ML060860028

NRC Request for Information Relating to Event Notification 42129

ML060860028
Person / Time
Site:	Point Beach
Issue date:	02/16/2006
From:	Koehl D L Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
42129, FOIA/PA-2010-0209, NRC 2006-0009
Download: ML060860028 (143)

Text

Point Beach Nuclear Plant N MC IJI Operated by Nuclear Management Company,LL C Carmrtied to Nzxi ExceIene February 16, 2006 NRC 2006-0009 10 CFR 50.46(b)(5)

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington,DC 20555 Point Beach Nuclear Plant,U nits 1 and 2 Dockets 50-266 and 50-301 License Nos. DPR-24 and DPR-27 NRC Request for Information Relating to EventN otification 42129 Reference (1) NRC Letter to NMC Dated January 10, 2006 On November 8, 2005, Nuclear Management Company, LLC (NMC) notified the NRC staff in accordance with 10 CFR 50.72 that the design basis for long-term cooling att he Point Beach Nuclear Plant Units 1 and 2 (PBNP), was notc orrectly modeled. This notification stated that, "These errors in the modeling fidelity potentially impact the analytical basis for demonstrating compliance with the acceptance criteria of 10 CFR 46(b)(5), long-term cooling.'Via Reference (1), the Nuclear Regulatory Commission (NRC) requested additional information to enable NRC staff review of this event. Enclosure 1 of this letter provides an executive summary of NMC's response to this request for additional information.

Enclosure 2 of this letter provides the detailed response to each of the questions in the RAI. Enclosure 3 provides a listing of the documents requested by the RAI, which are enclosed on a CD-ROM.Summary of Commitments:

Resolution of the nonconformances listed in response to Question 1.C in Enclosure 2, consistent with the existing NMC commitment to resolve GSI-1 91, will be achieved by December 31, 2007.Please contactm e at 920/755-7624 if you need further information on this matter.Dennis L.K oehl Site Vice-President, Point Beach Nuclear Plant Nuclear Management Company, LLC Enclosures AOO(cc: NRR Project Manager PBNP Resident Inspector 6610 Nuclear Road

• Two Rivers, Wisconsin 54241-9516 eq Ia l cis Telephone:

920.755.2321

.op- Rat' ur go 1 PDF created with pdfFactory Pro trial version www.pdffactorv.com ENCLOSURE I EXECUTIVE

SUMMARY

OF RESPONSE TO RAI ON EN 42129 On November 8, 2005, in accordance with 10 CFR 50.72, Nuclear Management Company,L LC (NMC) notified the NRC thatt he design basis for long-term cooling at the Point Beach Nuclear Plant Units 1 and 2 (PBNP), was notc orrectly modeled. This notification stated that, "These errors in the modeling fidelity potentially impact the analytical basis for demonstrating compliance with the acceptance criteria of 10 CFR 46(b)(5), long-term cooling." Region IlIl dispatched an inspector to the plant site to assist the PBNP resident inspectors in evaluating the potential safety significance of the notification.

During the approximate three-week period of NRC inspection, inspector questions,c oupled with NMC's continuing evaluation of emergency core cooling system (ECCS) long-term cooling issues, resulted in the identification of additional issues. These issues are discussed in detail in Enclosure 2.Via Reference (1), the Nuclear Regulatory Commission (NRC) requested additional information to enable NRC staff review of these issues. Enclosure 2 of this letter provides NMC's specific response to this request for additional information.

Enclosure 2 contains several attachments.

These attachments depict the degraded or nonconforming coatings on the three elevations in each containment, an elevation view of containmentE levation 8' (showing residual heat removal system piping extending below the containment floor and ECCS suction screens), a plan view of the piping from containment to the auxiliary building, and provides details of the SI-850 valves (including the valve cylinder and operator).

Enclosure 3 of this letter provides a CD-ROM containing the records thatw ere also requested in the NRC letter, along with additional operability determinations that were made following the discovery of several related issues during resolution of the ECCS evaluation modeling errors that were reported.

The following discussion summarizes the efforts taken to ensure that the discovery of issues to the analytical design basis associated with long-term core cooling do not pose a safety hazard.The initial discovery of potential ECCS modeling errors occurred while answering inspector questions related to degraded and non-conforming coatings.

This discovery centered on identification that the condition of coatings in the PBNP Unitl containment were beyond the analytical bases established in the licensee response to Generic Letter (GL) 98-04. Evaluation of the existing conditions of coatings,w hen compared to those assumed in the licensing basis calculations, led to discovery of several errors in the calculations (as reported in LER 266/301/2005-006-00).

Existing conditions in containmentp ertaining to coatings and other sources with potential to impact sump strainer capabilities,e quipment configuration and conditions (specifically sump suction valves and suction piping for residual heat removal), and analyses were reviewed and evaluated to verify long-term core cooling capability.

The containmentc oatings conditions were initially evaluated, in concert with other potential debris sources within containment, to determine potential for clogging the ECCS suction strainers.

The aggregate condition was determined to be technically Page 1 of 2 PDF created with pdfFactory Pro trial version www.pdffactorv.com acceptable, but posed a non-conforming condition with the PBNP licensing basis due to the errors in the analyses.During review of the containmentc onditions, it was discovered that the existing net positive suction head (NPSH) analyses for the RHR pumps did not consider the effects of a potential debris accumulation on the ECCS suction strainers.

An evaluation of this condition concluded that adequate NPSH existed. However, this evaluation concluded that a compensatory measure was required to restrict RHR flow during a specific system post-accident alignment for boron precipitation, and posed a nonconformance with the PBNP licensing basis due to the need to credit a containmentp ressure increase associated with equilibrium pressure/temperature conditions in the sump (a nonconformance with the PBNP response to GL 97-04). Specific modeling associated with the pressure drop of the postulated debris collar,a long with refined modeling of the pressure drop past the SI-850 valve; require 1.45 psig of containment pressure to preclude flashing.Reviews of the material condition,o perating history, maintenance history,a nd licensing basis associated with the SI-850 valves revealed inconsistencies in the safety function and testing/design basis requirements of these valves. The condition was evaluated and determined to be technically acceptable, but the safety function must be incorporated into various programs.The review of issues and conditions to assure long-term core cooling has revealed a number of issues with the responses to GLs 97-04, 98-04 and 2004-02,a nd Bulletin 2003-01. Additionally, the resolution of the issues identified in GL 2004-02 will revise the design and license bases associated with the ECCS suction strainers via detailed analyses, plant modifications, and establishment of additional administrative controls.

Resolution of the nonconformances will be integrated to ensure thatf inal resolution of the issues identified in GL 2004-02 will encompass completion of an aggregate resolution of each of the above currentc onditions.

As a resulto f deficiencies identified with the PBNP containmentc oating program, an RCE was performed.

Significantc orrective actions include an upgrade of the PBNP coatings program and implementing procedures.

NMC has concluded that the ability has been maintained to safely assure long-term core cooling in accordance with the provisions of 10 CFR 50.46. NMC will continue to evaluate the need to update our responses to various generic communications, including the need for license amendments, as we finalize our resolution of each nonconformance and final resolution of GL 2004-02. We will continue to evaluate issues associated with the plant's ability to satisfy long-term core cooling in an integrated and comprehensive manner.Specific details of each issue are provided in Enclosure 2.Page 2 of 2 PDF created with pdfFactory Pro trial version www.pdffactorv.com ENCLOSURE2 RESPONSE TO RAI ON EN 42129 On November 8, 2005, Nuclear Management Company, LLC (NMC, the licensee), notified the U. S.N uclear Regulatory Commission (NRC) in accordance with 10 CFR 50.72 (Event Notification 42129) that the design basis for long-term cooling at the Point Beach Nuclear Plant (PBNP),U nits 1 and 2, was not correctly modeled. NMC's notification stated that,"These errors in the modeling fidelity potentially impact the analytical basis for demonstrating compliance with the acceptance criteria of 10 CFR 50.46(b)(5),L ong-term cooling." On January 10, 2006, the NRC staff issued a Request for Additional Information (RAI) to NMC. The texto f the RAI follows as Enclosure 2, with NMC's response to each of the items."The NRC staff is reviewing NMC's actions to establish that the requirements of 10 CFR 50.46(b)(5) continue to be met. The staffs review includes the potential blockage of the containment sump and its effect on the ability to sustain long-term cooling, the potential impact of the SI-850 valves to operate so as to sustain long-term cooling, and the potential impact of leakage from the recirculation line, particularly regarding dose to operators.

The NRC staff has determined thatr esponses to the following questions are needed to proceed with this review.For each of the questions below,p lease ensure thaty our responses describe your assumptions,m ethods,a nd conclusions in sufficient detail to support the NRC staffs independent review. If technical reports are referenced, you should provide a copy of the report and the technical basis for the applicability of the reports to your facility." 1. General A. Provide a discussion of actions taken to demonstrate the ability to establish and maintain long-term cooling in accordance with 10 CFR 50.46(b )(5).NMC Response: There are substantial ongoing reviews thath ave resulted in proposed changes in the area of containment sump screen design criteria, effects on downstream components, debris sources,d ebris transport, etc. These changes are being developed and implemented in response to Generic Safety Issue 191 (GSI-191)as communicated to the industry in NRC Bulletin (BL) 2003-01, Generic Letter (GL) 2004-02 and other associated communications.

PBNP will comply with the revised acceptance criteria established via GSI-191 by December 2007 as stated in NMC's response to GL 2004-02 dated September 1, 2005.The following NMC responses are provided within the context of events and discoveries made pertaining to containment coatings and related issues during late 2005. The scope of the responses is confined to the station design and license bases, as they existed at the time of discovery unless otherwise noted.Page 1 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com PBNP has met the requirements of 10 CFR 50.46(b)(5) via the following actions: 1969: The design of the screens predated the issuance of RG 1.82, Revision 0,'Sumps for Emergency Core Cooling and Containment Spray Systems," (issued in 1974), but reflects portions of that guidance.

Features include:* Separation from high energy piping systems by structural barriers (C.2)* Located on the lowestf loor elevation of containment with a trash rack and a fine inner screen (C.3)* No drains terminating so as to impinge water (and entrained debris) on the screens (C.5)* A substantial vertically mounted trash rack (C.6)* A vertically mounted inner screen designed for 0.2 fps with 50% screen blockage (C.7)* A solid top deck that would be submerged after completion of safety injection (C.8)* Seismic rack and screens (C.9)* Screen openings sized based on the minimum restrictions of downstream components (C.1 0).* Corrosion resistant materials (C.12)1989-1990:

In response to concerns about blockage of the sump screens by debris, including GL 85-22, "Potential For Loss of Post-LOCA Recirculation Capability Due to Insulation Debris Blockage," and Regulatory Guide (RG) 1.82, Revision 1, detailed unit-specific analysis of debris generation and transport were performed to consider debris from both coatings and insulation.

The potential adverse effects on downstream components and the reactor core were also evaluated.

The analyses concluded acceptable performance of the ECCS system without additional modifications or changes to the plant. Copies of these analyses are provided on CD-ROM as listed in Enclosure 3.1992: The station developed a hydraulic model of the integrated ECCS system. This model provided the ability to perform more sophisticated analyses and evaluation of alignments and scenarios not previously considered.

As a direct resulto f this model and subsequent refinements to it, the station identified that operation of containment spray in the "piggyback" mode may result in insufficient net positive suction head (NPSH) to the operating residual heat removal (RHR) pump. The option to operate in this alignmentw as removed from the emergency operating procedures.

1994: Recognized that prolonged simultaneous operation of both trains of emergency core cooling system (ECCS) during the injection phase would rapidly deplete the RWST inventory and challenge the ability to successfully transition to sump recirculation prior to losing the suction source. Emergency Page 2 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com operating procedures were revised to direct securing a single train and prolong the suction source. (This change pre-dated the Candidate Operator Actions of BL 2003-01).1997: Responded to GL 97-04, "Assurance of Sufficient Net Positive Suction Head for Emergency Core Cooling and ContainmentH eat Removal Pumps," reaffirming adequate NPSH for the ECCS pumps.1998: Established a refueling frequency inspection of containment coatings and maintaining a detailed inventory of those thata re unqualified or degraded.The inspections are reflected in the station's response to GL 98-04, "Potential for Degradation of the Emergency Core Cooling System and the Containment Spray System After a Loss-of-Coolant Accident Because of Construction and Protective Coating Deficiencies and Foreign Material in Containment." 1998-1999:

To support the response to GL 98-04, two more unit-specific transport analyses were performed to evaluate the potential for transporting potential coatings debris to the sump screens, and the effect of such blockage on the NPSH available to the emergency core cooling (ECCS) pumps. These analyses used NUREG/CR-6224, "Parametric Study of the Potential for BWR ECCS Strainer Blockage due to LOCA Generated Debris," October 1995, to estimate the potential head loss. The analyses established a "Zone of Influence" that required heightened awareness and maintenance of coatings within the zone.1999-2005:

PBNP maintained an informal inventory of degraded and unqualified coatings inside containment.

The as-found deficientc oatings were evaluated by informally re-performing the transport and head loss analyses performed in 1998-99 and ensuring thatsc reen head losses were still acceptable.

2003-2005:

PBNP implemented Candidate Operator Actions (COAs), as appropriate, and consistentw ith the NMC response to BL 2003-01. These actions included:* Operator training on sump clogging.* Stopping unnecessary redundant pumps in 1994..Implementing more aggressive foreign material control of containment.

.Ensuring containment drainage paths are unblocked.

This was previously performed by installation of a strainer in the refueling cavity drain,a nd use of reflective metal insulation (RMI) on the reactor vessel head..Ensuring sump screens are free of adverse gaps and breaches (Technical Specification surveillance requirements R 3.5.2.6)..Initiating analyses necessary to resolve GSI-191 issues.Page 3 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com 2003-2005 (continued)

.Providing a more aggressive cooldown and depressurization following a small break loss of coolante vent( LOCA)..Providing guidance to refill a depleted refueling water storage tank (RWST);providing symptoms and identification of containments ump blockage, and developing contingency plans in response to sump blockage, loss of suction, and cavitation.

.Injecting more than one RWST volume.2001: NMC determined a potential for a higher head loss across the containment sump outletv alves (1&2SI-850A&B) than previously recognized.

This was because the earlier models used the head loss value for a standard valve, yet these outlet valves have an unusual configuration and typical valve head loss factors should not have been used. The head loss was recalculated using a summation of entrance, exit, contraction, and expansion head losses, and the NPSH calculation (N-92-086) was revised accordingly.

2005: NMC evaluated the potential impact of emergentc oncerns related to the containment sump outlet (1 &2S[-850A&B) valve due to the postulated formation of a 'debris collar" (see the NMC responses to Question 4 below).2005: In October/November,i t was discovered or recognized that:.The inventory of degraded and unqualified coatings was no longer bounded by the 1998-1999 analyses..The methodology that had been used for calculating the head loss across the screens in those analyses was non-conservative.

.Air entrainmentr ather than NPSH is the limiting factor for the RHR pumps when operated in the post-accident sump recirculation mode (due to the partially submerged sump screens)..A postulated "debris collar' around the sump outlet valves could lead to a significantly higher head loss at the sump outlet than previously evaluated.

As a result of these and other findings documented in the corrective action programs everal operability determinations (OPRs) and associated corrective action items were initiated.

Each of the issues addressed in the operability determinations,t he technical basis for the operability determinations, and the conclusions of the determinations are summarized below.OPR 149, Part2 1 Notification of Failed Coatings on Fans: An industry notification of coatings on fans supplied to the nuclear industry was found to be applicable to the replacementc ontrol rod drive mechanism (CRDM) cooling fans that had just been installed in Unit 2 (and was operating at full power), as well as, the fans staged for installation with the new Unitl reactor vessel head.A knife/pull-off testo f the coatings on the Unit 1 replacement fans confirmed that the coatings were deficient.

Prior to installation in Unitl , the vendor Page 4 of 84 PDF created with pdfFactory Pro trial version www.Ddffactorv.com removed the deficientc oatings and recoated the fans in accordance with the requirements of ANSIN 101.4-1972, "Quality Assurance for Protective Coatings Applied to Nuclear Facilities." OPR 149 was developed to address the increase in quantity of unqualified coatings present in Unit 2. Revision 0 of this OPR was developed prior to the discovery of deficiencies in the calculations of sump screen head loss due to coating debris blockage (discussed above). The OPR concluded thatt he incremental increase in overall degraded coatings inventory by the addition of-160 ft 2 (Revision 1 later estimated the area at 173 ft 2) was minimal and would not impact the operability of the ECCS sump screens.Following the discovery of deficiencies in the sump screen head loss calculations, the OPR was revised. The revision evaluated the location of the CRDM fans and the potential for transport of fragments of the unqualified coatings to the ECCS sump screens. It concluded that such transportw ould not occur due to the high density of the fragments, their remote location from the sump,a nd the low transport velocities that would exist in containmenta fter a design basis LOCA.The OPR concluded that the sump screens were not challenged by the additional degraded coating inventory,a Ithough this condition is a nonconformance to the station license basis commitment thatq ualified coatings be used for activities comparable in scope and nature to those of the construction phase. Remediation of these degraded coatings is discussed in the response for Question 1.C below. No compensatory actions are associated with this OPR.OPR 161, ContainmentC oatings Not Maintained within Analyzed Limits: This evaluation was prompted by the discovery that the total quantity of unqualified and degraded coatings inside the containments was not bounded by the coatings transport and sump screen debris blockage analyses performed in 1998-1999.

These analyses had not been recognized as absolute limits,a nd the analyses had been informally re-performed after each coatings inspection using the increased inventory to check that the acceptable conclusions of the analyses were still valid.During the development of the OPR, a deficiency was identified in the underlying analyses from 1998-1999.

As a result, the head loss portion of the analyses was determined not to be valid.The deficiency was using the head loss correlation established in NUREG/CR-6224," Parametric Study of the Potential for BWR ECCS Strainer Blockage Due to LOCA Generated Debris," to estimate the head loss across the sump screens. The debris bed being analyzed consisted entirely of flatc hips or flakes of coatings,a nd was notp ostulated to contain any fibrous debris.However, the correlation published in NUREG/CR-6224 had been established from empirical testing using a mixed debris-type bed consisting of fine particulates and fibers. Use of the NUREG/CR-6224 correlation was, therefore, not appropriate.

In the absence of established methods and correlations, there Page 5 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com was not a valid, peer-reviewed approach for calculating the head drop across a debris bed consisting entirely of flat plates or chips. The analyses will be superseded by GL 2004-02 analyses and only the portions evaluating the potential transports f coatings debris will remain valid.To address the immediate concerns of operability, the results of the transport portions of the 1998-1999 analyses were scaled using less limiting,b ut bounding, values of sump depth and withdrawal rates to determine the critical areas of interest for degraded coatings.

