrumentation and control calculations reference actual environmental conditions existing at the location where the device is physically located. EFE concluded that the cause, extent, and corrective action for the concern were adequately. addresse The NRC team agreed that this EFE concern has~been adequately resolve Action Item No. C04 Drawing 2323-El-0049, Sheet 3, Revision CP-2, which depicts MOV-1-LV-4754, does not show any motor overload device in the control circuit. The motor overload is only shown in the alarm circuit device. EFE noted that the intent of Regulatory Guide 1.106 is to bypass the motor overload circuits during accident conditions, and to have the valve motor protection for normal operation of the valve during testing by not bypassing the' motor overload in the control circuit. As shown on the above drawing, the motor overload is removed from the circuit entirely which means that, during normal testing and stroking

.of the valve, there is no motor overload protection. EFE expressed concern regarding compliance with Regulatory Guide-1.10 The project's response justifies the motor operated valve-protection design by indicating that the thermal overload relays are connected to alarm only and, as such, this meets position C.1 of Regulatory Guide 1.106. The response also states that fuses or thermal magnetic breakers are provided in each motor control center cubicle to protect against degradation of the lines. EFE noted that although the response reveals that the fuses / breakers do not assure that the motor operated valve would be protected for sustained locked rotor current, they would be sized to protect the circuit leading to the motor. This being the case, this satisfies Regulatory Guide 1.106 concerns that the design not result in degrading other safety systems. On this basis, EFE closed this action ite The NRC team agrees that the EFE concern was adequately resolve Action Item No. C08 The review of validated G&H setpoint calculation IC-026 revealed that inputs requiring confirmation have not always been identified. The iny t used in determining the maximum temperature in room 64 was from G&H calculations 701 and 713 and there was no evidence that these calculations had been

, validate EFE expressed concern regarding existing

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l justifications to determine that no confirmation is required as stated in validation record DVP-1-llJ-C-038- The project response noted that the preparer inadvertently omitted the requirement that the reference document for room temperature needed confirmation. In addition, EFE reviewed SWEC calculations 16345-IC(B)-30 through -045 and for these calculations" confirmation requirements were clearly identifie Further, all G&H calculations have been superseded by SWEC issued setpoint calculations (safety related balance of plant)

as directed by TU Electric. Based on the above, EFE concluded that this item was satisfactorily resolve The NRC team agrees with the resolution of the EFE concer Action Item No. C15 FSAR Section 7.3.2.2.8 states that the operator can manually override the automatic operation of individual components by the use of that component's control switch which is located on the main control board. To meet the requirements of IEEE-279-1971 (single failure criteria) under operating bypasses, the FSAR has justified this action by stating that:

if an operator were to turn the control switch to a position that would remove this component from its safety position, the control switches are of a design that, once the operator were to let go of this switch, the automatic initiation signal would again place this component in its safety position. EFE noted that most drawings comply with the above criteria with the following exceptio Electrical schematic drawing 2323-El-0031, Sheet 29, which depicts the control circuit for containment spray pump (cpl-CTAPCS-01) breaker lAPSC1 is in disagreement with that stated in FSAR Section 7.3.2.2.8 in that after the pump had been started on a safety injection actuation sequencer signal, 99 seconds after this input signal was received by the safety injection actuation signal (SIAS), the sequencer will rese After the sequencer had reset, the operator could trip this pump (bypass the SIAS signal for this component) by the use of the control switch on the main control boar If this were to happen, the breaker for the above pump would not reclos The project response noted that FSAR paragraph 7.3.2.2.8 was inadvertently overlooked during review of the containment spray syste The FSAR paragraph 7.3.2.2.8 was revised to correctly identify the design. The response also noted that once a component is bypassed, control of that component is by manual means and that continuous indication of the bypass status is provided by Class lE monitor light boxes. If loss of offsite power were to occur following operator reset of the safety injection signal, the safety injection sequencer is blocked and

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/31 the operator must wait:until<the blackout sequencer allows a restart'of:.the equipmen (Also, See Action. Item No. C17 below.) . Based on this response, EFE concluded that their concerns were adequately addresse TheLNRC team agrees.that the'EFE concerns were satisfactorily resolved.- (Also see Action Item No. C17 below.)

Action Item No. C23 l

L Instrument detail drawing 2323-MI-2103-05, Revision 7, contains

' instrument tubing schematics for connections to pressure sensors. Note 4.of this drawing indicates that the schematics may also be used for differential pressure (D/P) cells with one side vented to atmosphere. EFE expressed concern that details

'on removing vent plugs and how to protect the vented ports on the-D/P cell'are not provided or referenced within the above drawing or the installation specification (CPES-I-1018).

The project response states that note 4 references'.the associated tabulation sheet which. indicates which D/P. cell side requires connection to a. root' valve. In addition, specification CPSES-I-1018-was revised to add requirements for protection of vented. ports of differential pressure type' transmitters and switches. Further, a review of the installation drawings to-find protection methods of vented ports found detail:5G on'

drawing 2323-MI-2105-01. This detail provided the requirements for the control room differential pressure cell transmitter .Therefore, installation' requirements.for: tubing / bug screens have been properly provided for atmospheric pressure applications and the installation of these control room transmitters agrees with the detail drawing. .EFE concluded

.that.these actions appropriately addressed the concer The NRC team agrees that the EFE concerns were appropriately

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resolved.

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Action Item No. C17 Review cf electrical elementary diagram 2323-El-0031, sheet 29, revealed that if the operator tripped the containment spray pumps for any reason after the safety injection system sequencer reset itself and if a blackout and/or LOCA were to occur, the control circuits for the pump breakers would prevent the' containment spray pumps from starting for an additional 99 seconds. EFE noted that existing fluid studies assume in their calculations that during these conditions the total time required to start these pumps and open the containment spray pump discharge valves with the emergency diesels is 25 seconds, not 99 seconds.

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l The project response noted that upon receipt of an 'S' signal l the safety injection system sequencer will load the bus in a predetermined sequence. If a loss of bus voltage were to occur after the 'S' signal up until the operator resets the

'S' signal, upon restoration of bus voltage,,the safety

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injection system sequencer will reload the bu If the operator resets the 'S' signal and the bus loses voltage, upon restoration of voltage, the safety injection system sequencer will not reload the bus. The blackout sequencer would allow the bus to be reloaded manually or by a high containment pressure signal after 99 seconds. Current fluid system studies assume a loss of.offsite power coincident with the event

'instead of this hypothetical case which assumes a loss of power after the operator resets the 'S' signal. EFE concluded the response justifies that if the operator were to reset the safety injection system to switch over to the recirculation mode of operation, the additional time delay of 99 seconds due to loss of offsite power would not affect mitigation of the event since the containment spray pumps would not be required at that point. Based on the above, EFE concluded that the response adequately addressed the concer The NRC team agrees that the EFE concern was appropriately resolve In summary, the NRC team concludes that the reviewed confirmations for the instrumentation and control calculations and the above identified EFE concerns were adequately and appropriately addressed except for the open item identified abov . Electrical Equipment and Cables (51061, 51051, 37051, 51055, and 51065)

The systematic validation process implemented at Comanche Peak for safety-related electrical systems, as stated in the applicant's Corrective Action Program, included the final reconciliation of issues identified in the Post-Construction Hardware Validation program. The final reconciliation process included the consolidation of design validation analysis, hardware modifications, preoperational test results, and inspection activities intended to assure consistent electrical design contro Specifically, the NRC inspector evaluated the resolution of electrical hardware related issues where ,l confirmation of identified design assumptions were require l In order to evaluate the applicant's reconciliation process, .

the NRC inspector reviewed selected Electrical Engineering !

Design Calculation packages. These completed documents were (

reviewed to ensure the validity of design inputs with  !

particular emphasis on final engineering evaluations and/or confirmation of design inputs and assumptions utilized in the design validation proces _ _ _ _ -

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In particular, the NRC inspector reviewed the following Electrical Engineering Calculation packages which contained required confirmations: Calculation EE(B)-142, Revision 0, "Thermolag Tray Interference Analysis." As determined by the applicant, certain safety-related conduits require wrapping (enclosure) with Thermolag because of Appendix R considerations. Due to close tolerance between these conduits and adjacent trays and the requirement to totally enclose the conduit, sections of the adjacent tray may have to be enclosed. Therefore, this calculation determined the maximum section length of power cable trays enclosed by Thermolag which would not impact the basis of the cable sizing calculation (i.e., no additional derating factor or consideration would be necessary).

One confirmation was required relative to the calculational assumption that all level 1 and 2 cable trays utilized had a depth of.four inches or less. Change notice CN-001 for calculation EE(B)-142 adequately addressed the confirmation requirement. In the limited applications where cable tray depth exceeded four inches, the applicant included the appropriate design considerations and controls in an existing progra Procedurally, these cable tray sections are identified in the applicant's drawing M1-1700, Revision 2, "Thermolag and RES Schedule." The NRC inspector reviewed this insulation schedule and determined that the identified cable tray sections which exceeded the assumed design depth were properly characterized. The NRC inspector concluded that calculation EE(B)-142 adequately satisfied the required confirmatio Calculation EE(B)-069, Revision 2, " Voltage Drop Verification - Misc. DC Control Circuits - CPSES Unit 1."