In addition, recent testing results, documented in EPRIT echnical Report 1011753, "Design Basis Accident Testing of Pressurized Water Reactor Unqualified Original Equipment Manufacturer Coatings," September 2005, supported the deletion of unqualified coatings as challenges to the operability of the sump screens. It was concluded that there are nots ufficientd egraded coatings in proximity to the sump screens to challenge operability of the screens.This OPR contained a new and conservative assumption equating the area of degraded coatings thatc ould reach the sump screens to the area of the sump screens that would be blocked (e.g., one square foot of degraded coatings equals one square foot of blocked screen surface area). This is conservative because the visual inspections for degraded coatings intentionally round upwardly the areas of degradation observed and because flatp latelets (chips, flakes, etc.) would be expected to form a porous debris bed at least a few plates deep, rather than spread out evenly to form an impervious layer one platelet thick. The resulting debris bed would effectively block a considerably smaller screen surface area than the area of degraded coatings that created the debris.After the initial issuance of this OPR,c ontinuing internal reviews of the coatings inspection results from the prior refueling outages identified a previously unrecognized area of reported degraded coating in close proximity to the Unit 2 containment sump screens. An entry into containmentw as performed to inspect the area of concern. The reported degradation was confirmed,a nd a reactor shutdown was commenced in accordance with Technical Specification

3.0.3. During

the shutdown, the degraded coatings were reduced to an acceptable level, leaving only coatings thatw ere not accessible without the erection of scaffolding.

The shutdown was terminated and the unit was returned to full power operation.

The OPR was revised (Revision

1) to address degraded coatings remaining in proximity to the sump screens. The OPR concluded that these inaccessible remnants were located too far from the sump screens to present a challenge to them (i.e., would notb e transportable).

Other emergent concerns were also addressed in this revision of the OPR, including other containmentl atent debris such as tape, labels, and remnants of mineral wool used during the construction of the facility thata re still adhering to the bottom surface of overhead floor slabs, and thermal insulation.

Both revisions of the OPR concluded thatw hile the screens are operable, the increased quantity of unqualified or degraded coatings in containment constituted a nonconformance to the license basis as communicated in the Page 6 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com station response to GL 98-04. No compensatory measures were indicated, and resolution of the nonconformance will be achieved by completion of the GSI-191 project. That project will supersede the existing criteria for coatings, insulatione tc, and replace them with the design bases assumptions and analyses for the replacement sump screens. For further details of this OPR, please refer to the responses under Question 3 below, OPR 161, Revision 1, and Engineering Evaluation 2005-0024, Revision 1 (contained on CD-ROM as listed in Enclosure 3).OPR 162, Ability of Sump Screens to Pass Required Flow: The NRC prompted this evaluation when it was observed that the containment sump screens are in close proximity to the sump outlet valve disc. In the event that a small debris"collar" formed at the base of the sump screens, it would cause outlet flow to be channeled through a narrow annulus between the valve disc and the sump screen. The concern was that the resultant head loss could cause a loss of required NPSH to the RHR pumps.During development of the OPR, additional concerns (the potential for flashing of hots ump fluid just downstream of the annular constriction, and for air entrainment by vortexing) were also addressed.

The OPR evaluated the potential for both excessive head loss and flashing of the sump fluid. It concluded that head losses would remain acceptably low, provided that the outlet flow rate is limited to thata chievable by a non-degraded RHR pump delivering flow to only the reactor core. If the pump were to be aligned to discharge to both the core and a high head SI pump without throttling the total flow,e xcessive head losses could result. This 'piggyback" alignment would occur during operation to flush postulated concentrated boric acid from the vessel outletp lenum many hours after a DBA LOCA. A compensatory measure to limitRH R flow to prevent a loss of NPSH was therefore implemented.

The OPR also concluded thatf lashing would not occur, but that it was necessary to credit static containment

'overpressure" to arrive at this conclusion.

The 'overpressure" is due to the air present in the post-accident containment.

The pressure contribution of this air had been intentionally omitted from previous NPSH analyses as a conservative and bounding assumption.

Since crediting this "overpressure" was notc onsistent with the station's response to GL 97-041, this is considered to be a nonconformance with the license basis. Revision 1 of the OPR expanded and clarified the contents of Revision 0 to address additional questions posed by the NRC.Continued reviews in response to an inspector question about the potential for the containment sump outletv alves to gradually "drift" shut during long-term containment sump recirculation prompted Revision 2 of the OPR. The revision GL 97-04 was concerned with available NPSH as calculated by the customary two-pointm ethod, and did not identify concerns with flashing of hots ump fluid en-route between the sump and the pump impeller.

The potential for flashing at some intermediate points, and specifically when passing through a sump screen, was recognized as a result of the new guidance contained in the safety evaluation for NEI 04-07,'Pressurized Water Reactor Sump Performance Evaluation Methodology," dated December 2004.Page 7 of 84 PDF created with pdfFactory Pro trial version www.Ddffactorv.com evaluated how far the outlet valves could drift in the shut direction from fully open until unacceptable frictional head losses and/or flashing would occur. The results established that there would be adequate time to take remedial action, such as reopening the valve,b etween indications of valve drifting and the loss of the RHR pump suction source.All three revisions of the OPR concluded that while the ECCS system remained operable, the crediting of containment "overpressure" was a nonconformance to the license basis, and thata compensatory measure to procedurally limit the total sump outletf low is necessary to ensure thata dequate NPSH is available to the RHR pumps. Resolution will be achieved by completion of GSI-191 activities, and may require a license basis change to credit either containment"overpressure" or sump fluid subcooling to demonstrate that flashing at the outlet valve disc will not occur. The current revision of OPR 162 is contained on the enclosed CD-ROM.OPR 164, Wax Deposits on Unit 2 Containment Floor. During the Unit 2 at-power containment entry to inspect and remove suspected degraded coatings (discussed in connection with OPR 161 above), areas of dark deposits on the containment floor coatings were observed in the vicinity of the containment sump screens and on an upper elevation of containment.

These deposits had been previously documented in containmentc oatings inspections as remnants of floor "wax". Previous efforts in the mid-1 990s to remove the"wax" deposits throughout containment had been largely successful, however, there were still isolated areas of coatings that had not been removed after repeated attempts.OPR 164 addressed the presence of these remaining deposits, and established that they were not a wax, but rather, an acrylic co-polymer floor coating. The tenacious nature of the deposits, their limited extent (-40 ft 2),a nd their benign failure mode (i.e., into fine particulates that would pass through the sump screen perforations) contributed to the conclusion that their presence did not pose a challenge to the operability of the ECCS sump screens.While the OPR concluded that the Unit 2 sump screens would not be challenged by the presence of the remaining unqualified floor coatings, their existence is considered a non-compliance with the station license basis that will be resolved by completion of GSI-191 activities.

No compensatory measures are indicated.

Further removal of the tightly adherent remnants of acrylic floor sealer ("wax")will not be attempted because previous attempts have resulted in damage to the underlying qualified coatings and concrete.

Removal would pose a challenge to quality that is disproportionate to their continued presence.

The sizing of the replacement screens is taking into account these unqualified coatings as part of the design basis particulate debris loading.Page 8 of 84 PDF created with pdfFactory Pro trial version www.Ddffactorv.com OPR 170, Design Basis Leakage Detection Capability Defeated.

During continuing reviews of the design and license basis for the ECCS sumps and related systems, itw as found that an original design feature of the system had been defeated by later actions. In the original design, ECCS leakage originating in piping in the tendon gallery underneath the containment structure would collect in the gallery sump and be channeled through open pipe sleeves to the RHR pump room. This room has instrumented sumps thata larm in the control room in the event of a high level, and this would alert the operator to an abnormal conditions uch as excessive leakage in the ECCS system.Subsequents ite activities grouted the pipe sleeves closed. These grouted closures have since been found to be credited as limiting the intrusion of ground water into the RHR pump rooms,p roviding seismic supports for the RHR piping (in the case of Unitl ), and limiting potential flooding of the RHR pump room in the event of an RWST rupture.OPR 170 evaluated whether adequate indication of ECCS leakage in the tendon gallery remained despite the closing of the intended drain paths. It concluded that the safety-related containment sump level instrumentation provided ample indication of a loss of sump inventory caused by leakage before itc ould jeopardize the functioning of the ECCS system. The OPR also considered the potential dose consequences of postulated leakage and found them to be acceptably bounded as well. For further details and information, please refer to the responses to Question 5 below.OPR 170 concluded that removal of the leakage path from the tendon gallery to the RHR pump room did not jeopardize the operability of ECCS or supporting equipment, but that it did constitute a nonconformance with the design and license basis description of leakage detection capability.

Since there are other reliable means of leakage detection (i.e., the redundant and environmentally qualified containment sump level indications in the control room),n o compensatory measures were required.OPR 171, Safety Functions of Containment AccidentS ump Isolation Valves: Pursuant to NRC inspection activities and continued internal reviews,i t was determined that the containment sump outlet valves have an active function to shut to isolate a postulated system leak occurring downstream of the valves.Since this function had notb een explicitly identified previously in station inservice testing (IST) documentation, OPR 171 evaluated whether there was reasonable assurance that this function would be achieved.

Further details of the technical issues pertaining to this OPR can be found in the NMC responses to Questions 4 and 5 below.The OPR concluded that, based upon stroke testing performed incidental to the open stroke testing,r efueling frequency leakage testing of the downstream piping, and an initial review of environmental qualification of the supporting components (such as the position limits witches, solenoid pilot valves,a nd hydraulic power packs), the valves would perform the identified function to shut reliably in the event of a design basis event. However, the condition is nonconforming to the station's license basis because there is not sufficient Page 9 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com environmental qualification documentation for the shuts afety function and the testing protocol for this function are not complete.Quarterly stroke testing procedures for the valves have been revised to verify close-stroke capability.

No compensatory measures are necessary.

Additional corrective actions are to be taken as discussed in response to Question 1.C below.In summary, the six OPRs described above concluded that in each case a nonconformance to the license basis existed. However, in each case, the potentially affected systems, structures or components (SSCs) were also determined to be operable.B. Have you completed a 10 CFR 50.59 evaluation of compensatory measures (e.g., ECCS flow reduction) taken as part of your OPRs? If so, provide a copy of those evaluations.1 f not, please explain why?NMC Response: The only compensatory action directed was to limit the flows through an RHR pump operating on containments ump recirculation to 1560 gpm or less when operating an SI pump in 'piggyback." During safety injection, a single train of RHR discharging through its piping system and against a depressurized RCS has been analyzed to deliver<1582 gpm (there are slightv ariations from train to train and unit to unit due to differences in pipe routing).

This is more than adequate for decay heatr emoval (-200 gpm of boil-off at 20 minutes post-trip), even assuming that 50% of the flow spills to the containment prior to reaching the reactor vessel.However,w hen a parallel flow path is aligned from the RHR pump to both the RCS and an SI pump, the decrease in RHR pump discharge back pressure will result in a marked increase in RHR pump flow if no other actions are taken to limit it.T he procedural direction to limit the flow results in keeping the RHR pump within the analyzed acceptable condition of 1582 gpm while ensuring sufficient flow for decay heat removal.Please refer to the response to Question 4.A below for further details. The 10 CFR 50.59 screening of the change to the procedures (SCR 2005-0260) was completed and is included on the CD-ROM being provided per Enclosure 3.Page 10 of 84 PDF created with pdfFactory Pro trial version www.pdffactory.com C. Provide a detailed discussion including planned actions and schedule for resolution of any nonconformances with the current licensing basis or degraded conditions.

NMC Response: The following actions and schedule for resolution of nonconformances or degraded conditions is provided.

The actions and schedule are provided reflect the due dates that are listed and are being tracked to completion in the PBNP corrective action program. At the latest, these actions will be completed consistentw ith the existing NMC commitment to resolve GSI-191 by December 31, 2007.Specific Items to be Resolved External to GSI-191 Refueling Frequency Testing of SI-850 Valves: The procedures to stroke test the valves on a refueling frequency will be revised with appropriate acceptance criteria prior to the nextp erformance of each test during each unit's upcoming refueling outage.Sump Outlet Valve Position Indication Qualification:

The position indication limit switches for the SI-850 valves will be dedicated or upgraded to be able to withstand an anticipated harsh environmentd ue to integrated gamma dose prior to the end of the next refueling outage on each unit.Sump Outlet Valve Motive Power: The hydraulic power packages and positioning solenoid valves for the SI-850 valves will be dedicated or upgraded to be able to withstand an anticipated harsh environment due to integrated gamma dose prior to the end of the next refueling outage on each unit.Detection of Si System Leakage into the Tendon Gallery: Alternatives to the grouting that currently exists in the tendon gallery are being evaluated.

Resolution of tendon gallery grouting issues will be consistent with resolution of GSI-1 91 and will be completed by the end of the next refueling outage of each unit (Fall 2006 for Unit 2 and Spring 2007 for Unitl ).Programmatic Guidance for Monitoring Containment Sump Level: Post-accident, long-term programmatic guidance will be implemented by June 2006 to include explicitd irection for monitoring the containment accident sump level for adverse trends thatm ay indicate a leak of service water into containment (uncontrolled rise in sump level), or a leak of sump inventory out of containment (uncontrolled drop in containment sump level), and to investigate the condition accordingly.

Remediation of the Unit 2 CRDM Fan Coatings:

The non-conforming coatings on the Unit 2 CRDM fans will be removed or the fans replaced with ones that are either uncoated or coated with qualified coatings prior to the end of the fall 2006 refueling outage.Page 11 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com Specific Items of Concern to be Resolved Under GSI-1 91 NMC continues to pursue resolution of GSI-191 issues in accordance with GL 2004-02 requirements and will provide status updates to the Commission in accordance with the provisions of the GL.Control of Containment Coatings:

The design basis for the replacement sump screens defines the limits of unqualified and degraded coatings that may exist in containment and the location of those coatings.

Prior to the end of the next refueling outage on each unit, containment coatings will be removed, repaired, or restored to the extent necessary to be enveloped by this design basis.Subsequent refueling frequency coatings inspections will ensure the total inventory of coatings and other sources of particulate debris will remain bounded.Sump Screen Replacement:

Replacemento f the existing sump screens with the GSI-191 replacement screens will eliminate the potential for a 'debris collar" flow restriction.

Replacement of the sump screens will occur consistent with NMC's commitment to GL 2004-02, no later than December 2007.Crediting of Containment Overpressure:

Assuming no containment overpressure, there may be a potential for fluid flashing under the sump outlet valve discs,e ven after installation of the new strainers.

However, a minor"overpressure" would suppress such flashing.

Substantial overpressure would be available due to trapped air and non-condensibles inside the containment building.

Resolution of this issue will occur concurrentw ith resolution of GSI-191.2. Zone of Influence A. What is the zone of influence?

How was this determined?

What is the basis for this answer?NMC Response: Attachments 1 and 2 of Enclosure 2 contain graphical depictions of the Zone of Influence for each containment thata re used to assess operability.

The term "Zone of Influence" is defined in the Purpose/Objective section of calculations M-09334-345-RH.1 and M09334-431-RH.1: "The zone of influence is defined as the horizontal distance extending from sump screen projected onto the water surface into which failed coating debris would be transported to the sump screen by the flow of water rather than settling on the containment floor." Page 12 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com These two calculations were concerned with the potential for failed coatings interacting with the sump screens. The calculations also considered a second Zone of Influence due to particles sliding along containment floor. This extended zone encompasses the area around a screen where coatings debris would settle to the floor, and once on the floor of the containment, could be transported to the screen surface by sliding along it.While the term "Zone of Influence" was notu sed in the earlier 1989-1990 unit-specific evaluations of paint and insulation debris effects on containmente mergency sump performance, the methodology used to determine the quantity of debris that could be transported to and accumulate on the debris screens was comparable.

The resultw as a graphical depiction of a"Debris Transport Zone" in Figure 6.2-3 of the evaluations.

A term with the equivalent meaning of the "Zone of Influence" (as historically used) is "Zone of Transport". "Zone of Transport" denotes the region surrounding the sump screens where suspended debris would ultimately arrive at the surface of the screen by all modes of transportc ombined. This response will state "ZOI/ZOT" when describing this region of potential debris transport.

Upon resolution of issues related to GSI-1 91, the previous analyses will be obsolete and the terms "Zone of Influence" and "Zone of Transport" will be used consistent with their use in NEI 04-07, 'Pressurized Water Reactor Sump Performance Evaluation Methodology," and its associated NRC Safety Evaluation.

How the ZOl/ZOT was Determined

-The method for determining the ZOI/ZOT in calculations M-09334-345-RH.1 and M09334-431-RH.1 is described in the Methodology/Acceptance Criteria sections of those documents.

The methodology used for calculating the horizontal water velocities and coating transport mechanisms is based on NUREG/CR-2791, "Methodology for Evaluation of Insulation Debris Effects," September 1982.Conceptuallyt he settling velocity of a postulated coating fragment is determined using the coating density, the assumed characteristic dimensions of the coating fragment( establishing the drag coefficients),a nd the density and viscosity of the liquid that it is sinking through. The time it takes the postulated fragment to sink through a sump of given depth is then determined.

In the subject calculations,b oth the minimum and maximum sump depths were used to ensure bounding results were obtained.

Using hydraulic flow modeling methods, the flow field velocities for the areas surrounding the containment sump screens were determined.

Multiplying this radial flow velocity approaching the sump screens by the settling time for a debris fragmenti n a given depth of water results in the characteristic length of the ZOI/ZOT for direct impacto n the sump screen surface.The extended ZOI/ZOT that includes transport by sliding along the containment floor is a two-step process. In the extended ZOI/ZOT, the minimum bulk velocity to cause sliding of a postulated coating fragmenti s first determined, and the region surrounding the screens with flows at or above this velocity is Page 13 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com then established.

Particles reaching the floor within this region may be expected to transport to the base of the sump screens.After this sliding region has been established, the process of determining horizontal transport during settling, the same as was done for the direct screen impact, is repeated.

Adding the two characteristic lengths (one for sliding transport and one for settling transport) results in the final characteristic length of the ZOI/ZOT for sliding transport.

The calculation of the ZOI/ZOT was refined, where appropriate,t o differentiate between flows originating from different areas surrounding the sump screens.This was because the calculated horizontal velocities varied depending upon the flow channels being considered.

The results are illustrated in Figures 9,10, and 12 of calculation M-09334-345-RH.1, and Figures 7, 8, and 9 of calculation M-09334-431 -RH. 1.Recent Revisions to the ZOI/ZOT: In late 2005, the ZOI/ZOTs contained in the previous evaluations were re-reviewed in Engineering Evaluation 2005-0024.