This particular set of calculations determined the voltage drops between the DC fuse panels and the circuit load devices (relays, solenoids, etc.) in order to establish the minimum voltage required at the fuse panels that would permit the circuit to function as designed. Additionally, these calculations compared the minimum voltage required to the minimum voltage available at the DC fuse panels in order to establish that sufficient voltage was availabl The confirmation required for this design calculation consisted of establishing that the assumed minimum voltage required at ASCO valve solenoids was 75% of the nominal voltage or approximately 93.75v DC. The NRC inspector determined that change notice CN-002 for this calculation addressed the confirmation requirements by reference to the applicable ASCO installation and maintenance

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34 instructions. As stated in the subject change notice, the vendors documentation indicated that for these solenoid operated valves, 72% of the nominal voltage was the minimum pickup voltage' required. Based on the above reviews and examination of the vendors catalog, the NRC inspector concluded the the confirmation required for this design calculation was adequately addresse c. Calculation EE(S)-749, Revision 0, " Adequacy of Cable size and Breaker. Trip Settings for Cables EG0002/29 and EG000429A." The subject cables are in series and supply power to the control room air-conditioning Unit 3, CPX-VAACC3.*03 from MCC XEB2-E. The confirmation required the deterraination of actual cable lengths subsequent to the implementation of Design Change Authorization (DCA)

75006, Revision The NRC inspector determined that the actual lengths of the subject cables were established by field walkdown Based on these field walkdowns and as documented in change notice CN-001 to calculation package EE(S)-749, the confirmation required was reconciled. The NRC inspector concluded that CN-001 to calculation EE(S)-749 adequately satisfied the required calculational confirmatio d. Calculation EE(S)-763, Revision 0, " Adequacy of Cable and Breaker Size for cables E0050537 and EG050538." Cable E0050537 is a branch MCC feeder cable which supplies power to the primary plant HVAC heater, CPX-VAFUPK-15 from MCC lEB1-1. Cable EG050538 is a branch MCC feeder cable which supplies power to the primary plant HVAC heater, CpX-VAFUPK-16 for MCC lEB2- In particular, the inspector reviewed the configuration required aspects of this calculation which concerned the determination of actual cable length. The NRC inspector determined that change notice CN-001 to this engineering calculation adequately addressed the required configuration in that the design cable lengths were shown to have adequate margin to allow for expected field run variation e. Calculation EE(B)-074, Revision 0, "XST1 and XST2 Startup Transformers Load Study." The stated objective of this electrical engineering calculation was to verify that the start-up transformers XSTl and XST2 were adequately sized for all operational modes of the modified station service syste The confirmation required aspect of this calculation package involved the verification that the loading of the Unit 2 switchgear was equivalent to the Unit 1 switchgea The NRC inspector reviewed the associated documentation including change notice CN-001 to the subject calculatio _ _ - _

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-35 This. review indicated that for the equipment identified in the design calculation, the' applicant had utilized similar loading characteristics for~both units. The NRC inspector concluded that the required confirmation was adequately addresse f. ' Calculation TNE-EE-CA-0008-267, Revision 01, " Protective Relay Settings.for Emergency Diesel Generators."

Reaffirmation of the emergency diesel generator protective relay settings was necessary in order to incorporate the contents.of the original G&H calculations with the revised

' desig In particular, the validation record specified the resolution of differences between selected' settings requested ~by Operations and those specified in Revision 0 of the subject calculatio The NRC inspector reviewed the subsequent revision of the validation record and the supporting analytical calculations. Based on the. review of these. documents, the NRC inspector' determined that the revision adequately addressed the protective relay settings selected by operations including the above noted' difference Calculation EE(S)-689, Revision 0, " Determination of 125v DC ' Fuse ' Size in Vertical Panel X-CB-01 for Damper Solenoids." The NRC inspector examined the engineering calculations associated with this issue and discussed the

. technical aspects.of this package with the cognizant TU Electric engineer. The required confirmation consisted of the verification of fuse type and size at control panel X-CB-01 and the determination of the maximum slot circuit current of control panel X-CV-01 from the' distribution panels XED1-lCKT No. 7 and SEDZ-1CKT No. .The NRC inspector reviewed CN-001 to the subject calculation and determined that the required confirmations were adequately addressed since the fuse size was verified by the applicant using field inspection techniques and the maximum slot circuit current was established in accordance with S&W calculation 16345-EE(B)-04 Calculation TNE-EE-CA-0006-157, Revision 0, " Coordination Study-6.9kv Power Distribution System." This calculation established the degree of coordination between protective devices on the 6.9kV distribution system (lE and non-1E)

for Unit:1 and' Common busse As' stated in the objectives of this-calculation, the analysis was limited to the coordination among phase and ground over-current device The required confirmation consisted of verifying that all breakers' coordinate as stated in the calculation, with the exception of the high and low side breakers associated with the 6.9kv - 480v transformers.

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The NRC inspector reviewed the supporting documentation for this. calculation including.the associated validation record and Revision 1 to the package. Based on these reviews and examination of.the applicable breaker coordination curves, the inspector determined that the applicant had adequately addressed the required confirmations for the safeguards buses LEA 1 and 1EA2 for both phase fault'and ground fault condition Calculation TNE-EE-CA-0011-504, Revision 0, " Class lE A Lighting Panel Coordination Study." Specifically, this calculation demonstrated that a fault on an AC lighting branch circuit would be isolated from the Class lE MC The required confirmations for this engineering calculation were stipulated to include the verification of referenced electrical one line drawings and supporting calculation The NRC inspector reviewed the calculation validation records (Revision 0 and 1) for calculation TNE-EE-CA-0011-504 and selectively examined the reference drawings and supporting analysis for this package. Based on these reviews the NRC inspector concluded that the applicant had performed the required confirmations and that the supporting documentation adequately reflected this validation proces . Large and Small Bore Piping and Pipe Supports (49065, 50090, 37055, and 35061)

Final reconciliation in the large and small bore piping and pipe supports area was to be accomplished in accordance with TU Electric Procedure CPPP-23, " Pipe Stress / Support Final Reconciliation." This procedure established requirements to reconcile the as-built configuration of the' piping and pipe supports with the validated design input and design criteria and included requirements for completion of piping DVP The scope of the procedure was limited to: (1) all ASME Code,Section III, Class 1 small bore piping of 1-inch nominal pipe size or less; (2) all ASME Code,Section III, Class 2 and 3 piping; (3) all ASME Code,Section III, Class 1, 2, and 3 pipe supports; (4) all Class 5 or 6 piping and pipe supports within ASME Code,Section III, stress problem boundaries; and (5) all high energy piping identified in TU Electric letter NE-1852 dated November 4, 198 As-built configuration inputs for the final reconciliation program were to be obtained from the As-built Inspection of Piping and Pipe Supports Program and the PCHVP, and validation

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design inputs from design requirements in applicable DBDs and design procedure . _ _ _ _ _ _ _ _ _ _

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- 37 The NRCTteam determined that the~ effective issue of Procedure CPPP-23 was Revision 2 dated September 19, 1988, including.

l change notices 1 through 8' dated October 3, 1988, through l July ll, 1989. It.was also determined.that.95 of the approximately 800 DVPs for Unit l'and common areas within th scope.of CPPP-23 had been' complete ~

These approximately.

l 800 DVPs include the ASME Code,Section III,-safety-related L piping and pipe supports in 24 service. systems in Unit 1 and common areas. Of the 95 completed DVPs,Lthe NRC team reviewed a random sample of 9 to assess the adequacy of'the implementation of the final reconciliation progra The NRC review of Procedure CPPP-23, Revision 2, Change Notices 1 through 8 found that the final reconciliation process to consolidate'the results of' design validation, hardware modification, and inspection documentation was adequate to assure consistency of the hardware with the validated desig The-review found that Section 6.0, " Final Reconciliation

~ Procedure," of CPPP-23 specified extensive requirements for: The preparation and control of input information for pipe stress reconciliation (Section 6.2); Engineering evaluations (Section 6.3) including:

(1) Problem Review Documentation Package (PRDP), Thermal Mode Sketch (TMS), and Pipe Support Review Data (PSRD) calculation activities (Section 6.3.2);

(2) Pipe stress calculation activities (Section 6.3.3);

(3) Pipe support calculation activities (Section 6.3.4);

(4) Processing of modifications and design changes (Section 6.3.6); and (5) Post-reconciliation changes (Section 6.3.8); Notification of Approval (NOA) of completion of final reconciliation (section 6.4);

. Control of final reconciliation of calculations (Section 6.5); Document control (Section 6.6); Transmittal of results to and from other organizations (Section 6.7); and Preparation of design validation packages (Section 6.8).

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In addition, the NRC review found that due to the high volume of information to be exchanged between the pipe stress and support and other organizations, special requirements were deemed necessary to control and track the exchange of information. These requirements were defined in Project Procedure PP-107 (Procedure CECO-12), " Preparation, Approval and Distribution of the Information of Pipe Stress Analysis (IPSA)," Revisions 0 dated December 14, 1988, including change notices 1 through 3 dated February 14 through July 14, 198 This procedure established a consistent method for the identification of validated source documents requisite as input for the reconciliation of ASME Code,Section III pipe stress and pipe support analyse !

The sample DVPs selected for review are identified in the following table. Each DVP pertained to a particular stress problem; i.e., a portion of piping and associated pipe supports in a system which could be analyzed independently from the connected piping or components in the system. The stress problem, service system, and piping and pipe support data for each of the nine sample packages are as follows:

Stress Service Nominal Approximate No. of i Problem System * Pipe Size Length of Pipe Supports  !