It was recognized that the earlier evaluations had assumed lower water levels and higher flow rates than would exist under the current operating procedures and equipmentl imitations.

By taking a ratio of the maximum supportable sump flow rate to the flow rate assumed in the earlier evaluations, the size of the direct impact ZOI/ZOT was reduced accordingly.

The details of the reduction are presented in Engineering Evaluation 2005-0024, Revision 1, which is contained on the CD-ROM provided as Enclosure

3. That evaluation determined that the largest horizontal projection of the ZOI/ZOT in either unit is bounded by a maximum of 2.4' based on a flow channel in Unit 2.The worst-case ZOI/ZOT dimension for direct embedment on the screen surface determined in calculations M-09334-345-RH.1 and M-09334-431-RH.1 was 7.3'. This had been calculated to exist in Unit2 at a minimum flood level of 2.68' and a flow rate of 4,847 gpm. The corresponding calculated ZOI/ZOT for a maximum flood level of 6.18' was only 6.6'. This demonstrates a diminishing ZOI/ZOT size for this flow channel with increasing flood depth. After scaling to account for the actual expected lower flow rates (reflecting equipment limitations and the use of only a single train during sump recirculation), the characteristic size of this worst-case (Unit 2) ZOI/ZOT was reduced to 2.4'.Based on recently completed testr esults contained in EPRI Technical Report 1011753," Design Basis AccidentT esting of Pressurized Water Reactor Unqualified Original Equipment Manufacturer Coatings," September 2005, the ZOI/ZOT for sliding transport was eliminated.

ZOI/ZOT had been based on postulated low-density alkyd coatings specific gravity of 1.12, and that had been assumed to deteriorate to transportable chips, which could then block the screen perforations.

The recently completed EPRI testing demonstrated that disintegration products from such coatings would be small particulates that are notc apable of lodging in the screen perforations.

The effects of ingestion of these small particulates are addressed in the NMC response to Question 3.E below.Page 14 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com In summary, the current ZOI/ZOT of concern is based on the settling velocity of qualified (acceptable) epoxy coatings only. This ZOIUZOT is based on impingement of debris on the screen surface prior to settling on the sump floor.It also includes all locations where transportb y other credible mechanisms could result in the deposition of the coatings fragments at the surface of the water within the ZOI/ZOT. Examples are degraded epoxy coatings which are located on the containment liner plate directly above the containment sump ZOI/ZOT or containment spray water wash-down of the vertical liner plate in this region could result in the fragments being carried to the sump area adjacent to the screens. The worst-case characteristic size for the ZOI/ZOT is 2.4' from the sump screen surface. With additional refinement, this dimension could be further reduced by considering unit and scenario-specific parameters.

The NMC response to Questions 3.A and 3.D(2) discuss the coatings in this ZOI/ZOT.Attachments 1 and 2 of this enclosure contain maps of containment depicting the ZOI/ZOT on the elevations in the PBNP reactor containments.

The ZOI/ZOT depicted on El. 8' is limited to the area immediately surrounding the screens as discussed above. In addition, it has been the practice to include an arc of the containment liner adjacentt o the sump screens and extending all the way to the containmentd ome as also being within the ZOI/ZOT. Piping and components in proximity to this arc have also been considered within the ZOI/ZOT unless otherwise evaluated.

These inclusions were based on postulated wash-down of degraded or unqualified coatings in these areas reaching the screen surface during containment spray operation.

This was applicable when the ZOI/ZOT was large enough to extend to the containment liner wall. Although the ZOI/ZOT has contracted (as discussed above),t he vertical extensions of the ZOI/ZOT are retained due to the turbulence of the pool adjacent to the liner caused by the sheeting and cascading of water coming down the vertical liner plate during containments pray operation.

The arcs associated with the vertical extensions of the ZOI/ZOTs are also depicted at each elevation of the containment on the maps provided.3. PotentialB lockage of the Sump/Long-term Cooling A. Containment Coatings (1) How much (percentagea rea, and volume) of the coatings will fail? Include the location of the failed coatings, the type of coating, and qualification level of the coatings.

What is the basis for this answer?NMC Response: There are two general types of coatings thata re assumed to fail and be released to containment during or after a design basis loss of coolant (LOCA) event. These are unqualified coatings, and coatings that are qualified (Acceptable) but have become degraded by means of de-bonding or delaminating.

The following tables summarize the Page 15 of 84 PDF created with pdfFactory Pro trial version www.pdffactory.com quantity of each type located in each of the two containments based on inspections performed during the last refueling outage on each unit.Percent of Total Total Area Total Volume Unitl(It 2) Coatings (ft 3)Area (%)Unqualified Coatings 19,747 5.6 3.5 Acceptable but De-bonding/

996 0.028 1.6 Delaminating Coatings*Percent of Total Total Area Total Volume Unit 2 (It 2) Coatings (ft 3)Area (%)Unqualified Coatings 21,826 6.2 3.9 Acceptable but De-bonding/

3,940* 1.1 6.2 Delaminating Coatings*

II*An additional

=173 ft 2 of degraded coatings were subsequently identified as a result of a 10 CFR 21 notification.

Thatn otification dealt with improperly applied coatings on the recently replaced control rod drive mechanism (CRDM) fan housings.

It was determined that this additional inventory was insignificant and was located outside of a ZOI/ZZOT of concern. The fans with deficient coatings designated for installation in Unit 1 were replaced with fans that had fully qualified coatings prior to actual installation.

During the most recent refueling outage (Unit 1), the coatings inspection differentiated between de-bonding/delaminating coatings and those that were degraded in other benign modes,s uch as mechanical abrasion or impact damage,c racking butt ightly adherent, etc. This distinction had not been previously applied, and results in the Unit2 inventory being substantially larger. The quantity of degraded coatings in Unit 2 is inferred from the textual descriptions contained in the inspection reports and is believed to be conservative because it does not differentiate between types of degradation Unqualified coatings are widely distributed throughout the containments in relatively small quantities, but the main sources are attributable to a few discrete components:( 1) Polar crane (-5,500 ft 2); (2) Polar crane rail girder (-4,950 ft 2);( 3) Manipulator crane (1,500 ft );( 4) Reactor coolant pump motors (600 ft 2).Page 16 of 84 PDF created with pdfFactory Pro trial version www.pdffactory.com Attachments 1 and 2 of this enclosure provide a detailed listing of the delaminating "Acceptable" coatings in each containment, followed by graphical depictions of their approximate locations.

Acceptable coatings are coatings that include coating systems which have been reviewed for suitability for application inside containments, and there is reasonable assurance that the coatings will not detach under normal or accident conditions.

At PBNP, the coating systems specified for use on major structures during original construction were tested and qualified for the design basis accident (DBA) environment by WCAP-7198-L, "Topical Report- Evaluation of Protective Coatings for Use in Reactor Containment," dated April 1, 1968.Unqualified coatings are those coatings do notm eet the above criteria.

These are mostly original equipment manufacturer (OEM)applied alkyd (oil) based coatings.

A coating lacking sufficient documentation to establish it as a "Qualified" (Acceptable) coating is classified as unqualified.

Unqualified coatings are assumed to all be alkyd-based and 100% of them are assumed to fail. As discussed in the NMC response to Question 3.A.(2) below, the failure products of these coatings are benign,d o not challenge the functioning of the ECCS sump screens, and are not represented by a detailed listing or graphical depiction of location.

Additionally, only qualified (Acceptable) coatings that exhibit delamination or de-bonding are assumed to fail.The total coverage of coatings is approximately 353,100 ft2 per containment.

This value was used as the basis for determining area percentages of failed coatings.

The total volume of coatings is determined by multiplying the thickness by the area (see the NMC response to Question 3.A.(2) below).(2) What are the physical characteristics of the failed coatings (particle size, thickness, and specific gravity)?

What is the basis for this answer?NMC Response: Size: Unqualified coatings are assumed to fail to minute particles bounded by 1128 microns or less in characteristic dimension.

This is based on EPRI Technical Report 1011753," Design Basis Accident Testing of Pressurized Water Reactor Unqualified Original Equipment Manufacturer Coatings." This recently issued report demonstrates that a broad range of coatings, including epoxies and alkyds, when they deteriorate, do so in the form of fine particulates.

Acceptable coatings that fail (de-bond and become available for transport) are assumed to be flat discs 1/8" in diameter.

This assumption is based on having the smallestp ossible fragment that Page 17 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com could physically lodge in or on the 1/8" screen perforations.

By minimizing the size the transportability of the fragments is maximized.

The flat disc also maximizes the drag coefficients uch thatb oth the settling velocity and the velocity of water necessary to transport horizontally across a surface are minimized.

Thickness:

The thickness of coatings varies by application and location.

The values used in various analyses depend upon the purpose of the analysis (i.e., whether it is evaluating the heat transfer to containment heats inks to calculate the pressure and temperature response to a LOCA, whether it is evaluating the quantity of debris that may be generated,e tc.). A sampling of existing coatings thickness was used to establish a conservative value for the debris generation analyses of interest.The Dry Film Thickness (DFT) of unqualified alkyd coatings was measured to be between 0.0003 and 0.0038", with an average of 0.00212".

This value is appropriate when estimating the total volume of such coatings.

The DFT of acceptable (epoxy) coatings was measured between 0.0045" and 0.0187" with an average of 0.01 16".When evaluating the transportability of these coatings, a conservatively low value is appropriate for determining transportability (0.005" was used in mostc ases,a Ithough 0.015" was used where justified for the concrete floor coatings in Unit 1).Based on the above, for the purposes of estimating the total volume of epoxy coatings, a bounding high value of 0.019" was used for dry film thickness.

Specific Gravity: The specific gravity of unqualified (alkyd) coatings used in the previous transport analysis was 1.12. However, this value is not relevant since PBNP is assuming that these coatings fail to fine particulates and are highly transportable.

The specific gravity of the acceptable (qualified) coatings used at PBNP is bounded by a low value of 1.6. This reflects the specific gravity of the Phenoline 305 coatings used on concrete surfaces inside containment

(-85,000 ft 2 per containment).

The other acceptable coating systems consisto f Dimetcote 6 primer (specific gravity 3.2), and Amercote 66 (specific gravity of 2.6). These two higher density coatings were used on the major steel surfaces of containments uch as the containment liner and structural steel (-268,000 ft 2 per containment).

Page 18 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com Summary Of Failed Coating Particle Characteristics Failed Particle Characteristic Coatin TypeSpecific Coating Type Size Thickness Gravity Qualified

/ 1/8" diameter 0.019" 1.6 minimum Acceptable (Epoxy) discs 0.1_16miiu Unqualified (Alkyd) <1128 microns 0.00212" 1.12 (3) Will the failed coatings be transported, including during the blow down phase of the event, to the sump? What is the basis for this answer?NMC Response: Unqualified Coatings:

The disintegration products of unqualified coatings are conservatively assumed to be 100% transportable to the sump owing to their minute sizes.Acceptable butD egraded (Delaminating)

Coatings:

Acceptable coatings that fail into chips or flakes large enough to be a challenge to the screens are also too dense to be readily transported by the low velocity flows that would exist during sump recirculation.

As a result, provided delaminating coatings are located outside of the small ZOI/ZOT, they would not be transported to the sump screens. The deep, flat-bottomed pool with relatively wide, open flow channels, and a low withdrawal rate leads to very low flow velocities (less than 0.1 fps) thata re not conducive to the transport of negatively buoyant debris. As a result, Acceptable coatings debris large enough to pose a challenge to the ECCS sump screens will not transport to the screens.Supporting Details: The PBNP "sumps" are not conducive to the transports f debris. The sumps are not depressed sumps, but rather, comprise the entire El.8 ' of the containment.

This floor is nominally flat with the sump outlet pipes dropping vertically downward from the floor level. The opening of the 10" outlet pipes (one per train) is flush with the floor. As can be seen in Attachments 1 and 2 of this enclosure, the floor plan of El. 8' of containment is relatively open and unobstructed.

This minimizes high velocity channels and choke points,a nd therefore,m inimizes re-suspension of settled debris.The absence of a depressed sump precludes a "trap" that could collect debris during the energetic blowdown phase of a transient.

Switchover to the containment sump is directed when indicated refueling water storage tank (RWST) level is 34% or less. This corresponds to an actual level in the containments ump of-42" above the containment floor. This figure discounts a Page 19 of 84 PDF created with pdfFactory Pro trial version www.pdffactory.com contribution from the breached reactor coolant system (RCS) and the safety injection (SI) accumulators.

After initial switchover to recirculation, the containment sump continues to fill as the RWST is depleted using the containment spray system. The final level when spray is terminated at 9% indicated RWST level, the depth of water in the containment sump is -60".During containments ump recirculation,o nly a single train is placed in operation.

The hydraulic analysis of the SI and RHR system shows that system piping friction limits total flow to s1,582 gpm in this mode of operation.

Later, if concurrent upper plenum injection or "core deluge" (the nominal recirculation flow path supplied by an RHR pump) and cold leg injection (supplied by an Sip ump operated in series with an RHR pump) is desired for prevention of boron concentration and precipitation, the total flow is procedurally limited to 1,560 gpm indicated flow.How much (percentage,v olume,p article size) of the coatings will be transported including during the blow down phase of the event? What is the basis for the answer?NMC Response: Transportd urine Blowdown:

Degraded qualified coatings that fail by delamination are mostl ikely to do so as flakes or chips during the blowdown phase of a postulated transient.

However, in that case, they would come to rest on the containment floor before sump recirculation is initiated, would be sequestered,a nd not available for transport to the surface of the sump screens when sump recirculation is initiated.

This is because the horizontal fluid velocities during sump recirculation (less than 0.1 fps) are less than thatn eeded to transport settled debris (0.2 fps). This would be true regardless what mechanism may be postulated to generate the coatings fragments.

As a result, transport of coating chips or flakes to the screens is not considered during the blowdown phase of the transient.

The current approach reserves the full inventory of degraded (delaminating) coatings for the recirculation phase. During this phase,a moving fluid field exists that could transport the coatings debris to the screen while they were sinking if the debris landed in close proximity to the screen. The assumption of no transport during the blowdown phase is more conservative than assuming otherwise.

Transporto f fines from erosion of qualified coatings and disintegration of unqualified coatings is stipulated.

These, however, are incapable of producing sump screen blockage due to their small size. These fines are therefore not considered analytically when evaluating sump screen blockage.Page 20 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com TransportD urine Recirculation

-Acceptable

("Qualified")

Coatings: The 1/8" flakes or chips of acceptable

("qualified")

coatings discussed in the NMC response to Question 3.A.(2) above are too dense to be transported horizontally across the nominally level floor of containmentb y the low velocity prevailing flows (less than 0.1 fps)during sump recirculation.

Therefore, unless degraded coating fragments are deposited on the surface of the pool at or within the ZOI/ZOT (as described in the NMC response to Question 2.A above), they will not be transported to the sump screens during sump recirculation.

Quantitv of Acceptable Coatings that Transport to the Sump Screens Unit 1 Unit 2 Area of coverage (ft 2) 0 0 Percentage of all coatings 0% 0%Volume (ft 3) 0 0 Particulate Size (inches) 0.125 0.125 TransportD urine Recirculation

-Unqualified Coatings:

It is assumed that 100% of the unqualified (alkyd) coatings will be transported to the sump area. This is because the fine particulate nature of the disintegrated coatings renders them highly transportable.

The area of unqualified coatings in containment was tabulated in the NMC response to Question 3.A.(1) above and is provided below with the percentages and volumes that they represent.

The range in particle size is from EPRIT echnical Reportl 011753, "Design Basis Accident Testing of Pressurized Water Reactor Unqualified Original EquipmentM anufacturer Coatings." Quantity of Unqualified Coatings thatT ransport to the Sump Screens Unit 1 Unit 2 Area of coverage (ft 2) 19,747 21,826 Percentage of all coatings 5.6% 6.2%Volume (ft 3) 3.5 3.9 Particulate Size (microns) 5 -1128 5 -1128 Page 21 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com How much of the degraded qualified and unqualified coatings are on the containment floor (both pre-existing and event generated) in the zone of influence around the sump, and how much of those will be transported to the sump? What is the basis for this answer?NMC Response: Acceptable (butD egraded) Coatings:

Walkdowns of the floor areas on El.8 'of containment determined that large portions have some extent of mechanical damage such as abrasions or "dings." There was no delaminating noted. Therefore,n o pre-existing debris from otherwise acceptable coatings is anticipated on the floor in the area immediately adjacent to the sump screens. As discussed in the NMC response to Question 2.A above, the ZOI/ZOT atP BNP is relevant only for coating debris that may be settling through a moving water column. This is because itw as determined thatc oatings debris large enough to pose a challenge to the sump screens is nott ransportable across the floor of containment.

The maximum flow velocities on containmentE I 8' (the entire "sump")are less than 0.1 fps. This is below the 0.2 fps threshold necessary to transport debris across the floor. As flow converges near a containments ump screen, itw ill accelerate due to the decreasing flow area normal to the direction of flow. Using a minimal sump depth of 3.2' (from Calculation 2000-0044, Revision 3; assumes a minimum RWST draw-down, no contribution from spilled RCS inventory, no contribution from Sla ccumulators, and no expansion due to thermal heating of the sump contents), and a flow rate of 1600 gpm (3.56 ft 3/sec) flowing toward the cylindrical screen, the 0.2 fps threshold would be a cylinder of 1.8' in diameter.

This is smaller than the minor dimension of the trash rack covering the screens (as seen in Figure 4.1-2 of the Gibbs & Hill reports, the trash rack covering the screens is 2' wide and 5' long). Therefore,p articles large enough to lodge on the screen surface and originating outside of the trash rack are not subject to transport to the sump screens, even at minimum sump depth.Since there is no ZOI/ZOT for horizontal transport of acceptable coatings debris large enough to challenge the sump screens, there are no acceptable coatings located on the floor within the ZOI/ZOT.Unqualified Coatings:

Unqualified coatings are expected to disintegrate into fines that would not pose a challenge to the functioning of the ECCS screens. They are,h owever, assumed to be 100% transportable to (and through) the sump screens. The total quantity of potential debris was provided in the NMC response to Question 3.A.(1) above. See the response to Question 3.B below for the basis of noth aving fibrous debris loading on the screens (no thin bed effect).Page 22 of 84 PDF created with pdfFactory Pro trial version wwvw.pdffactorv.com

4) What percentage of the sump screen will be blocked by failed coatings or by coatings in combination with other material?What is the basis for this answer?NMC Response: Since the only coatings that are transportable to the sump screens are those that are smaller than the screen perforation size,n o blockage of the sump screens due to coatings is anticipated (0%blockage per analysis).