1-S265 RH 3/4 i f N007 CS 3 i f S035B CC 3/4 i f ** CC 4 i f S116 CC 3/4 i f DD 4 i f S061 RH 3/4 i f S178 CC 3/4 i f B DD 3/4 & 4 i f *RH = Residual Heat Removal system CS = Chemical and Volume Control system  !

CC = Component Cooling Water system )

DD = Demineralized and Reactor Makeup Water system

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identified as reconciled (details provided below).

Each of the 9 DVPs was found to contain the following l documents:  ;

IPSA document, BRP drawings - piping isometrics, BRHL/GHH drawing - hanger location drawings,

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PSRD (Pipe Support Review data) - pipe stress and support system review calculatio SID (System Information Document),

PRDP (Problem Review Documentation package),

TMS (Thermal Mode Sketch),

Pipe stress calculation, Pipe support calculations / drawings, and DVP validation letter and ASME Code, Section.XI, NIS-2 form (if issued).

For each package, the NRC team also reviewed the applicable DB The NRC team found the contents of the packages and their ,

indices to be in accordance with the requirements of  !

Section G.8 of Procedure CPPP-23. Subsequently, the NRC team reviewed each package for compliance with the applicable requirements of Sections 6.2 through 6.7 of Procedure CPPP-2 The results of the reviews were similar and the results of one of the reviews will be documented herein to demonstrate their exten The NRC review of the DVP for stress problem 1-313 found the following: DBD-MG-229, " Design Basis Document, Component Cooling Water System," Revision 4, December 22, 1988, did not contain any confirmation required item Tracking systems for the reconciliation process were developed and maintained in accordance with the requirements of Section 6.1 of Procedure CPPP-2 Tracking systems developed included the "PSAS Calculations and CCNS by Calculation Number," " Pipe Support Calculation Index," and " Pipe Stress / Supports Confirmation Log" indices. These indices appeared to the NRC team to be complet The following reconciliation inputs were collected and compiled in accordance with the requirements of Section 6.2 of Procedure CPPP-23: (1) IPSA 1100,

"Information for Pipe Stress Analysis (IPSA),

System 1100 - Component Cooling System," Revision 0, February 1, 1989, including change notices 1 through 2, July 12 through July 26, 1989; (2) SID-CC, " System Information Document, Component Cooling Water,"

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Revision 7, December 6, 1988; (3) PRDP-CC-1-313, " Stress

- Problem CC-1-313," Revision 0, November 4, 1988; (4) TMS-CC-1-313, " Stress Problem 1 - 313," Revision 1, December 19, 1988; and (5) PSRD, " Problem 1-313,"

Revision 0,-including CCN 1, July 10, 1989. The NRC team's review of the confirmation required items in the

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revisions and change notices to the documents verified i that no outstanding confirmation required items relating L to stress problem 1-313 were outstanding and that closure of confirmation required items were satisfactor The NRC team's review of 1PSA 1100, Revision 0, found that

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the cover page of the IPSA indicated that except as specified in attachment A to the IPSA, all information identified in the IPSA had been verified and contained no requirements for confirmation of items pertinent to the stress analysis for stress problem 1-313. Attachment A to the IPSA was found to identify the following items which affected seven sections of the IPSA which required confirmatio Item Affected N IPSA Section Description 1 Displacements from El. 960 and above are not appropriate for stress analysi .1 Systems Interaction Progra .2 Work in progres . Pending revision to SID-C .4. Pending valve end load requirement .4. Pending accelerations for non-Q valve . Pending Westinghouse review of final reconciliation loads on equipmen . Pending resolution of seismic classification of equipment issue '

in IMT-285 The NRC team found that, except for item 3, all of the preceding items were referenced in the body of IPSA 1100, Revision i

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In addition, the NRC team found many parenthetical stipulations in IPSA 1100, Revision 0,.that verification of certain types of information was required. For example, Sections 1.1 and 1.2 stipulated "SWEC/ Mechanical to Verify" and "SWEC/ Mechanical ~to Verify Drawing &