B. Containment Insulation:

(1) How much (percentage,v olumet ype and size) of the insulation will fail, including during the blow down phase of the event?What is the basis for this answer?NMC Response:

The following tables summarize the quantities of insulation debris generated from the limiting break locations in each unit: Unit 1 Insulation Debris Insulation Type Area or Volume of Debris Reflective Metallic foils (ft 2) 19,438 Asbestos & Calcium Silicate* (ft 3) 222 Encapsulated Fiberglass (ft 3) 95 Temp-MatB lankets (ft 3) 67 Encapsulated Mineral Wool (ft 3) 12 Unit 2 Insulation Debris Insulation Type Area or Volume of Debris Reflective Metallic foils (ft 2) 8,862 Asbestos & Calcium Silicate* (ft 3) 301 Encapsulated Fiberglass (ft 3) 95 Temp-MatB lankets (ft 3) 67 Encapsulated Mineral Wool (ft 3) 12*NRC Information Notices IN 2005-26 and IN 2005-26a communicated a concern with the presence ofc alcium silicate (CalSil) insulation in containments that use tri-sodium phosphate (TSP) as a pH buffer. TSP is not used at PBNP. The concern is that the relatively insoluble compound of calcium phosphate will precipitate if there is an appreciable quantity of dissolved Ca 2 and PO43 ions present in the post-accident solution.

It has been postulated that CalSil, while a relatively inert covalent compound, could still contribute significant concentrations of Ca 2 into a phosphate-rich sump from the resulting in-clogging (or 'blinding")

of a pre-existentf ibrous debris bed. NMC is aware of these concerns, and has been participating in industry efforts to further quantify these and other potential "chemical effects." Page 23 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com A TSP buffer is not used at PBNP. At PBNP, a sodium hydroxide additive (NaOH) to the containment spray buffers the sump pH. To date, sodium hydroxide buffers have exhibited some potential for the formation of sodium aluminum silicate and aluminum oxyhydroxide (AIOOH) precipitates.

This research is being incorporated in the GSI-191 project, as applicable.

In 1989-1990, prior to the debris generation analyses performed in support of the continuing GSI-191 resolution effort, Gibbs & Hill performed debris generation and transport analyses for PBNP (Enclosure 3 CD-ROM). These analyses followed the general methodology of NUREG/CR-2791, NUREG/CR-3616,a nd NUREG-0897,R evision 1. The analyses form the current design bases for insulation debris transport.

The analyses evaluated the generation of debris from five categories of insulation installed in the PBNP containments and on or in close proximity to the RCS piping:* Reflective Metallic* Asbestos and Calcium Silicate Blocks (with stainless steel jackets)* Encapsulated Fiberglass

• Temp-mat Blankets* Encapsulated Mineral Wool The evaluated mechanisms for debris generation were:* Jet Impingement (7-pipe diameter zone of destruction)

• Pipe Whip (all insulation between the break and the plastic hinge)* Pipe Impact( 5 fabricated lengths of installed insulation on the impacted pipe)Although the PBNP licensing basis does not require the consideration of the dynamic effects of a LOCA (modified GDC-4 per 10 CFR 50 Appendix A GDC-4),p ipe whip and pipe impact were included in the evaluations.

The limiting break in each containment was determined to be in the"B" steam generator cubicle because of its proximity to the sump screens. A hot leg break was found to be the worst case.Page 24 of 84 PDF created with pdfFactory Pro trial version www.pdffactory.com (2) What are the physical characteristics of the failed insulation (particle size, thickness, and specific gravity)?

What is the basis for this answer?NMC Response: The characterization and evaluation of debris from failed insulation was performed in the Gibbs & Hill reports (Enclosure 3 CD-ROM).The following are excerpts from Section 7.4.3 of those analyses.Owing to the non-transportability of most of the debris considered, a more detailed characterization of the failed insulation was not performed.

• "Reflective Metallic Insulation (From Alden Research Laboratory test data reported in NUREG-0897 Revision 1 and NUREG/CR-3616)

..."* "Single sheets of thin stainless steel materials (such as the 0.00025" -0.004" thick foils used within reflective metallic insulation units)..."* 'As fabricated reflective metallic insulation units..."* "Outer covers (0.037" thick)..."* "Inner covers..." (no thickness cited, but apparently comparable to the outer covers)* "End covers..." (no thickness cited, but apparently comparable to the other covers)Asbestos, Mineral Wool, and Calcium Silicate Blocks* "...hard, cast material like mortar with a minimum specific gravity which is greater than water. This material is covered with stainless steel jacketing.

If the jacketing is destroyed by jet forces and the block material is also damaged, this material will break into large chunks and fall to the floor..." Encapsulated Fiberglass and Temp-MatB lankets: "...Type "E" glass... [in] jacketing"* "...[intact]

dislodged panels..." as well as "loose insulation":

• "...type "E" glass... density...

11 lb/cu ft....""Unlike conventional fiberglass, Type "E" glass is a woven material, notr eadily subject to ripping and shredding...

not anticipated that the Type "E" glass material would disintegrate in such a manner as to allow transportation of glass fibers to the sump screens."* "...Absorbs water,p articularly hot water, and sinks rapidly (from 20 seconds to 30 seconds in 120"F water)..." Page 25 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com Encapsulated Mineral Wool:* "...encapsulated in welded stainless steel jackets" or"...encapsulated in welded and riveted stainless steel jackets"* "In the event of a pipe break... highly unlikely [to be] removed from the jacketing..."* "Although intact mineral wool mats could be lighter than water, the fragmented fibers have a specific gravity greater than 2..." (3) Will the failed insulation be transported, including during the blow down phase of the event,t o the sump? What is the basis for this answer?NMC Response: No significantq uantity of failed insulation is expected to be transported to the sump screens, including during the blowdown phase of an event.Blowdown Transport:

Transportation of debris during the blowdown phase of a LOCA event is acknowledged.

This mode of transport has notb een analyzed in detail,e xcept within the contexto f the continuing effort to resolve GSI-191 concerns.The chaotic relocation of such debris during blowdown would tend to be a general dispersal away from the break location, but would not tend to deposit debris preferentially upon the trash rack nor fine screens located within it (for a depiction of the sump, trash rack, and screen configurations, please refer to Figures 4.1-1 through 4.1-3 of the Gibbs & Hill reports included on CD-ROM as parto f Enclosure 3).Since the "sump" is the entire El. 8' of containment, there would be no tendency to trap and retain transitory debris passing through the vicinity of the sump screens as could be the case for screens located in a depressed sump.Further, the debris would subsequently be covered by the rising water level in the containment, be washed down into the deepening pool by continued spray or break flow (and subsequently sink), or remain where deposited on higher elevations.

In any case, they would be sequestered and would not be available for further transport once the recirculation flow was initiated.

The current design basis analyses assumed a deposition of debris generated on the floor of the lowest level beneath the loop compartments.

This concentrated the maximum quantity of debris in the pool ata location close to the sump screens (in this case, the limiting B loop rupture discussed in the response to Question 3.B.1 above). No deduction was taken for debris blown up to the higher Page 26 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com elevations of containment or held up on the bar grate platforms underlying most of the RCS loop compartments.

Subsequent Transport:

During the time period between the blowdown transient and the initiation of sump recirculation (while the containment sump fills), there is sufficient time for all debris generated to become thoroughly wetted and sink to the bottom of the containment sump. Subsequent transport to the sump screens would require horizontal transport, and the flow field necessary to cause such transport could not exist until sump recirculation is initiated.

The minimum velocity required to transport submerged insulation debris is 0.2 ft/sec as established in NUREG-0897, Revision 1. The drag force of a submerged object in a freely moving fluid is proportional to the square of the velocity,a nd the velocity in containment is less than 0.1 fps (see the response to Question 3.A.3 above). Since this is less than half of the empirically observed threshold for transporting sunken objects, there is a drag force margin of at least four (4) between the forces that could exist under post-DBA recirculation and the force necessary to transport the postulated debris. The margin is even higher when it is recognized that the fluid along the floor of the containment is not a freely flowing fluid, butr ather has a significant stagnant or slower moving boundary layer thatw ill tend to trap fines and fibers small enough to be fully enveloped in it.Based on the above considerations,n one of the insulation debris that may be generated is expected to be transported to the existing sump screens.How much (percentage, volume, particle size) of the insulation will be transported, including during the blow down phase of the event? What is the basis for this answer?NMC Response: As discussed in the previous response, no insulation is expected to be transported to the sump screens, including during the blow down phase of an event.Page 27 of 84 PDF created with pdfFactory Pro trial version www.pdffactory.com (4) What percentage of the sump screen will be blocked by failed insulation or by insulation in combination with other material?What is the basis for this answer?NMC Response: No blockage of the screens is anticipated.

As was discussed in response to Question 3.A above, and will be discussed in the responses to Questions 3.C and 3.D below, analytical treatment consistent with the current license bases of debris other than insulation, found none that are transportable to the containment sump screens. Therefore,n o aggregate effect is indicated.

In addition, since none of the insulation debris is transportable to the sump screens, no blockage of the sump screens due to insulation is expected.C. Containment Debris: (1) How much (volume, type and size) containment debris will be transported to the sump? What will happen during the blowdown phase? What is the basis for this answer?NMC Response: No containment debris is expected to be transported to the sump, including during the blow down phase of a postulated transient.T he approach taken to determine transportability of containment debris is the same as that for debris originating from coatings and insulation.

The debris sources specifically evaluated are tape and adhesive labels known to reside or suspected to remain in small quantities in the containmentb uildings.

This type of debris would pose the greatest potential of both transport (due to relatively low density and high surface area) and screen blockage (due to potential for blocking a significantf raction of the screen surface with an impervious membrane).

The tape widely used for various purposes during refueling outages is a 2" wide fabric reinforced tape commonly referred to as "Duct Tape".Common experience indicates that the adhesive of this tape is thermoplastic, and remnants that may be inadvertently left in containmenta fter an outage cannot be expected to remain adherent under accident conditions.

Additionally, an undetermined quantity of conduit marking labels and striped reflective tape remain in each of the containments.

Though not tested, it is expected that the adhesive on these items would similarly fail under accident conditions.

The specific gravity of samples of these tapes and labels were measured under ambientc onditions, resulting in a specific gravity measurementr eferenced to room temperature water. However, the density of sump water early in an accident sequence would be lower Page 28 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com and the specific gravity of the debris correspondingly higher. This conservatively maximized the analyzed potential for horizontal transport.

This is because the frictional forces on debris from contact with the containment floor (those that tend to retard or prevent flow-induced transport) are proportional to the negative buoyancy of the debris. The average specific gravities measured ranged from a low of 1.1 to a high of 1.3.Since the potential debris sources tested have a specific gravity greater than 1.05, they are not expected to be subject to horizontal transporta cross the floor of containment with the analyzed flow velocities of less than 0.1 fps. This is based on the guidance provided in RG 1.82, Revision 1 that indicates a velocity of 0.2 fps or greater is required to transport debris of this density.Transportation of debris during a postulated blowdown event would be inevitable.

The distribution of transported debris would be expected to be a general dispersal outward from the break location.This condition has not been analyzed in detail prior to the development of analyses supporting the continuing effort to resolve GSI-191 concerns.The evaluation performed to assess current operability (Engineering Evaluation 2005-0024, Revision 1 and OPR 161,R evision 1 assumed a non-specific deposition of debris on the floor of the lowest level of containment.

This is consistent with the design basis analyses previously performed for other debris types that are provided in Enclosure

3. No deduction was taken for debris blown up to the higher elevations of containment, or debris sequestered at other locations in containment.

During a postulated blowdown transient, labels and tape that may reside in the containmentc ould be relocated.

The chaotic relocation of such debris during blow down would tend to be a general dispersal away from the break location, and would not tend to depositd ebris preferentially upon the trash rack,m uch less the fine screens located within it. Since the "sump" is the entire El. 8' of containment, there would be no tendency to trap and retain transitory debris passing through the vicinity of the sump screens as could be the case for screens located in a depressed sump.Further, the debris would subsequently be covered by the rising water level in the containment, be washed down into the deepening pool by continued spray or break flow (and subsequently sink), or remain where deposited on higher elevations.

Thus, the debris would be sequestered and would not be available for further transport once the recirculation flow was initiated.

Other debris types thatc ould be postulated in the category of acontainmentd ebris" are latent dust and dirt, "tramp" (loose Page 29 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com individual) fibers, and particles resulting from the erosion of concrete during the blow down phase. Consideration of these debris types is currently outside of the PBNP licensing bases, but they are being included in the analyses necessary to resolve issues related to GSI-191. In the interim, the above evaluations of insulation and coatings debris transport provide reasonable assurance that the probability of transport of such postulated debris is very low. The same reasoning and evaluations methods used in considering those debris types are applied to miscellaneous fines below: Dust, dirt, and concrete erosion products would, by their nature, either be very fine and capable of passing through the screens unimpeded, or if sufficiently large, would be of a density too high to be transportable.

Mark's Handbook for Mechanical Engineers tabulates typical specific gravities for concrete (2.2-2.4), dry sand and packed gravel (1.6-1.9), and damp clay (1.7). These are considerably greater than the 1.05 minimum threshold for transport in a 0.2 fps fluid field established in response to Question 3.a above.This indicates that such particles will sink, and will not transport if already on the floor of containment at the time recirculation initiates.

Further, published studies of the transports f solids by moving fluids have demonstrated that fines transportable at velocities of 0.2 fps and lower are on the order of 1 mm (0.04") or less in size. As such, they would be too small to lodge on the strainer surface, and would pass unimpeded through the 1/8" perforations.

Loose clumps or individual fibers could be postulated to originate from fibrous insulation and would be expected to behave as described in the response to Question 3.B above.T his type of insulation debris would be sequestered on the floor of containment after having been wetted out. The source of this type of debris could be from clothing worn in containment( in which case itw ould be trace amounts whose effects would be too small to quantify), or from filter material residing in containment.

The only filter materials in the various containment ventilation systems are enclosed in plenums of the containment cleanup system.The plenums are located on or above the El. 66' refueling floor,a nd are located away from LOCA zones of destruction.

As a result, this material (or loose fibers originating from it) is nots ubject to transport during either the blow down or wash-down phases of a postulated event.Filters that may be brought in to support outage activities,s uch as high efficiency particulate filters, are required to be removed prior to returning the unit to operation.

This is assured by the containment closeout inspection,w hich requires inspectors to enter accessible containmenta reas and to ensure the area is free of tools, equipment, dirt or debris accumulation and/or materials that could inhibit Sump "B" recirculation.

Page 30 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com Based on the above considerations, transport of containment debris to the sump screens is not expected.(2) What are the physical characteristics of the debris (size, shape, thickness, and specific gravity)?

What is the basis for this answer?NMC Response: The physical characteristics of possible debris were described in the response to Question 3.C.1 above.(3) What percentage of the sump screen will be blocked by debris or by debris in combination with other material?

What is the basis for this answer?NMC Response: As was discussed in the NMC responses to Questions 3.A and 3.B, analytical treatmentc onsistent with the currentP BNP licensing bases for debris,o ther than containment debris, found none that are transportable to the containments ump screens. Therefore, an aggregate effect is not indicated.

In accordance with the analysis, since none of the containmentd ebris is transportable to the sump screens,b lockage of the sump screens due to miscellaneous containmentd ebris is not anticipated.

OPR 162 demonstrates an additional margin of safety by assuming, consistent with the original design of the screens, that5 0% of the submerged area is nota vailable due to blockage.

Although we expect no blockage, in OPR 162, consistentw ith the original licensing basis, we conservatively assume 50% screen blockage.D. Sump Blockage: (1) What are the safety functions of the emergency core cooling system (ECCS) sump? What is the basis for this answer?NMC Response: The ECCS sump (a) serves as the suction source for the RHR pumps during the recirculation phase of a LOCA;a nd (b) precludes the passage of particulate debris greater than 1/8" in diameter to downstream components, such as the RHR pumps and reactor core.The first function ensures a continued source of water for core cooling during the immediate and long-term post-LOCA recirculation phase. In fulfilling this first function, the sump serves as a collection point for spilled coolant, injected water, and containments pray run-off. The sump also ensures that excessive air entrainment does not occur, and that frictional head losses through the sump structure Page 31 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com are low enough thata dequate net positive suction head (NPSH) to the RHR pumps is assured.The second function is to ensure that the functioning of critical downstream components is notj eopardized by debris suspended in the recirculation flow stream. The establishment of the 1/8" size was originally predicated on the 3/8" diameter containment spray system nozzles. While the spray system is not required to function during recirculation under the current license bases, retention of the maximum particulate debris size is appropriate to ensure operation of other downstream components (the NMC response to Question 3.E provides further discussion of this aspect).(2) What percentage of the sump screen will be blocked by coatings, insulation, and debris? What is the basis for this answer?NMC Response: Blockage from such debris is not anticipated.

As discussed in the responses to Questions 3.A, 3.B, and 3.C above, the characteristics of the debris type postulated and the very low fluid velocities preclude the transport of such debris to the sump screens. The only potential debris source thatc ould pose a challenge to the screens are acceptable coatings thath ave degraded by delaminating or de-bonding. If present within a very limited ZOI/ZOT immediately surrounding the sump screens, the chips or flakes thats uch coatings could shed would be available to embed on the screen surface. The most recently completed coatings program inspections, together with Engineering Evaluation 2005-0024, show such degraded coatings currently do not exist in the area of interest.

Without a viable transportm echanism, the sump screens would remain unblocked by the postulated debris. Note that OPR 162 uses a license basis 50%screen blockage.Page 32 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com (3) What percentage of the sump screen is required to be unblocked (or, what head loss can be sustained) to fulfill its safety functions?

What is the basis for this answer?NMC Response: The head loss that can be sustained is 1.6' (19"). This is limited by the potential for direct air ingestion due to the partially submerged screens. The minimum screen submergence at switchover to recirculation is 3.22'. The average submergence is 1.6'.As the containment continues to fill following a DBA, this minimum sustainable head loss likewise increases.

With an expected final containments ump level of 60", the average screen submergence increases to 30".The above directr esult is complicated by the recognition thata relatively small "debris collar" around the base of the sump screens could cause a significant reduction in the minimum flow path area.As discussed in the NMC responses to Questions 3.A, B and C, no debris is expected to be transported to the screen and there is no mechanistic basis for positing the formation of such a collar.However, if a debris collar is assumed to form, the collar could cause a significant increase in frictional head loss and creates the potential for flashing if saturated fluid is assumed.Due to the configuration of the existing sump screens and their close proximity to the sump outlet valve discs, a relatively small accumulation of debris att he base of the screens could cause a disproportionate amount of head loss in the ECCS suction piping.This condition was evaluated in OPR 162, Revision 1. The findings of OPR 161, Revision 1 are summarized below. For a depiction of the flow details and the calculations involved, please refer to OPR 161, Revision 1,c ontained on the CD-ROM enclosed with this response.OPR 161, Revision 1, determined that, with a "debris collar" around the bottom -2.5" or more of the sump screen, all flow would be diverted through a small (-3/4" wide) annulus with -12" inner diameter.