Revision Level," respectively. None of these stipulations were contained in Appendix A to the'IPSA. Nonetheless, all organizations responsible fer verification of inputs to the IPSA had concurred with IPSA 1100, Revision 0. The final reconciliation coordinator explained to the NRC team that inclusion of these parenthetical stipulations were intended to be used as aids and were in addition to the basic requirements of the IPSA. The parenthetical stipulations served only to identify organizations responsible.for the input to various sections to the IPS The NRC team concluded that the explanation was satisfactor Change notice 1 to IPSA 1100, Revision 0, dated July 12, 1989, was extensive in scope. This change notice closed (by " Deleted" designations) items 2 through 5 and 7 of Appendix A to IPSA 1100. The NRC team verified that closure of these items was acceptable. For example, items 2 and 3 were closed on the basis of the revised inputs to Sections 6.1 and 6.2, respectively, provided by the change notic In addition, all the above described, erroneous, parenthetical stipulations were removed by change notice The two outstanding confirmations required items in IPSA 1100, Revision 0, change notice 1, were found by the NRC team to be not applicable to stress problem 1-31 The NRC team concluded that this issue of IPSA 1100 was acceptable for final reconciliation of stress problem 1-31 Changes effected by IPSA 1100, Revision 0, change notice 2, were of no significance to stress problem 1-31 The NRC team found that SID-CC, Revision 7, PRDP-CC-1-313, Revision 0, and TMS-CC-1-313, Revision 1, did not contain any confirmation required items. The issues of the PRDP and TMS reviewed by the team were issued specifically for the final reconciliation of stress problem 1-31 Reviews of the PRDP, TMS and PSRD required by Section 6.3.2 of Procedure CPPP-23 were completed and all confirmation required items were closed. The NRC team's review of the status of the confirmation required items verified, except as described in the following, that no confirmation required items were outstandin _ _ _ - _ - - _ _ _ _ _ -

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As reported in paragraph 10(c), the results of the NRC team's review of the PRDP and TMS confirmed that there were no confirmation required items.

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However, the NRC team's review of PSRD-313, Revision 0, October 19, 1988, found that PSRD-313 contained confirmation required items relating to issuance of the calculation for primary pipe support stress problem 1-31 During the inspection, the NRC team found that confirmation. required items in this PSRD were incorrectly closed-by an interoffice correspondence (IOC) dated July 10, 1989. Closure of confirmation required by IOC items was permitted by Section 6.5 of CPPP-23 which states that confirmation required items may be removed by means of a memorandum. However, closure of PSRD 1-313 was incorrectly based on the July 10, 1989, closure of confirmation required items in change notice No. 1 to PSRD-1-S313 instead of PSRD-1-313 (closure of confirmation required items in PSRD-1-313 was possible during the time of the audit since the calculation for stress problem 1-312 was issued in November 1988). Subsequently, PSRD 1-313 was closed by IOC dated August 23, 1989, on the basis of the issuance of PSRD-31 The NRC team found that closure of lPSRD 1-313 was acceptable on this basi TU Electric explained that similar errors had been found in the final reconciliation program and had developed a DVP checklist to verify the acceptability of inputs to DVP The NRC team concluded that PSRD 1-313 was incorrectly closed during the audit due to a documentation error which was subsequently corrected. However, a program had been developed to identify and correct such errors. The NRC team deemed the program to be necessary and appropriat e. Evaluations and reviews in the pipe stress area required by Section 6.3.3 of Procedure CPPP-23 were completed and all confirmation required items were closed. The NRC team's review of stress calculation 15454-NP(N)-CC-1-313, Revision 0, July 22, 1987, including change notices 1 and 2, October 18, 1987, and May 25, 1989, and IOC, July 14, 1989, verified that no confirmation required items were outstanding and that closure of the confirmation required items was satisfactor Section 2.0 of Revision 0 of the stress calculation indicated that 5 of the 6 assumptions used in the analysis required confirmation. These assumptions related to:

(1) the use of preliminary piping configurations and pipe support locations, (2) the support configuration of the component cooling water surge tank, (3) support weight on

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the piping, (4) the use of a thermal mode sketch based on a preliminary PRDP, and (5) the use of generic stiffnesses !

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for all new supports added to the piping syste Furthermore, Section 9.0 added a sixth confirmation required item relating to confirmation of modifications associated with the addition of'four new pipe support Change notice 1 to the stress calculation did not affect the confirmation required status of the stress calculatio Change notice 2 to the stress calculation closed all six of the previously identified confirmation required items but identified the following three new items: (1) fluid t.ansient inputs, (2) jet impingement loads / pipe whip data inputs, and (3) the exceeding of equipment allowable nozzle loads. Closure of the six previously identifi9d confirmation required items was acceptable to the applicant on the basis of the information available at the time of issuance of the change notic The IOC closed the confirmation required items identified in change notice 2 on the basis of IPSA 1100, Revision 0, change notice 1; PM-165, Revision 2; and IMPELL letter IM-T-210-62-0683. The NRC team found that closure of the items was acceptable on the basis cited. In addition, the Ioc formally closed the confirmation required items in Revision 0 of the stress calculatio f. Evaluations and reviews in the pipe support area required by Section 6.3.4 of the procedure were also completed and all confirmation required items were closed. The NRC team found that hanger calculations 15454-NZ(S)-CC-1-993-700-A73R, Revision 0, through 15454-NZ(S)-CC-1-993-703-A73R, Revision 0, all including change notice 1, all dated May 18, 1989, were satisfactorily closed by similar Iocs all dated July 10, 1989. All the confirmation required items in the hanger calculations related to the final issuance of the pipe stress calculatio In addition, the NRC team's review of hanger drawings and DCA/CRs CC-1-993-700-A73R, Revision CP-1; CC-1-993-701-A73R, Revision CP-1; CC-1-993-702-A73R, Revision CP-1 and DCA/CR 71041, Revision 0; and CC-1-993-703-A73R, Revision CP-2 ind DCA/CR 71908, Revision 4, verified that there wcre no confirmation required items outstandin g. Based on the results in paragraph a through f, it was concluded that except for the documentation error in paragraph d, implementation of the final reconciliation process of the DVP for stress problem 1-313 was

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acceptable. Similar results were obtained for the other B DVPs in the sample selecte Based on the results of: (1) the review of the final reconciliation program as defined in Procedure CPPP-23, and (2) the review of the final reconciliation process in the 9 sample DVPs, the NRC team concluded that the final reconciliation program and its implementation in the piping and pipe support area are acceptabl . Train A and B and Train C Larger Than 2" Diameter Conduit and Conduit Supports (37055, 37051, 52055, 51051, 51055)

For final reconciliation in this area, Ebasco developed Procedure SAG CP35, Revision 5, " Technical Guidelines for Conduit System Design Validation Package Close-Out." This procedure provides the guidelines for backfit design reconciliation and closeout of those Unit 1 conduit packages which were completed prior to the latest revisions of the implementing procedures and the S-0910 drawings which had design impact on previously completed wor Appendix C to this procedure provided a checklist which contained 28 items which required reverificatio In Revision 5 to the above procedure, it is stated that based on a review of a minimum of 10 percent of the total number of conduit packages in each building, it was determined that the enhanced design criteria had no design impact on already completed calculations for all of the SAG CP35 check list items except for clamp This is documented in Ebasco's position paper on " Design Reconciliation of Seismic Conduit Packages Train A, B and Train C > 2" Diameter" dated February 3, 198 The NRC inspector reviewed the results presented in this position paper which revealed that out of a total of 6097 calculation packages, a detailed review, in accordance with the requirements of SAG CP35, was performed on a sample of 822 (13.5%) packages. These packages contained approximately 4100 supports. The only adverse impact that was encountered was that 14 clamps, out of a total of 5141 evaluated, exceeded design allowables. For the remaining 27 items reviewed for each system / support, there were no adverse finding Therefore, based on results found in this sample, Ebasco's position was that for these 27 items, no further evaluation was necessar The NRC inspector concurs with this conclusio Based on the conclusions outlined above for this functional area, the NRC inspector's review of reconciliation in this area consisted of a review of weld related Component Modification Cards (CMCs) for weld size and configuration changes, confirmation of footprint loads (FPL) and clamp loads, as well as closecut of open items such as NCRs DCAs, et _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ - _ _ _ _ - _ _ - _ _ _ _ -

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The following calculation packages were reviewed by the NRC inspector:

Calculation N Conduit Building No. of Supports 12372 C12K12372 Reactor 5 21336 C14021336 Safeguards 5 07975 C12007975 Electrical 7 03416 C12003416 Safeguards 13 21432 C14G21432 Reactor 21 11969 CO2011969 Electrical 5 10539 C15B10539 Electrical 3 14143 Cl3G14143 Safeguards 1 14202 C13G14202 Safeguards 3 32179 CO3032179 Auxiliary 11 In these calculation packages there were no assumptions which required confirmation. The NRC inspector confirmed that reconciliation of clang loadings had been performed when required. The NRC inspector reviewed document packages in the plant records center to insure that work required by the DCAs generated as a result of the design validation process had, in fact, been implemented as required and that the necessary QC inspections were satisfactorily completed. The NRC review also included whether any NCRs written against either the conduit or the conduit supports had been completely dispositioned and that concurrence of footprint loads had been received by Ebasco prior to closecut of the calculation package. Based on the above actions, the NRC inspector determined that the final reconciliation process was satisfactorily implemented for these calculation No violations or deviations were identifie . Cable Tray and Cable Tray Supports (37051, 37055, 51051, 51055)

The NRC team reviewed the reconciliation packages for the following cable tray system analyses:

String N Revision RSM-1-AUX-39 3 RSM-1-AUX-42 4 176-64-03 2 192-154-02 3 X-037-03 3 The NRC team reviewed Technical Procedure ECS 5101, Revision 1, which is the guideline for design and analysis of nuclear safety-related cable trays (electrical raceways) and cable tray hangers (CTHs). IMPELL's final reconciliation in this area consisted of reviewing each calculation package to ensure that f.