The effect of the "debris collar" was to increase the hydraulic frictional losses by an additional 4.8' under the maximum permissible flow rate (1582 gpm). Existing calculations had previously determined that the NPSH margin available at the same flow rate would be 10.64', and neglected the 3.22' of submergence.

The net effect is that there is 9'of NPSH margin in excess of the RHR pump requirements, even with the lower few inches already 100%occluded by postulated debris.Page 33 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com Thereforew hile the potential detrimental effect of a "debris collar'has been recognized, the effect does not result in a reduction of ECCS capability beyond that already inherent by the partially submerged screen configuration.

As indicated by the responses to Questions 3.A through 3.C above, no blockage of the sump screens by debris is expected.

The results of OPR 162, Revision 2, indicate additional margin to accommodate debris, even when none is expected.(4) Is there a reasonable expectation that the sump will fulfill its safety function?

What are the major uncertainties and the sensitivity of the answer to those uncertainties?

What is the basis for this answer?NMC Response: There is reasonable expectation that the sump will fulfill its safety functions.

The responses to Questions 3.A, 3.B, 3.C, and Questions 3.D.(1) through (3) show that there is a high degree of confidence that the postulated debris types do not pose a challenge to the ability of the ECCS sump to perform its safety functions.

This is based on regulatory guidance supporting the inability of low velocity fluid fields to transport negatively buoyanto bjects.Additionally, the quiescent period between the blowdown transient and the initiation of sump recirculation provides time for initially suspended debris to settle to the floor of the large pool, whereupon it would not be available for subsequent transport to the sump screens.Uncertainties in the quantities and specific mix of debris that could be generated by various sizes and locations of LOCAs are not significant because the debris would be negatively buoyant to the degree that they would behave similarly for any break location.Since the highest expected flow velocities are less than 0.1 fps,a nd the minimum velocity necessary to transport debris with a specific gravity of 1.05 or greater is twice this value defined in RG 1.82 Revisions 0 and 1,a nd NUREGICR-6773,A ppendix B, a margin of at least two (2) exists to accommodate uncertainties in calculated flow velocities.

Page 34 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com Additionally, as discussed in the response to Question 3.C.(1) above, the lightest potential debris source is a tape with a specific gravity of 1.1. This is a 100% increase in negative buoyancy beyond the threshold specific gravity of 1.05 cited in RG 1.82,R evisions 0 and 1.The other debris sources considered have substantially higher specific gravities.

This represents another conservative factor of two (2) that can accommodate uncertainties in the measurement of this minor debris constituent, and a much larger conservatism for the other debris types considered.

In aggregate, the conclusions of non-transportability and screen operability are based on a foundation of empirical evidence and established regulatory guidance.

Significant uncertainties are limited to the exact flow field velocities.

However, the hydraulics surrounding the sump screens are not complicated by convoluted flow passages, and the limiting flow rates are based on pump capacities and system hydraulic resistances.

As a result, it is estimated that the uncertainties in flow rate (and therefore velocity)are on the order of 10%. With the margins described above, these considerations are enveloped.

E. Affects on Downstream Components:

(1) What types,p article sizes and quantity of materials are expected to pass through the sump screens? What is the basis for this answer?NMC Response: Prior to resolution of concerns related to GSI-191, the types of debris explicitly evaluated to pass through the sump screens have been limited to fragments of disintegrated coatings.

These evaluations are contained in Section 9 of the unit-specific 1989-90 consultant reports, and in Engineering Evaluation 2005-0024, Revision 1. The various evaluations estimated the total quantity of debris fines that pass through the screens and reach the reactor vessel to be from less than 10 ft 3 to up to 27.5 ft 3.These particles have been estimated to have sizes ranging from 10 microns to 0.125" (the size of the ECCS screen perforations).

.1989-90 Evaluations:

These evaluations cite the following assumptions when considering the potential for transport of failed coating fines:* All coatings inside containment fail (353,100 ft 2).* The failed coatings have a particle size distribution ranging from 10 microns to 1.0",w ith the peak at 0.5".* The transportv elocities of the fines can be calculated using the same methods as that used for larger particles.

Page 35 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com In assessing the concentration of these fines within the recirculating fluid, it was estimated that they would be less than 0.1%. When evaluating the potential for the accumulation of these fines in the reactor vessel, two additional assumptions are made:* Coating particles less than 0.015" in size reach the sump screens from far-field transport.

• Coating particles less than 1/8" which reach the near sump screen zone (ZOI/ZOT) are available for transport to the ECCS and reactor vessel.These evaluations estimated that the quantity of debris fines which can pass through the sump screens and reach the reactor vessel would be less than 10 ft 3.Engineering Evaluation 2005-0024, Revision 1: This evaluation assumed disintegration of 100% of all unqualified coatings inside containment to fines with a size range of 1128 microns or less. The total quantity of unqualified coatings assumed was a bounding figure of 22,000 ft .It was further assumed thatb ecause of the small size of the particles, 100% remained in suspension and passed through the sump screens.In performing this evaluation, it was further acknowledged thats ome of the unqualified (and presumed alkyd) coatings may be epoxy coatings thatw ould notb e susceptible to disintegration.

However, assuming failure of all the coatings not known to be acceptable would result in a conservatively high calculated concentration of suspended fines. The total volume of the fines was determined to be 27.5 ft 3 , giving a volumetric fraction of -0.13%.Comparison of Evaluations:

Both the 1989-90 reports and Engineering Evaluation 2005-0024 found comparable quantities of suspended fines (same order of magnitude;d iffering by a factor of 2.8,d espite differing approaches to the question).

The sizes of the particles were also comparable;a 11 particles

<0.015" in the 1989-90 reports, and <1128 microns (0.044") in Engineering Evaluation 2005-0024.

Page 36 of 84 PDF created with pdfFactory Pro trial version wwvi.pdffactorv.com (2) What ECCS equipment/components have tight clearances that could potentially be affected by foreign materials that pass through the sump screens (e.g., pump seals,f low orifices, throttle valve trim,e tc.)? What is the basis for this answer?NMC Response: There are no ECCS components that have tight clearances and/or materials that could be unacceptably degraded by foreign materials that pass through the sump screens. This conclusion is based on evaluations performed by the consultant in 1989-90 and Engineering Evaluation 2005-0024, Revision 1.1989-90 Consultant Evaluations:

These evaluations explicitly considered four differenta spects of potential effects of suspended debris in the recirculating ECCS fluid:* Blockage of fluid systems (Section 9.2)* Effect of Abrasives in the Coatings Debris (Section 9.3)* Debris Accumulation in the Reactor Vessel (Section 9.5)* Potential for Core Blockage (Section 9.6)Components specifically addressed were the containments pray system nozzles, the RHR pumps, Sip umps,c ontainment spray pumps, the reactor vessel, and the fuel assemblies.

In each case, the conclusions were favorable, in that:* The spray nozzles are considerably larger than the maximum debris size postulated.

• The pumps have hard-wear bearing surfaces that will exhibit low wear rates.* The concentration of coatings debris was estimated to be below the threshold of 0.1% established in NUREG/CR-2792 for negligible effecto n pump performance.

• The reactor vessel has considerable free volume in the vessel lower plenum to accommodate accumulated debris (>300 ft 3).* AtO .15", the passages in the grid plates of the fuel assemblies have dimensions greater than the 0.125" screen perforation size.Engineering Evaluation 2005-0024:

This evaluation relied on information compiled for evaluation of components for downstream effects under the on-going effort to resolve concerns related to GSI-191. It determined thatf ailure of the mechanical seals on the RHR and SI pumps from operation with suspended coating Page 37 of 84 PDF created with pdfFactory Pro trial version www.Pdffactorv.com decomposition particles is not expected.

The same design seals are used in applications with similar debris laden fluid such as pulp and paper, petrochemical,f ood processing, and waste water treatment.

The evaluation concluded that the orifices (flow instrumentation and flow limiting orifices) in the credited ECCS flow paths are on the order of inches,a nd therefore,a re nots ubject to blockage by the fine particulates.

Also, after sump recirculation is established, valves in the flow path are not relied upon to reposition.

The valves in these flow paths are 2" or greater in size, and have stainless steel or harder wearing surfaces.

While itm ay be desirable to throttle the 8" diameter RHR heat exchanger outlet butterfly valves, these large diameter valves are not expected to be susceptible to significant degradation from suspended particulates.

Based on these considerations; wear,e rosion and blockage of valve components are not a factor.The evaluation also concluded, as did the 1989-90 reports, that the reactor vessel and core flow passages are on the order of fractional inches or more, and are not susceptible to fouling by the fine particulates.

4. SI-850 Valves A. What are the safety functions of the valves (e.g., to open/stay open, to shut/stay shut, to maintain leak tightness) and what is the basis for this determination?

What are the ECCS pump minimum and maximum recirculation flows and net positive suction head (NPSH)requirements?

What is the basis for this answer?NMC Response: The SI-850A&B have a safety function in both the open and shut directions.

Safety-related systems,s tructures and components (SSCs) include SSCs that are relied upon to remain functional during and following design basis events to assure the capability to shut down the reactor and maintain it in a safe shutdown condition.

The ECCS system is designed such that the failure of a single active componento r the failure of a passive component during the long-term cooling period does not interfere with the ability to meet the necessary long-term cooling objectives.

The RHR system is designed to provide the following safety-related functions:

• Deliver borated cooling water to the RCS during the injection phase of SI* Recirculate and cool the water that is collected in the containments ump and return it to the RCS during the recirculation phase of SI Page 38 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com

• Provide the means to preclude containment leakage through the RHR system piping penetrations following accidents* For piping and components that are part of the reactor coolant pressure boundary,m aintain pressure boundary integrity during all modes of plant operation Recirculating and cooling water that is collected in the containments ump and returning it to the RCS requires the SI-850 valves to have a safety function to open and stay open. Since the SI-850 valves are the only installed valves thatc an isolate a passive failure of the piping between containment and the 1(2)SI-851A&B, the Sl-850 valves have a safety function to shut and remain shut to minimize the effect of a passive component failure through the RHR system post-accident.

This leakage is a passive failure of one suction line (excessive packing or weld leakage) and will not impair the operation of the redundant valve. Shutting the SI-850 valve for the affected train stops excessive leakage during the long-term cooling, and therefore, this excessive leakage cannot interfere with the other system from performing its long-term cooling objectives.

Dose consequences and the licensing basis are addressed in Question 5.E.The industry definition of passive failure evolved during PBNP's original licensing and culminated in the redesign of the ECCS system, including the inclusion of the SI-850 valves. The SI-850 valves were installed specifically to isolate the ECCS line to "minimize" the effect of a passive component failure.Containment Sumg Outlet Flows (Recirculation Flows): At the initiation of containments ump recirculation, a single operating RHR pump's suction is switched from the RWST to the containment sump. The flow rate from the sump would be bounded by a maximum of 1582 gpm. The operating containments pray pump(s) would continue to draw down the RWST inventory until the criteria to secure the pumps is reached. The SI pumps would be secured once it was verified that the RHR pumps were providing adequate injection flow (this action prolongs the period that the RWST is available for injection).

The flows thatc ould be provided under other alignments have been analyzed, but these flows are notp rocedurally permitted because adequate NPSH is not available to the RHR pumps. RHR flow is injected into the core through the core deluge nozzles. Therefore, as long as the injection flow is greater than the core cooling flow requirement, the core will receive adequate cooling and the excess will be diverted out the RCS break. The procedurally limited flow of 1560 gpm exceeds the core cooling required flow at the starto f recirculation.

The procedurally limited flow is a resulto f NPSH concerns as described in OPR 162. Resolution of this OPR will address this condition.

Page 39 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com Boron Precipitation Control: The flow values contained in the emergency operating procedures (EOPs) are maximum allowable flows and not minimums.

Boron precipitation control is obtained at a reduced maximum flow of 1560 gpm. In the May 7, 1975, licensee response to an NRC letter dated March 14, 1975, on this subject, ECCS long-term cooling requirements were provided.

The evaluation was based on the generic Westinghouse evaluation," Long Term Cooling -Boron Considerations." This submittal stated thatf low from a single pump (187 lbs/sec or-1350 gpm) during a large break LOCA was more than adequate to prevent boron precipitation.

NetP ositive Suction Head: Recirculation operation gives the limiting NPSH requirement.

The available NPSH is determined from the containment water level, and the pressure drop in the suction piping from the sump to the pumps. The RHR pumps need at least -8' of NPSH atl 582 gpm. The RHR pumps are located atE 1. -19'3" to assure the necessary NPSH at the pump suction when the recirculating water is at 212 0 F with atmospheric pressure in the containment.

B. Have the valves been adequately tested to demonstrate that they will perform each of their safety functions identified above? Explain and identify what testing has been performed?

What is the frequency of this testing and how do the test acceptance criteria demonstrate/relate to the valve safety function?

What is the basis for this answer?NMC Response: The Sl-850 valves have always had a safety function to open and are tested to ensure the open function is maintained as discussed below. During the review of the safety functions for the valves (see response to Question 4.A), it was identified that the valves also have a safety function to shut. A corrective action program document was initiated since the valves were not currently credited in the PBNP inservice testing program as performing a safety function in the shut direction.

The shut safety function testing has been incorporated into IST program. Seatl eakage testing that would be required for a safety-related function in the shut direction is addressed in the response to Question 5.1.Open Safety Function:

These normally shut, hydraulically-operated valves are located inside containment in the line leading from Sump "B" to the suction of RHR pumps. The valves perform an active safety function in the open position.

The SI-850 valves must be capable of opening, by remote manual switch actuation,w hen transitioning from the injection mode of Si to the recirculation mode of SI. When the initial supply source of SI water from the RWST is effectively depleted following a LOCA,s uction for the SI and RHR pumps is switched from the RWST to the containment sump to provide long-term core cooling. The SI-850 valves receive no automatic actuation signals to open and must be aligned from the control room using their Page 40 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com associated control switches.

They have no maximum design stroke time limits associated with the safety function in the open position and fail "as-is" on a loss of power.The SI-850 valves are tested quarterly in the open direction via the inservice testing (IST) program using Inservice Test (IT) 40, 'Safety Injection Valves (Quarterly)

Unit 1",a nd IT 45 "Safety Injection Valves (Quarterly)

Unit 2".ASME OM Code Paragraph ISTC 4.2.8 provides the basis for IST acceptance criteria for open stroke time and for position indication verification (PIV) testing.Valve stroke time acceptance criteria are based on ASME OM Code,w hich sets the acceptance criteria at the baseline reference value +/- 50%. IT 40 and IT 45 check the output pressure of the SI-850A(B) hydraulic pumps in both the open and shutv alve stroke directions.

PBNP currently requires that the hydraulic pressure not exceed 1500 psig when shutting the valve to prevent damage to the hydraulic operator.

In the open direction, the hydraulic pressure for each valve is required to be between 1150 and 1500 psig. The 1150 psig lower limit is to ensure that the hydraulic operator is capable of opening the valve against the weight of the valve, packing friction, head of the containment sump and post-accident containment pressure.ShutS afetv Function:

The valves had previously not been credited with a having a shut safety function.

This is documented in a corrective action program document.

The valves' shuts troke time and shut position indication verification is performed under IT 40 and IT 45 for trending of valve degradation using the same guidance of ASME OM Code Paragraph ISTC 4.2.8 as the open direction test.The response to Question 5.1 documents the allowable leakage requirements of the Sl-850s in the shut position.C. Is there a reasonable expectation that the valves will perform their safety functions for the duration of the events, as defined in the safety analyses?

What is the basis for this answer?NMC Response: There is a reasonable expectation that the SI-850A(B) valves (containment sump "B" isolation) will be able to perform their safety functions for the duration of the events as described in the safety analyses.The design basis for these valves has previously identified that the valves only had an open active safety function.

CAP 069891 has identified that an active safety function in the shut direction also applied.The ECCS system is designed such that the failure of any single active componento r the failure of a passive component during the long-term cooling period does not interfere with the ability to meet the necessary Page 41 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com long-term cooling objectives.

The PBNP licensing basis assumes a passive failure or an active failure during the long-term cooling phase. Either an active or a passive failure would remove one train of ECCSI eaving the other train operable to recirculate and cool the water that is collected in the containment sump and return it to the RCS. Therefore, the only credible mechanisms to prevent the SI-850 valves from performing their safety function would be environmental qualification considerations.

The containment sump isolation valves have hydraulic cylinders for opening and shutting the valves, which are mounted directly to the piping in the containment tendon gallery. The hydraulic pumps which provide pressure to the cylinders for operation are mounted in the PAB pipeway hallway. A review of the valves' ability to perform intended design functions was performed.

The review determined that the hydraulic cylinder components pressurized to shut the valve and the hydraulic units located in the PAB pipeway meet the required service conditions for valve operation.

It was determined that the containment sump "B" isolation valves are not in compliance with the environmental qualification (EQ) program requirements.

As evaluated in OPR 171, SI-850A(B) are capable of performing their safety-related function throughoutt he recirculation phase.Presuming one SI-850 valve was shut to minimize the effect of a passive component failure, the valve would remain in the shut position.

The forces from the containment recirculation sump liquid level in addition to a containmento verpressure would maintain the valve shut.T he remaining valve is maintained opened to ensure one train of RHR is in operation for core cooling.The open containments ump isolation valve needs to remain open to provide a suction path for the RHR pump.T he valve and its operator are designed for the post-accident environmental conditions in the PAB pipeway hallway and the tendon gallery. The system was designed to mitigate either an active or a passive failure during recirculation operation.

In the evento f a single active or passive failure during long term cooling the second train of RHR would be placed in service to ensure core cooling is maintained.

The potential for valve drift and resulting effects are discussed in the response to Question 4.D.Page 42 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com D. What are the consequences if the valves fail closed? If the valves fail closed, can they be re-opened?

If these valves can drift shut, what amount of closure will cause the open indication in the controlr oom to be lost and what will be the effect on recirculation flowINPSH/pump operation with these valves in this partially closed position?

Is the equipment that provides control room position indication qualified in accordance with 10 CFR 50.49, "Environmental qualification of electric equipment important to safety for nuclear power plants? If they can be re-opened what are the consequences of the time period the valves are not fully open? What is the justification for the time period assumed?Can the valves be opened with pumps in operation?

What is the basis for this answer?NMC Response: Consequences of Valve Closure: Failure of both valves to open would require two active failures, which is outside the design basis for PBNP.Failure of one valve to open is considered the single active failure of the RHR system and the second train of the RHR system would be started to provide containment sump recirculation.

The valves in one train are verified as open prior to establishing containment sump recirculation using plant procedures.

There are no common cause failures that could affect both trains.The hydraulic units located in the PAB pipeway are environmentally qualified and would allow remote operation of the valves from the control room during the complete duration of a LOCA event requiring long-term recirculation.

Failure of either containments ump "B" isolation valve to open requires operator action to manually open the valves using a staged hydraulic hand pump. If containments ump recirculation can not be established, there is procedural guidance which directs operators to utilize contingency actions.Additional information regarding radiological considerations for local operation is provided in the response to Question 4.E.The SI-850A/B valves can be reopened if they move from the open position to an intermediate or shut position.

The valves can be remotely opened, or the staged hand hydraulic pump can be used in accordance with established emergency operating procedures and contingency actions.Use of the hydraulic operator requires placing the associated valve's control room hand switch to open position to reestablish the hydraulic force to open the valve. This is an expected response of the control operator to the valve being out of position as specified per procedures.

Page 43 of 84 PDF created with pdfFactory Pro trial version wwv/.Pdffactorv.com Control Room Indication:

The valve travel distance that will bring in the control room intermediate indication from the open valve position was measured via performance of work orders that obtained stroke distances.

The calculation performed for OPR 162 shows that there will be adequate NPSH for the RHR pumps when an SI-850A (B) valve is at an intermediate indicated valve position.

OPR 162 also established that there would be adequate time to take remedial action, such as reopening the valve, between indications of valve drifting and the loss of the RHR pump suction source.Environmental Qualification:

The equipmentn ecessary for the control room to remotely observe valve position and operate the SI-850A(B) valves during the duration of the recirculation phase is capable of operating within the environment expected.

EQ documentation is currently nonconforming and will be upgraded.

See the response to Question 1.C for additional detail.Consequences of Partial Valve Closure: There is nota defined drift closure time period or inadvertent closure period assumed for the valves with RHR in operation either in the licensing or the design basis for PBNP. Failure of one valve to open is considered the single active failure and the second train of containments ump recirculation would be started. The valves are verified as open prior to establishing containment sump recirculation.

A corrective action program document was initiated to investigate the potential for a recirculation valve to drift shut.C losure of the valve would impact the NPSH to the recirculation RHR pumps as the friction factor of the water flowing to the RHR pump suction would increase as the valve drifted shut. Drifting of the valve to the shut position would be apparent to operating personnel as a loss of RHR pump flow via main control board indication.

Sump "B" recirculation would be restored in accordance with approved plant procedures.

Each containment sump 'B" isolation valve has position indication in the control room with red and green position indicating lights adjacent to the control switches.

Changes in the status of this indication would be apparent to operators.

Prior to assuming control room duties each shift, and frequently during the shift, licensed operators are required to perform a main control board walkdown thatw ould identify potential changes in valve position indication.

A corrective action program document was initiated to investigate potential hydraulic fluid leakage that could affect SI-850A(B) valve drift. Hydraulic fluid leakage paths to consider would be past the hydraulic cylinder internal piston ring, out of the system, or through the solenoid-operated pilot valve under the influence of gravity and the differential pressure generated across the valve by flowing sump or RHR fluid. A qualitative assessments f leakage is provided below based on the valve operator design.Page 44 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com The impacto f internal piston ring leakage would be expected to be a long-term degradation mechanism as minor leakage from the hydraulic cylinder could be postulated through the hydraulic seals to the environment.

Gross leakage is not assumed as it would be observed during functional testing since the stroke time of the valve would change as a result of this failure mechanism.

Failure of the valve hydraulics would be considered an"active" failure and would apply to only one of the two valves.The hydraulic system solenoid-operated valve utilizes close tolerance metal seating surfaces.

Leakage past the hydraulic system solenoid-operated pilot valve would be related to hydraulic pressure created as a result of forces placed on the hydraulic cylinder.

The force on the hydraulic cylinder would be generated by the influence of gravity and the differential pressure generated across the SI-850A(B) valves by flowing sump/RHR fluid creating a valve stem force. The SI-850A(B) valve stem load trying to shut the valve (weighto f valve, stem ejection forces) verses stem forces maintaining the valve open( piston friction,s tem friction) is very small. Therefore, if drift occurs, it would be relatively slow. A drift rate is notd efinable in the amount of time it would take to cause the intermediate valve indication to actuate.PBNP has not tested potential drifting of the SI-850A(B) in the closing direction to date. A corrective action program document has been initiated to determine testing methods and to establish acceptance criteria.In OPR 162, a sensitivity evaluation of NPSH verses valve position was performed.

The results indicate thatv alve drift would have to occur before the partially shutv alve would begin to create more of a pressure drop than when it was full open. The maximum possible open stroke is 2.5". Field measurements determined that the valves are set to provide a full open stroke of at least 2". The point atw hich the valve begins to increase head losses above the acceptable head loss is .85" of open travel. As noted previously, intermediate position indication lights for these valves on the main control boards would show drift by at least 1.25" of open valve travel.Therefore, based upon the control room light indications and industry experience that a hydraulic leak is expected to be slow, licensed operators would recognize potential valve drifting shutp rior to impacting core cooling.Can the Valves be Open with the RHR Pumps in Operation:

The containments ump "B" recirculation valves can be opened with the RHR pumps operating when still aligned to the RWST as a suction source. The differential pressure forces on the valves at this time are less than the forces assumed in Calculation 2001-0001, "Hydraulic Pressures Associated with the SI-850 Valves," since RWST water level head will reduce the forces assumed in the calculation.

Page 45 of 84 PDF created with pdfFactory Pro trial version ww\w'.pdffactorv.com In the event of valve drift, the S1-850 valves would be capable of stroking from a partially shut position to full open. The forces required to reopen the valve to the full open position are significantly less than maximum loads used in Calculation 2001-0001 since containmentp ressures would be lower and the differential pressure across the valve disc would be minimal because the valve is still open.E. What is the radiation exposure to the operator if local manual action is necessary?

What is the basis for this answer?NMC Response: The SI-850 motor-hydraulic units,r eferred to as the valve operators,t hat would be accessed to manually change the position of the SI-850 are located outside Pipeways 2 and 3 in the access gallery for Unitl and Unit2, respectively.

The valve operator is shown in the FSAR Figure 6.2-2 and as Attachment 5 of this enclosure.

The valve operators for Unit 1 and Unit 2 are on the EI.8 'located near 1(2) RK-51/52 Pipeway 2 (3) instrument panels,w hich are shown on FSAR Figure 1.2-4. During the design and construction of the plantt he motor-hydraulic units for each of the SI-850 valves were intentionally located in the PAB such thata ccess post-accidentc ould be made if needed. The SI-850 valves can be operated remotely from the control room.Since the passive failure of one suction line (presumably excessive packing or weld leakage) will not impair the operation of the redundant valve, multiple failures would have to occur to require local operator action. For example, assuming a failure occurred on the inservice recirculation train, local operator action would be necessary if the operator is unable to isolate the failed train from the control room or the operator is unable to place into service the opposite train. Multiple failures are not taken in conjunction with a design basis event. Therefore, consistentw ith the design basis for PBNP, access to these valves was notc onsidered required,b utw as possible, based on the intentional selected physical location.Dose considerations for local manual action in the PAB post-accident are described in FSAR 11.6 under auxiliary shielding.

The auxiliary shielding is based on a design basis LOCA with minimum safeguards that results in a gap release of all of the fuel rods, as determined by the 10 CFR 50.46 evaluation presented in FSAR 14.3.2 and discussed in FSAR 14.3.5.Specifically, FSAR 11.6 states the following: "All components necessary for the operation of the external recirculation loop following a loss-of-coolanta ccidenta re capable of remote manual operation from the control room and can be powered by the emergency diesel-generators so that its hould not be necessary to enter the auxiliary building in the vicinity of the recirculation loops." Page 46 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com This section of the FSAR goes on to state that if access is essential to the continued operation of the engineered safeguards system during the recirculation phase dose reduction measures would be applied. Such dose reduction measures would be additional shielding, limited duration and respiratory protection.

Estimated dose rates in the vicinity of the RHR recirculation piping are stated as 25 R/hr one-hour,p ost-accidentw hereas dose rates on the recirculation loop are stated as 200-300 rem/hr immediately following the initiation of recirculation.

The basis for these dose rates is provided in Table 11.6-6 of the FSAR.In response to NUREG-0578 Item 2.1.6.b, "Design Review of Plant Shielding of Spaces for Post-Accident Operations," reissued as NUREG-0737 Item II.B.2,P BNP re-evaluated the shielding design of the PAB to ensure areas requiring post-accident access were habitable.

This review was performed under the assumption of a fuel-melt accidenta nd resulted in several shielding modifications.

The location of the containment sump suction isolation valve operators were not identified as vital areas, that is, areas requiring access post-accident; however, these areas were shown to be inaccessible.

Access to this area is limited due to the direct radiation from unshielded low pressure safety injection lines transporting liquid from the RHR heate xchanger to the safety injection/containments pray pump room. Acceptance of the implementation of NUREG-0578 Item 2.1.6.b was provided to PBNP on April 9, 1980,w hen the NRC acknowledged that shielding was generally adequate and additional shielding of the C-59 control panel was under consideration.

Permanent and portable shielding was later placed in the area of the C-59 panel as well as other areas of the PAB. This work was communicated to the NRC via responses to NUREG-0737 Item II.B.2. The NRC accepted the NUREG-0737 Item II.B.2 vital access response on November 3, 1983.Therefore, based on a review of the current licensing basis and design basis of PBNP,I ocal operator action is not necessary to open/shut the containments ump suction valves post-LOCA.

This is because they are remote-operated valves and a single failure on one recirculation train will not prevent the other train from performing its design function.

Under the presumption of a radiological design basis LOCA (i.e., fuel melt), the location of the valve operators is nota ccessible due to the unshielded recirculation lines in the vicinity of the operators.

However,u nder the presumption of a design basis LOCA that credits minimal safeguards on injection (i.e., gap release);

these areas would be accessible on a limited basis if additional protective measures were taken into consideration.

Page 47 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com F. Will flashing occur in the piping below the valves when they are opened to perform their safety function during an event, including the long term? Consider containment overpressure, ECCS flow, and the number of ECCS trains in operation.

If containment overpressure is needed, has it been analytically shown that the minimum overpressure assumed in the analysis will be present for the limiting combination of conditions (e.g., including inadvertent operation of secured equipment that could reduce containment pressure),i ncluding the long term?What is the basis for this answer?NMC Response: Flashing will not occur in the piping below the valves as described in OPR 162 (enclosed on CD-ROM). During the recirculation phase containmente quilibrium pressure due to the partial pressure from air existing in containmentb efore an accident and a partial pressure from steam at 212 0 F due to a pool of water att he bottom of containment are credited.

However,a s discussed in OPR 162, crediting the containment equilibrium pressure is not in conformance with the current licensing basis.To change the conclusions of the OPR, air would need to be removed from containment.

The containment structure is designed for the pressure and temperature resulting from a design basis accident; however,a breach of containmenti s notw ithin the design basis of PBNP.In order for a large amount of air to be quickly removed from containment, a relatively large opening that can vent air must be made in containment.

Large openings that could communicate directly with the atmosphere can be made by inadvertent operation of the purge system, opening of both containmentd oors, or opening of the fuel transfer canal. The purge system contains blind flanges on both penetrations during Modes 1 to 4 that would need to be removed in order to use the purge penetrations.

The containmentd oors are mechanically interlocked such that only one door at a time can be opened. The fuel transfer canal contains a blind flange that must be removed prior to the use of the penetration.

The next largest penetrations are the main steam and main feed penetrations.

These penetrations are connected to a closed system inside containment.

Therefore it is not likely that large amounts of air could be released by inadvertento peration of secured equipment.

OPR 162 used the mostl imiting train of RHR for the determination of flashing.

Each train has an independentf ine screen suction strainer.Cross-connection of both trains of RHR is not an alignmentd irected by the EOPs. Therefore, both trains cannot draw suction off the same fine strainer, so the flow rate assumed in the OPR is bounding.The PBNP licensing basis assumes the failure of a single active component or the failure of a passive component during the long-term cooling period.Inadvertent operation of a second train is not within the design basis for this system. However as stated above, if the second train was started, it would Page 48 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com draw from its own strainer and S1-850 valve,a nd would not have an effect on the flashing considerations.

G. If flashing occurs, what are the potential consequences?

What is the basis for this answer?NMC Response: Flashing does not occur as documented in OPR-1 62 and discussed in the previous question.5. ECCS Leakage from the Recirculation Line (flange/body-bonnetlpacking/weld)

A. What is a technically defensible failure (leakage rate) to consider and when and where are these leaks postulated to occur? What is the basis for this answer?NMC Response: As defined in FSAR 6.2, the passive failure of one suction line is assumed to be due to excessive packing or weld leakage that will not impair the operation of the redundant recirculation train. This FSAR section also indicates that a RHR pump seal failure rate is 50 gpm.During normal plant operation, the leakage limit from the ECCS is maintained to be 400 cc/min or less.T his 400 cc/min value is constrained by the control room dose analyses and is described in FSAR Sections 6.2 and 14.3.5. The control room close analyses assumed an ECCS leak rate of 400 cc/min, for 30 days following an accident.The final form of the currentr adiological analysis for control room habitability was communicated by NMC to the NRC on June 3, 1997. This submittal provided additional information as a basis for the exclusion of a passive failure post-LOCA.

The analyses of record were approved in a Safety Evaluation Report dated July 9,1 997, "Issuance of Amendments Re: Technical Specifications Changes for Revised System Requirements to Ensure Post-AccidentC ontainment Cooling Capability." The primary basis for the exclusion of passive failure was the assertion thatr adiological dose post-LOCA for PBNP had not previously assumed a passive failure in conjunction with the design basis radiological analysis.

The only assumed failure for the LOCA radiological design basis dose analysis has been the loss of an emergency diesel generator, which limits the containment spray and ventilation systems to one train each.The credible leak sources for this type of leak consist of a malfunctioning residual heat removal pump seal, flange gasket, or a valve with degraded packing. The flow rate from any one of these sources will be less than 50 gpm. Original Technical Specification 15.4.4 for PBNP (April 1970)Page 49 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com stated, "The limiting leakage rates from the residual heat removal system are a judgment value primarily based on assuring that the components could operate without mechanical failure for a period on the order of 200 days after a design basis accident." This value was used in Chapter 14 of the Final Facility Description and Safety Analysis Report (FFDSAR).During the recirculation phase, continuous ECCS leakage may become airborne and escape through the PAB vent stack to the environment.

This leakage is not expected to exceed 400 cc/min. Radiological analyses of offsite dose due to this leakage have conservatively doubled the expected ECCS leak rate; assuming a combined ECCS leak rate of 800 cc/min during the accident.

Offsite radiological consequences of the LOCA,i ncluding this ECCS leakage, are described in FSAR Section 14.3.5. The 50 gpm passive failure is not included in the dose analysis, since this leakage is expected to occur around 200 days following an accident, which is after the 30 days assumed in the offsite and control room dose analysis.To maintain the leakage limit of 400 cc/min or less for the dose analyses,a series of Leakage Reduction and Preventive Maintenance (LRPM) tests are performed during each refueling outage. These tests measure and quantify the leakage from the system to the atmosphere by looking at leakage from individual components outside containment (i.e., valves, body-to-bonnet joints, packing) and portions of trains or systems. Seat leakage at boundary valves is included in the total leakage value as leakage to other systems may ultimately be exposed to atmosphere.

The leakage determined in these tests is collected at conservatively higher test pressures than would be experienced during a design basis event.Leakage-to-atmosphere in the LRPM program is maintained "as low as reasonably achievable." Active leakage, typically on the order of drops per minutei s corrected prior to completion of a refueling outage. Acceptance of an active leak requires a corrective action program document to be initiated and the active leak evaluated as acceptable for unit restart.In summary, the design basis leak rate for a passive failure in the ECCS containments uction line is 50 gpm. This leak is the expected worst case for a RHR pump seal failure that bounds all other leakage in the suction line through packing or weld leakage. This passive failure is not included in radiological analyses as currently defined in PBNP licensing basis. The time for such a passive failure to occur is on the order of 200 days following a design basis accident.Page 50 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com B. What compensatory measures are available to detect and isolate this leakage? If non-safety related equipment is relied on to support detection and isolation explain why this is appropriate.

What is the basis for this answer?NMC Response: There are three general areas where a passive failure in the containment sump recirculation line to the RHR pumps could occur: The tendon gallery;the RHR system pipeways in the PAB;a nd the RHR pump compartments.

Leakage resulting from a passive failure in the tendon gallery: The plant design is such that leakage in the tendon gallery would have a flow path to the "A" RHR pump compartmenta nd would be detected by the level transmitter in this compartment.

However, it was discovered that the tendon gallery sleeves are grouted closed and this leak path is currently not available.

CAP 069723, "Design Basis Leakage Detection Capability May Have been Defeated," was submitted on January 10,2 006, in response to this discovery.

OPR 170 concluded this condition as operable but nonconforming.

The OPR determined thats ufficient time and containment sump volume is available for detection by means of control board indications prior to challenging core cooling post-accident.

.If the leakage from the ECCS system to within the tendon gallery occurred, the means for detection of this leakage would depend in part on which systems are operable/functional post-event.A nonsafety-related sump pump automatically starts when tendon gallery water level increases.

The tendon gallery sump pump automatically pumps water to the facade sump.As long as the tendon gallery sump pump is functional, plant operators would receive a facade sump alarm. When this alarm is received, the sump is pumped outu sing approved plantp rocedures, 1 (2)-SOP-WL-002,"Pumping Facade Sump Unitl (2)." Sump samples are taken prior to pumping the facade sumps. Adverse chemistry results would prompt an immediate investigation by operators, as skill of the craft, into the source of the leak so the leak location could be identified.

Similarly, if repeated pumping of the sump occurred over a short period of time, an immediate investigation would be initiated into the source of the leak.A passive leak would result in a reduction of containment sump "B" level.During recirculation, the two safety-related redundantc ontainments ump B level transmitters are monitored,s o the control room staff would detect a gross change in the containment sump "B" level. A gross change in the containments ump "B" level would be noticed within at least one shift. Once a gross change in the containment sump "B" level has been observed, an immediate investigation into the source of the level change would be initiated by the control room staff.OPR 170,D esign Basis Leakage Detection Capability May Have Been Defeated, and OPR 171,S afety Function for ContainmentS ump "B" Page 51 of 84 PDF created with pdfFactory Pro trial version wwvi.pdffactorv.com Isolation Valves,d emonstrate that despite a passive failure in the recirculation lines of the RHR system, the safety function of the system can be maintained.