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I . 4 cable tray fill loads and thermolag. weights were properly evaluated in light of current conditions. If conditions had changed so that the existing analysis was no longer accurate or l conservative, a complete reanalysis was performed to meet all the initial design requirements of ECS 510 The NRC inspector reviewed each of the above analyses to ensure that cable fill load for each tray segment had been recalculated and that the weight of thermolag had been included where required. Also, each analysis was reviewed to ensure that there were no unresolved open items, all the assumptions were reasonable, all DCAs and NCRs had been adequately factored into the tray analyses, and there were no further items which required confirmatio The NRC team also reviewed the analysis packages for the following cable tray hangers:

Support N Revision CTH-1-2954 1 CTH-1-365 1 CTH-1-3355 1 CTH-1-539 2 CTH-104621 2 CTH-1-4731 2 CTH-1-731 2 CTH-1-733 2 CTH-1-2377 2 CTH-1-2378 2 These supperts were reviewed to insure that the results of the reconciled tray analyses, including current cable fill and thermolag data, had been utilized as the input to the support analyses, and that all assumptions and open items were reasonable and adequately resolve Based on the above review, the NRC team determined that the final reconciliation process was satisfactorily implemented for cable tray and cable tray support . Review of Resolution of Technical Audit Program (TAP) Findings and Engineering Functional Evaluation (EFE) Action Items (35061, 37051, 37.055)

The NRC inspector selected three TAP audits for inspection to determine if previous audit findings and EFE action items were properly addressed during final reconciliatio Audits selected were:

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Audit N Scope ATP-88-120 Unit 1 pipe stress and support (PSAS)

final reconciliation proces ATP-89-134 Unit 1 pipe stress and support (PSAS)'

final reconciliation proces ATP-89-142 Electrical corrective action program final reconciliation proces The NRC inspector found that prior to performing an audit, the audit team leader reviewed previous TAP audits for open andit findings or for audit items requiring verification by the TAP audit group for closure. Further, when applicable, open EFE action items were included in the scope of the audit to verify that the EFE action items were properly resolved. Results of the NRC inspectors review of each audit follows:

Audit ATP-88-120 This audit specifically addressed 12 open EFE action items and certain commitments made in response to previous TAP audits and surveillance. Additionally, the audit addressed implementa-tion of the final reconciliation process for pipe stress and support Audit ATP-88-120 determined that (1) coverage of commitments made in response to previous TAP audits or surveillance findings was satisfactory, and (2) responses to the 12 EFE action items were also satisfactory. However, the audit determined that implementation of the final reconciliation process was, at that time, unsatisfactory for this area (due to a number of technical and programmatic inadequacies) and therefore would be the subject of a subsequent audit after completion of corrective action The NRC inspector reviewed the audit checklist, the evidence observed, and the auditor's conclusions, and determined that the checklist and evidence observed appropriately supported the conclusions. The NRC inspector concurred with the audit team leader's conclusio Audit ATP-89-134 This audit addressed implementation of the pipe stress and support (PSAS) final reconciliation process. The audit divided the process into three areas: (1) final reconciliation input, (2) final reconciliation process, and (3) technical adequac Additionally, the audit addressed closure of 15 previous TAP audit finding The NRC inspector reviewed the audit checklist, the evidence observed, and the auditor's conclusions. The NRC inspector

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determined that the checklist and the evidence observed

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~ supported the auditor's conclusions-that the final reconciliation process was adequately implemented for this are Audit ATP-89-142 This audit addressed the implementation of the final-reconciliation process for two electrical systems: (1) the uninterruptable power system (UPS), and (2) the 6.9kv. power l

system. The audit specifically addressed follow-up of open

items for. audits ATP-88-98 and ATP-88-10 The audit found numerous minor' deficiencies'in the design validation' packages; L

however, the audit determined that the' implementation of the

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' final reconciliation process for.this area was satisfactory.

l l The NRC inspector reviewed the audit checklist and the evidence I observed by the auditors. Additionally, the NRC inspector

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discussed.the' numerous deficiencies with the audit team leade Based on the discussion with-the audit team leader and further i

review of the evidence observed, the NRC inspector concurred L with the' audit team leader's conclusion.

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Based-on the satisfactory follow-up of 2AP audit findings and EFE action items, the NRC inspector determined that TAP findings and EFE action items are satisfactorily addressed'in

'the final reconciliation proces . Open Items Open items are matters which have been discussed with the applicant, which will be reviewed further by the inspector, and which involve some action on the part of the NRC or applicant or both. One open item disclosed during the inspection is discussed in paragraph 8, pages 25 and 2 . Exit Meeting (30703)

An exit meeting was conducted July 28, 1989, with the applicant's representatives identified in paragraph 1 of this report. No written material was provided to the applicant by the inspectors during this reporting period. The applicant did not identify as proprietary any of the materials provided to or reviewed by the inspectors during this inspection. During this meeting, the NRC inspectors summarized the scope and findings of the inspection.

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