Isolation of the passive failure would be accomplished by shutting the SI-850 and SI-851 valves on the failed train of containment sump recirculation.

If the tendon gallery sump pumps failed or would not function, the tendon gallery could potentially flood up to the facade floor (El. 6'-6") before the leak was detected and isolated.

Filling the tendon gallery would take about 82,400 gallons of water. If the bounding leak rate of 50 gpm was located somewhere within the tendon gallery, it would take approximately 27 hours3.125e-4 days <br />0.0075 hours <br />4.464286e-5 weeks <br />1.02735e-5 months <br /> to fill the tendon gallery with water to the facade floor. Additional detail for time availability for response to this failure is provided in the response to Question 5.C.Leakage resulting from a passive failure in the RHR valve pipewav in the PAB: Leakage into the RHR valve pipeway would reach the RHR pump compartments and would be detected by the non-safety related level transmitters in the RHR pump compartments.

Leakage in the RHR pipeways would collecto n the floor in the pipeways, which are located behind the RHR pump compartments.

It would then drain through the associated RHR pump compartment wall into the RHR pump compartment.

This pipeway is divided into two sections by a 7' wall. At the bottom of each of these sections of the pipeway there is a 4" square hole that runs through the RHR pump compartment wall into the RHR pump compartments.

Both SI-851 valves are located directly behind the 'A" RHR pump compartment in the RHR pipeway and leakage from the "A" and/or "B" train upstream of the SI-851 valves will show up in the "A" RHR pump compartment.

Once leakage drains into the RHR pipeway, it would be handled as if it was leakage in a RHR pump compartment.

Leakage resulting from a passive failure in the RHR pump compartments:

Leakage detection in the RHR pipeways and RHR pump compartments would be achieved through the use of sump level detection.

Leakage that reached the RHR pump compartments, either from the RHR pipeway or from within the RHR pump compartments, would be detected by the nonsafety-related level transmitter installed in each of the RHR pump compartments.

As a result, an RHR pump room high-level alarm would be indicated on the control room main control boards. Each RHR pump compartment is equipped with a floor drain and separated equipment drains.The floor drain from each RHR pump compartment flows through an individual pipe to the El. -19' PAB sump. Two 75-gpm sump pumps transfer the leakage collected in this sump to the waste disposal system for processing.

The supply and discharge piping and valves for the RHR pumps are located in a pipeway adjacent to the pump compartments.

Procedural guidance for detection and isolation of a leak that flows into the RHR pump compartment is provided in emergency operating procedure (EOP) EOP-1.3, "Transfer to Containment Sump Recirculation Low Head Page 52 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com Injection." This procedure firstd etects which RHR pump compartmenti s affected by use of the individual level indicators located in each pump compartment.

The procedure then directs the operator to open the affected drain valve to the El. -19' PAB sump. Once this drain is opened, the frequency of operation of the -19' PAB sump pumps is monitored to attempt to quantify the rate of the leakage from the passive failure in the recirculation line. If needed, the RHR pump in the train with leakage would be shut down and isolated by closing the SI-850 and SI-851 valves for that train. This prevents gross diversion of containment sump inventory through the failed containments ump recirculation line.The level transmitters, level switches and sump pumps used to detect a.passive failure within the PAB (RHR pipeway or RHR pump compartment) use nonsafety-related power supplies and components.

However, a safety-related bus, through a nonsafety-related power panel, powers the RHR pump compartment level switches.

This arrangement provides reasonable assurance the level transmitters will be available to detect a flooding concern within the RHR pump area.The RHR pump compartment drain isolation valves are also powered from nonsafety-related buses,w hich in turn, are powered by safety-related buses. Again, this arrangement provides reasonable assurance the drain isolation valves will be available to mitigate a flooding concern within the RHR pump area.In addition, the RHR pump compartment level switches are manually lifted each quarter to assure that they are working properly and producing control room alarm and indication.T he RHR pump compartment drain isolation valves are also operated quarterly from the control room to assure that they are functioning properly.The El. -19' PAB sump pumps are powered from two independent power supplies; one from a Unit 1 power supply and one from a Unit 2 power supply. While they are powered from nonsafety-related buses, these buses are powered off safety-related buses (2B03 and 1 B04) that have emergency diesel generator supplies.

During an accident, the nonsafety-related buses receive a safety-injection stripping signal. However,a s the accident progresses into the recovery phase and safeguards electrical demand decreases, operators would be able to reenergize the stripped bus, as needed, to support flood mitigation concerns within the RHR pump area.For a design basis LOCA coincident with a loss of offsite power,b oth El. -19' PAB sump pumps would have power stripped.

PAB sump level detection, however, would remain energized,t hus prompting operators to reenergize the sump pump power supplies, if needed. EOPs direct the motor control center for the PAB sump pumps to be restored and the PAB sump level to be monitored.

Page 53 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com As discussed above, isolation of a passive failure in the containment sump recirculation lines would be accomplished by shutting the SI-850A(B) and Sl-851A(B) valves. The Sl-850 valve discs are located inside the containment.

A dedicated hydraulic pump located in the PAB is used to control a hydraulic cylinder located in the tendon gallery, which opens and shuts the valve. The downstream SI-851 valves are motor-operated gate valves located in the PAB. Both the SI-850 and SJ-851 valves are considered to perform an active safety-related function in both the open and shutd irections.

The NMC response to Question 5.H discusses the recent change in safety classification associated with the SI-850 valves in the shut direction.

Both the SI-850 and SI-851 valves are included in the IST program and are tested quarterly in accordance with the ASME OM Code. A review of test data confirmed that none of these valves experienced an inservice testing failure over the last fuel cycle that would challenge the ability of the valve to perform its intended safety functions.

In addition to the inservice testing program, PBNP has several other programs in place to assure that the containments ump recirculation lines and associated components are capable of performing their intended safety functions.

The programs include testing the emergency core cooling system (ECCS) via the Leakage Reduction and Preventive Maintenance (LRPM)program. Leakage from the ECCS recirculation line is routinely checked and monitored during the performance of the LRPM tests on a refueling outage frequency.

The Units 1 and 2 LRPM databases are maintained and updated during the performance of the LRPM tests with the total ECCS leakage being recorded.

The total ECCS system leakage is verified to be less then the FSAR Chapter 6.2 limito f 400 cc/min.PBNP's preventive maintenance program also supports the reliability of the SI-850 and SI-851 valves. The valve operators for the SI-850 valves are disassembled and inspected every 10.5 years. The operators for the SI-851 valves are diagnostically tested every 4.5 years and are disassembled and inspected every 12 years.PBNP's design, testing and maintenance programs provide assurance that both the safety-related and nonsafety-related components within the containments ump recirculation lines and the sump systems used to detect and manage leakage, remain capable of performing their functions, including mitigating the consequences of a passive failure within the lines.Page 54 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com C. How long will detection and isolation of a passive leak take? What is the basis for this answer?NMC Response: Passive Failure in Tendon Gallery: The tendon gallery sump pump is expected to pump about 5 gpm to the fagade sump. The facade sump alarm corresponds to about 482 gallons. Should the facade sump have been emptied immediately prior to the passive failure,a n alarm in the control room would be received in less than two hours. It would take longer to identify a leak that is smaller than the tendon gallery sump pump's capacity.Grouting was discovered between the tendon gallery piping sleeve and pipe.The grouting prevents leakage from the ECCS suction line in the tendon gallery from entering the RHR pump compartment.

An operability recommendation (OPR 170) concluded the condition was operable but nonconforming because sufficient time and containments ump volume were available to detect a 50 gpm leak prior to the loss of net positive suction head on the RHR pumps. As a result of this nonconforming condition gross containments ump leakage would be used to identify the leakage source as discussed below.A passive leak would result in a reduction of containment sump "B" level.During recirculation, operators routinely monitor the two safety-related redundant containment sump "B" level transmitters and would notice a gross change in the containment sump "B" level. A gross change in the containments ump "B" level caused by a 50 gpm leak would be noticed within at leasto ne shift. This is based upon control board reviews and daily log sheets. Once a gross change in the containment sump "B" level has been noticed, an immediate investigation into the source of the level change would be conducted by control room personnel.

Recent operability evaluations (OPR 170 and OPR 171) have demonstrated that despite a passive failure in the recirculation lines of the RHR system, the safety function of the system can be maintained.

Isolation of the passive failure would be accomplished by closing the SI-850 and SI-851 valves on the failed train of containment sump recirculation.

Containment sump level would approach the minimum NPSH requirements for the RHR pumps in about 57 hours6.597222e-4 days <br />0.0158 hours <br />9.424603e-5 weeks <br />2.16885e-5 months <br /> assuming a 50 gpm leak rate. Sufficient time exists between detection (within one shift) and loss of decay heat removal capabilities (about 57 hours6.597222e-4 days <br />0.0158 hours <br />9.424603e-5 weeks <br />2.16885e-5 months <br />) to allow operators to isolate a postulated 50 gpm passive leak.Passive Failure in RHR Valve Gallery or Pump Cubicle: The original design of the RHR pump compartments and the adjacentc ompartments are designed so they have a flow path to the RHR pump compartment.

These RHR pump compartments are approximately 200 ft in size and will Page 55 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com completely fill in about 30 minutes at a flow rate of 50 gpm Additional information on the design of the RHR pump compartments, leakage detection and flow path are contained in the NMC response to Question 5.B.A passive leak within the RHR pipeway or one of the RHR pump compartments would result in the credible leak source flowing to one of the RHR pump compartments.

This would result in an RHR pump compartment high level alarm in the control room. The alarm would require that operators respond as directed by the associated alarm response book (ARB) procedure,w hich requires that the pump compartment be drained and the leakage source isolated, if possible, to prevent damage to the RHR pumps. Isolation of the passive failure would be accomplished by shutting the SI-850 and Sl-851 valves on the failed train of containment sump recirculation.

D. What are the consequences of leakage with regard to control room habitability for the limiting passive leak and where and when does leak this occur and what activity level is assumed during this leakage?What is the basis for this answer?NMC Response: For purposes of providing a limiting dose consequence due to a passive failure during recirculation post-LOCA, an evaluation was performed.

The approach and assumptions used are consistent with the current licensing basis radiological design basis LOCA analysis contained in FSAR 14.3.5 and RG 1.195, "Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors (May 2003)." The input used to estimate the dose consequences are delineated on Table 5.D-1.MethodoloQy The calculation methodology described in RG 1.195, Regulatory Position 2, was used to estimate the dose to the control room. Input values needed to complete the dose estimate were taken from FSAR 14.3.5, "Radiological Consequences of a Loss of CoolantA ccident." Values chosen for parameters not specifically identified in the FSAR were based on the guidance in RG 1.195. Core activities are based on a core power level of 1549 MWt, which is the current licensed power level including calorimetric uncertainty.

The thyroid dose conversion factors listed in FSAR 14.3.5, which are taken from Federal Guidance Report 11, were used. The whole body and skin dose conversion factors were taken from Federal Guidance Report 12 per RG 1.195,R egulatory Position 4.1.4. As further discussed below,d ecay of the activity in the sump is credited up to the point that the failure is assumed to occur. The release rate of the activity from the passive failure (Ci/min) is assumed to remain constant until the failure is isolated (i.e., removal processes such as decay is not taken into consideration).

However, determination of the integrated activity in the control room does credit decay and exhaust. No other activity removal processes are credited (e.g., plate-out, hold-up, ground deposition, etc.).Page 56 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com DBA Input and Assumptions Sump Coolant Source Term: Post-LOCA, 50% of the total core iodine is assumed to be in the sump coolanta vailable for recirculation.

All of the iodine released to the sump is assumed to be elemental.

This assumed chemical form is consistent with the current licensing basis LOCA radiological analysis.

Decay of the iodine activity in the sump coolant up to the point of the failure is credited.

No credit for decay of the iodine activity in the sump is applied after the passive failure is assumed to occur. At 30 days post-accident, only significantq uantities of 1-131 are remaining due to the relatively short half-lives of the other isotopes of iodine. Therefore, the only activity assumed to be in the sump is 1-131 based on a 200-day decay.Sump Volume: Consistent with the CLB LOCA ECCS leakage dose, the amount of coolanta vailable for recirculation is 197,000 gallons. However, it is expected that the amount of coolant available for recirculation would actually be 243,000 gallons. The increase in available sump volume is due to corrective actions taken since the licensing of the radiological LOCA analysis in 1997. At the time the LOCA analysis was under review by the NRC,i tw as assumed that coolant in the lower refueling cavity would not be able to drain to the "B" containment sump due to a component issue on the inlet to the cavity drain line. The cavity drain line has since been modified such thatc oolant in the lower refueling cavity can drain into the containment"B" sump and can be considered available for containment sump recirculation.

However, to maintain consistency with the analysis, credit for the additional volume of coolant is not taken into consideration.

Passive Failure Leak Rate, Occurrence, and Duration:

As discussed in the NMC responses to Questions 5.A and 5.B, the maximum passive failure leak rate is 50 gpm, which is postulated to occur "on the order of 200 days" following a loss of coolant accident.T herefore, the dose consequences are based on a passive failure leak rate of 50 gpm occurring at 200 days post-LOCA.

It is assumed that detection and isolation of this failure could take up to 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> after the onset of the failure. This assumed release duration is based on the identification of the gross leakage by loss of the suction to the decay heat removal pump without other detection methods as discussed in the NMC response to Question 5.C. However, the response to Question 5.C also states that detection of the passive failure could occur within a shift based on a gross containments ump level change. For conservatism, the more limiting detection/isolation time is assumed.Furthermore,s eat leakage past the isolated SI-850 valve is not taken into consideration because of the conservative passive failure leak rate and duration used to estimate the dose. In addition, once this failure is detected, measures would be taken to correct it.Page 57 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com Leakage Activity Release Fraction:

Consistentw ith FSAR 14.3.5, the fraction of iodine in the leakage that is released to the environment is 10%.As stated in the "Methodology" section above, no credit for plate-out, hold-up,o r filtration is assumed.Release Point: For purposes of assessing the limiting dose consequence due to a passive failure during the post-LOCA recirculation mode, a passive failure occurring in the Unit 2 tendon gallery is expected to be more bounding than a passive failure inside the PAB. This is due to the fact that under the worse case assumptions, itw ould take longer to identify a leak in the tendon gallery than inside the PAB. The release of activity from the tendon gallery is released directly to the environmentu nmonitored, whereas, a release from inside the PAB is readily detectable via either PAB sump level changes, area radiation monitors or vent stack radiation monitors.A release from the tendon gallery would be via the access hatches on El. 6.5' of the facades. The Unit 2 tendon gallery release is more limiting than the Unit 1 tendon gallery because the Unit2 tendon gallery has an access pointc loser to the intake of the control room ventilation system. Of the two tendon gallery access points in the Unit 2 facade, the more limiting release point is the access point near Pipeway 4. Since both Unit 2 tendon gallery access points are within the same wind direction sector, the access release point closestt o the control room intake results in larger atmospheric dispersion factors. Therefore, the bounding release point is the tendon gallery access point located in the Unit 2 facade at EI.6 .5' under Pipeway 4.FSAR Figures 1.2-5 and 1.2-12 illustrate the location of the tendon gallery access points.Atmospheric Dispersion Factors: The atmospheric dispersion factor (X/Q)associated with a release from the Unit 2 tendon gallery access hatch is 2.75E-03 sec/M 3.This dispersion factor was calculated using ARCON96 and is based on a point source, ground level release from the Unit 2 fagade near Pipeway 4. The cross sectional area of the facade is used to calculate the building wake. The X/Q assumed is that value calculated for the 0-2 hour interval post-accident to provide a bounding dose estimate.Control Room Occupancy:

The control room occupancy factor is assumed to be one (1) or 100% for the duration of the passive failure. During this phase of the accident, it is expected thato perators would be on 12-hour shifts. However, an occupancy factor of one (1) results in a bounding dose estimation.

Control Room Ventilation System Mode: As described in FSAR 9.8, the control room ventilation system has four modes of operation, whereby Mode I is the normal operating mode (outside air intake/recirculation) and Mode 4 is the emergency mode (filtered outside air intake/recirculation).

It is assumed that the control room ventilation system is operating in Mode I and remains in Mode 1 for the duration of the passive failure. Therefore, the Page 58 of 84 PDF created with pdfFactory Pro trial version www.pdffactory.com control room dose consequences are based on an unfiltered release. The dose consequence evaluation used an intake value of 1000 cfm, consistent with the FSAR 9.8 Mode 1 description.

Acceptance Criteria The dose acceptance criteria are the limits delineated in 10 CFR 100.11 and 10 CFR 50, Appendix A,c larified in NUREG-0800, Section 6.4, as well as the doses documented in FSAR 14.3.5, the licensing basis radiological design basis LOCA. Further discussion of the acceptance criteria is provided in the NMC response to Question 5.F.Dose Results: The thyroid dose to the control room operator based on the above DBA failure scenario is on the order of 0.06 rem. The whole body and skin doses are <0.0001 rem,a nd are therefore,n egligible.

This is primarily because 200 days provides a sufficient amounto f decay of iodine such that a release of activity to the environment would notr esult in a dose of any significance with regard to control room habitability.

Design Basis Dose Consequence Licensing Basis: Based on a historical review of the licensing bases, a passive failure as posed in FSAR 6.2 to be either excessive packing/weld leakage or RHR pump seal failure, has not been assumed to occur in conjunction with the radiological design basis LOCA analysis for purposes of demonstrating compliance with 10 CFR 100 or the dose limits of GDC-1 9. The current licensing basis radiological accidenta nalyses for LOCA is performed consistentw ith the approach used previously; namely maximum allowable containmentl eakage assuming failure of an emergency diesel generator resulting in one-train of containments pray and maximum allowable ECCS leakage. No additional failures are assumed during the recirculation phase.The most significant changes made to dose analysis for the recirculation leakage have been the assumed size of the leakage from ECCS. The assumed leakage from ECCS was based on the program limits defined by the Leakage Reduction and Preventive Maintenance program, which was developed in response to NUREG-0578, Item 2.1.6.a. This change was initially communicated to the NRC in the station's final response to NUREG-0737; Item III.D.3.4 dated September 4,1 984. In addition, the filtration capability of the PAB ventilation system was eliminated from the dose analysis via Technical Specification Change Request 192, which was subsequently approved by the Commission on July 9,1 997. Although the allowed operational leakage of ECCS has increased, the methods by which the recirculation portions of ECCS are maintained and tested have not changed.Page 59 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com Table 5.D-1 InputA ssumptions Used to Estimate Control Room Operator Dose Due to a Passive Failure in the Unit 2 Tendon Gallerv Post-LOCA Input Value Core Power (includes calorimetric uncertainty) 1549 MWTh Total Core Iodine 4.13E+07 Ci Fraction of Total Core Iodine in the Sump 0.50 Sump Volume 197,000 gal Passive Failure Leak Rate 50 gpm Iodine Re-evolution Release Fraction 10%Duration of Passive Failure Leak 60 hr Tendon Gallery -CR Atmospheric Dispersion Factor 2.8E-03 sec/m3 (0-2 hr)Height of Lower Instrumentation 10 m Control Room Parameters Breathing Rate 3.5E-04 m;/sec Occupancy I Control Room Volume 65,243 ft 3 Outside Air Intake 1000 cfm Filtered Outside Air Intake 0 cfm E. What are the consequences of leakage with regard to offsite dose for the limiting passive leak and where and when does this leak occur and what activity level is assumed during this leakage? What is the basis for this answer?NMC Response: The estimation of the offsite dose consequences due to the limiting passive leak (i.e., passive failure) follows the basis for the control room consequences as documented in the response to Question 5.D with two exceptions:O ne with regard to the atmospheric dispersion factors (X/Q), and the second with regard to assumed dose duration for the site boundary.Since the atmospheric dispersion factors are not release points pecific, there is no difference in offsite dose due to a passive failure in either the Unit 1 or Unit 2 tendon galleries or PAB. The offsite X/Qs represent an overall site dispersion of a release. Therefore, the 0-2 hour atmospheric dispersion factor for the site boundary (5.OE-04 sec/ M 3) and the 0-2 hr atmospheric dispersion factor for the low population zone (3.OE-05 sec/m 3)from the licensing basis LOCA radiological consequence analysis (FSAR 14.3.5) are used. The site boundary doses are calculated for the first two hours of the release, whereas, the low population zone doses are calculated for the duration of the release. Other source term assumptions and their bases as discussed in the Response to Question 5d remain the same for determining dose consequences to the offsite. Similarly, the Page 60 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com method presented in RG 1.195, Section 2,w as also used to estimate the offsite doses. Only decay up to the point of the failure is credited for reducing the source term in the containments ump.Dose to the Offsite Following a Passive Failure at2 00 days Post-LOCA:

The thyroid and whole body doses to the site boundary and low population zone based on the DBA passive failure scenario that occurs at 200 day post-LOCA are less than 0.001 rem; therefore, negligible.

This is primarily due the fact that 200 days provides a sufficient amounto f decay of iodine such thata ny release of activity to the environment would notr esult in a dose of any significance to the offsite. Acceptability of the results is discussed in the response to Question 5.F.F. Are the consequentialr adiation exposures within calculated results and regulatory limits? What is the basis for this answer?NMC Response: The passive failure dose consequences,a s well as, the current licensing basis radiological consequences documented in FSAR 14.3.5 for the control room habitability and offsite consequences and regulatory limits are provided in the Table 5.F-1 below. The use of the symbol "-" indicates no dose limit identified or that a dose is not required to be calculated.

The current licensing basis for PBNP control room habitability includes a factor of ten (10) dose reduction credit for the ingestion of potassium iodide (K!). The control room thyroid doses documented below include this credit.Credit for the ingestion of potassium iodide (KI) was nota pplied to the control room passive failure thyroid dose consequence.

Based on the table, it is seen that the doses due to the passive failure are within regulatory limits and bounded by the current licensing basis radiological design basis accident analysis.

The passive failure doses calculated for the control room are conservative since control room filtration is notc redited, 100% occupancy is assumed and the worst case meteorological conditions are applied for the duration of the accident.Page 61 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com Table 5.F-1 Dose Cons quences Location/Release Path Thyroid (rem) Whole Body Skin Regulatory Limit -CR 30 5 75*CLB CR -Total 29.27 1.37 43.18 Containment Leakage 18.60 1.366 43.14 ECCS Leakage 10.67 0.004 0.04 Passive Failure 0.06 4E-07 2E-05 Regulatory Limit -SB 300 25 _SB -Total 190.42 3.48 _Containment Leakage 133.3 3.24 _ECCS Leakage 57.12 0.24 _Passive Failure 4E-04 6E-08 Regulatory Limit -LPZ 300 25 _LPZ -Total 61.37 0.51 Containment Leakage 24.37 0.45 ECCS Leakage 37.0 0.06 Passive Failure 6E-04 1E-07*As defined in SRP 6.4, the skin dose limit is 30 rem, unless the licensee commits to use of protective clothing and goggles during a severe radiation release. Then the unprotected skin dose limit is not to exceed 75 rem. PBNP committed to maintain protective clothing and goggles in the control room in response to NUREG-0737, Item III.D.3.4, on February 23, 1981, and reconfirmed in letter dated September 4, 1984..G. What are the consequences of passive leakage and isolation capabilities with respect to ECCS functions (e.g., preservation of containment sump inventory to support post LOCA recirculation)?

What is the basis for this answer?NMC Response: A passive leak in the ECCS outside containment will be detected and isolated prior to the loss of containment sump inventory to the extentt hat core cooling capabilities will not be challenged.

As discussed in the response to Question 5.B, there is reasonable assurance that a postulated leak of 50 gpm in the containments ump suction line would be detected and isolated prior to loss of pump suction as a result of loss of containment sump inventory.

This is regardless of the leak location,w hether in the PAB or the tendon gallery.Flooding in the tendon gallery or PAB as a result of a passive failure in the RHR suction line will notp revent the ability of the ECCS system to perform its safety function for core cooling. Equipment lost in flooding of the tendon gallery would consist of the tendon gallery sump pump. Loss of this nonsafety-related pump has little effect on the ability to detect and isolate a passive failure as demonstrated in the NMC response to Question 5.B.Flooding in the PAB will likely result in the loss of one RHR pump prior to leak isolation but the other train RHR pump would be available to maintain Page 62 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com core cooling. Makeup water to the containments ump is available and is procedurally directed as a contingency if there is a loss of reactor injection flow as a result of inadequate containment sump performance.

H. Are the SI-850 valves credited with isolating a passive leak? If so,i s this a safety-related function?

If not,e xplain. What is the basis for this answer?NMC Response: The SI-850A(B) valves perform a safety-related function to isolate a passive failure in the containments ump recirculation line to prevent gross diversion of containment sump inventory.

The SI-850A(B) valves would be shut to support the following post-accident functions following a credible leak in the containmentr ecirculation line:* Maintain Sump "B" inventory* Protect the RHR system and pumps from flooding The shut safety-related function is discussed in FSAR Chapter 6.2.2 where it states: "Each recirculation sump line has two remotely operated valves. The first valve is located adjacent to the end of the pipe in the containments uch that the line inside the containment can be isolated in the event of a passive failure." In accordance with 10 CFR 50.2, the ability of the SI 850A(B) valves to isolate a passive failure is classified as a safety-related function.

The ability of the S18 50A(B) valves to isolate a passive failure supports Criteria 2 and 3 for a safety-related component.

Shutting these valves to isolate a passive failure prevents the gross diversion of containment sump inventory and ensures that at least one redundant train of long-term core cooling remains operable throughout the post-accidentp hase. Long-term decay heat removal is essential to maintain the plant in a safe shutdown condition and to ensure offsite doses are maintained within the limits of 10 CFR 100 and control room doses are within the limits of 10 CFR 50 Appendix A, GDC-1 9.PBNP is designed to withstand the maximum credible leakage of 50 gpm from the containment sump recirculation lines and the RHR system without a loss of capability to shutd own the reactor and maintain it in a safe shutdown condition.

PBNP did not previously consider that the SI-850A(B) valves performed a safety-related function in the shutp osition. A corrective action program document was initiated in response to NRC inspection questions during the November 2005 inspection.

Although not previously evaluated againsta cceptance criteria, previous test data taken on these valves in the shut direction for trending purposes were within the bounds of the ASME Code-required acceptance bands. This Page 63 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com demonstrates that the valves are fully capable of performing a safety-related function to shut. Since it was determined that the Sl-850A(B) valves perform a safety function to shut, the PBNP IST program documenth as been updated to reflect this function and the associated IST implementing procedures (IT 40 and 45) have been revised to include acceptance criteria for shutting the valves. These revised procedures have been implemented.

The four SI-850A(B) valves have had satisfactory test results.Please refer to the NMC response to Question 5.J related to isolation of a containments ump recirculation line following a passive failure for dose considerations.

If the SI-850 valves are credited with isolating a passive leak, explain how much this valve will continue to leak after closure and how this leak rate was determined.

If this leak rate has not been measured, explain what a limiting leak rate would be and your basis for this leak rate. What is the basis for this answer?NMC Response: The Sl-850A(B) valves are credited with isolating a passive leak in the containments ump recirculation lines. PBNP does not perform a seat leakage test on the valves in the direction of the containment sump to the recirculation lines. Based upon the design of the valves and their operating conditions, PBNP expects the valves to limit a passive pressure boundary failure sufficiently preventing a gross diversion of water from the containments ump.Although not previously credited,N MC has determined that the Sl-850A(B) valves perform a safety-related function to shut and isolate a passive failure.PBNP does not perform a seat leakage test on the SI-850 valves in the direction of the containment to the containment recirculation line. Therefore, PBNP does not have qualified seat leakage data on these valves.The safety-related function to shut is to isolate a passive failure in the RHR containments ump recirculation line. The design leakage rate of the passive failure is bounded by a 50 gpm leak. Shutting the SI-850 valves will reduce the bounding leakage rate of 50 gpm. A specific maximum seat leakage rate is not required to reach a manageable leak rate with respect to maintaining the decay heat removal function, as PBNP is designed to withstand the bounding 50 gpm leak rate. Based on this, the SI-850 valves were not intended to meet the requirements to be classified as Category A valves per ASME OM Code,P aragraph ISTC 1.4(a), which states: "Category A -valves for which seat leakage is limited to a specific maximum amounti n the closed position for fulfillment of their required functions." Page 64 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com The valves are, however, required to prevent gross diversion of water through a passive failure in the containments ump recirculation line and are classified as Category B valves. Per ASME OM Code, Table ISTC 3.6-1,"In-service Test Requirements," seat leakage testing is not required for Category B valves.While the PBNP licensing/design basis limits the leakage from this passive failure in the RHR to less than 50 gpm, the design of the SI-850 valves is expected to significantly reduce the leakage rate when they are shut.T he SI-850 valves are equipped with a resilient (soft) seat. Resilient seats are used to accomplish good seating performance with much lower contact force than is required in metal-to-metal seats. In the case of the Sl-850 valves, the resilient seat is formed by an O-ring and provides the primary seating seal with the metal-to-metal closure acting as a secondary seal.Based on ANSI B 16-104,A merican National Standard for Control Valve SeatL eakage, the allowable seat leakage for a valve with the design of the SI-850 valves (Class VI) would be approximately 15 ml/min atm aximum rated differential pressure.

While nota n element of the PBNP licensing basis, ANSI B 16-104 is an industry standard used to determine expected leakage of resilient seals. A review of the forces on the containment sump"B" isolation valves concluded that adequate sealing forces are applied for the O-ring to provide and adequate seal as ascertained in Engineering Evaluation 2006-0003.

Although the soft-seated design of the SI-850 valves would be expected to control seat leakage to a very nominal rate, as they are classified as Category B valves per ASME OM Code, no seat leakage testing is performed on these valves to quantify this leakage rate. Based on the above discussionP BNP would expect a shut Sl-850A(B) valve to prevent the gross diversion of water from the containments ump through the containments ump recirculation line.J. Was the continued leakage past the shut SI-850 valve considered in calculation of controlr oom dose, off-site dose or preservation of containment sump inventory?

If not, explain. What is the basis for this answer?NMC Response: As discussed in response to Question 5.D (calculation of control room dose), seat leakage past the isolated sump suction isolation valve, Sl-850, was not taken into consideration as a result of the conservative passive failure leak rate and duration used to estimate the dose consequences.

Since the calculation of offsite dose (response to Question 5E) used the same methodology as the control room dose calculation, the exclusion of seat leakage has the same basis.As stated in the NMC response to Question 5.G, makeup to the containment sump is available and is procedurally directed as a contingency if Page 65 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com containments ump performance is identified as a concern. Once a passive failure is identified and isolated, leakage past the shut containment sump isolation valve is negligible (refer to NMC response to Question 5.1).Page 66 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com ENCLOSURE 2 ATTACHMENT I The following attachments to Enclosure 2 are provided to assist in the review of the NMC response to the RAI: Attachment 1 Pages 1-3 Pages 4-7 2 Pages 1-3 Pages 4-7 3 4 5 Description Unit 1 Delaminating Qualified Coatings List Unit 1 ContainmentE levations Showing Degraded Or Nonconforming Coatings Unit 2 Delaminating Qualified Coatings List Unit 2 ContainmentE levations Showing Degraded Or Nonconforming Coatings 1(2)SI-850A(B)

SIS Drains Elevation Sketch 1(2)SI-850A(B),S IS Drains Plan 1 (2)SI-850A(B)

Valve Page 67 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com ENCLOSURE 2 ATTACHMENT I Unit 1 Delaminatinq Qualified Coatinqs_ _ __ APage 1 of 3 Map Location Area (ft 2) Description Reactor Cavity 1 l Reactor cavity l 10 Delaminated and cracking coatings on walls Subtotal 10 8' Elevation General Area 2 Az45, El 10 2 on ICW 3 Az 85 3 ICW, delamination 4 Az 90, El 20 8 Penetration 28, light rust 5 Az 140 12 WTUP on 3 columns and ceiling -qualified by adhesion test.S omeflaking 6 Az 150 0.5 By LP,s upport peeling atb olts 7 Az 170 8 Flaking topcoat on 2 columns. In general, steel embeds have red or zinc primer with white topcoat. Concrete has green surface (up to 1/8-inch thick), with white, gray, and white intermediate and top coats. Most areas appear tight, with a few areas having delaminating topcoat 8 Az 200 8 Delaminating topcoat on column 9 Az 200 1 On column, degraded coating over unprepared steel embed 10 Az205 3 Inner concrete column, total failure of coating; no adhesion of base coat 11 Az 235 6 ICW, support with poor application over red primer, loose and chipping off 12 Az 245, El 14 0.5 LP, topcoat loose and chalky 13 Az 270 3 In cubicle opening, delaminated concrete coating 14 Az 310 3 On column near LP, horizontal pipe member of support, poorly done over red primer 15 Az 357 1 Keyway wall, checking and delamination.

Subtotal 59 A Steam Generator Cubicle 16 1st Level 6 SG support struts. -Most loose coatings removed.17 1st Level 16 Northeast corner column, cracked and delaminated 18 3rd Level 2 Northwest wall.19 3rd Level 10 Northeast wall.20 3rd Level 20 East wall, degraded coatings on a penetration through the east wall towards the reactor.21 4th Level 10 South wall 71 5th Level 40 East wall 72 5th Level 30 South wall 73 5th Level 10 Floor, south 74 5th Level 40 West wall Subtotal 184 Page 68 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com ENCLOSURE 2 ATTACHMENT I Unit I Delaminating Qualified Coatinqs Page 2 of 3 Map # Location Area (ft 2) Description A Reactor Coolant Pump Cubicle 22 1 st Level 5 On the RCP support struts. -most loose coatings removed.23 3rd Level 8 South wall 24 4th Level 8 South wall cracking and delamination 25 5th Level 4 Top of the Upper Oil Cooler and pipe 26 5th Level 10 Slab joints/ledge above, delaminating 27 5th Level 18 South wall, a steel structure for HVAC -loosely adherent coating, easily removed 28 5th Level 5 Southwest wall, cracking, delaminating, WTUP.Subtotal 58 B Steam Generator Cubicle 29 Snubber Level 25 North wall d elaminating 30 Snubber Level 25 East wall, delaminating 31 Snubber Level 50 South wall, delaminating

-bad surface prep 32 Snubber Level 25 West walld elaminating Subtotal 125 B Reactor Coolant Pump Cubicle 33 1st Level 3 North, scratch, column delaminated 34 1 st Level 1 North openingl arge blister on top part 35 3rd Level 10 Oil pipes, degraded, poor surface prep (shiny or mill scale)36 3rd Level 4 South wall, cracking and delamination 37 3rd Level 8 North side, cracked, delaminating, embed 38 3rd Level 14 Northeast, cracked and delaminating 39 4th Level 12 North wall,c racks and delamination 40 4th Level 6 East wall, cracks and delamination 41 4th Level 6 South wall, cracks and delamination 42 5th Level 6 Northeast wall, cracking and delamination 43 5th Level 6 South wall, cracking and delamination 44 5th Level 6 West wall,c racking and delamination Subtotal 82 Pressurizer Cubicle 45 Top level 16 North walld elaminating.

Bad surface prep 46 Top level 12 East wall, orange-tan touchup, checking, delamination 47 Top level 6 Southeast wall, delamination,c hecking and cracking 48 Top level 10 South wall, delaminating and cracking 49 Top level 6 West wallc hecking and delamination 50 Base 42 Floor 35% abraded area about 120 ftW. Condition not good 51 Cubbyhole 19 Walls, ceiling Subtotal 111 Page 69 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com ENCLOSURE 2 ATTACHMENT I Unit I Delaminating Qualified Coatings Page3of3 Map # Location Area (ft 2) Description 21' Elevation General Area 52 Az 45 1 Delaminating coating on support column for 1W1 C1 -bad surface prep 53 Az 92,E 1 22 8 LP, penetration 34, light rust at welds 54 Az 130 1 By LP, steel column with delaminated topcoat, zinc primer intact, no rust 55 Az 210 to 249 100 ICW, cracked/delaminated coating 56 Az 310 50 ICW, cracked/delaminated coatings & WTUP Subtotal 160 46' Elevation General Area 57 Az 40 2 Penetration 27 -medium rust 58 Az 42,E 1 49 3 On penetration through the ICW toward the reactor cavity (NE-1 33 or M-300-7-1) 59 Az 49,E 1 60 10 delaminated coating on the LP, no rust 60 Az 150 4 By LP, steel columnd elaminating coating, applied over dirt or grease?61 Az 230, 4 Halfway downstairs, lCW chipping, grout holes 62 Az 245 10 Cracked and peeling on inner wall 63 Az 259 10 Cracked and peeling coating on the inner concrete wall, especially by the embeds Subtotal 43 _66' Elevation General Area 64 Az 66 10 Cracks and peeling coating on the inner containmentw all 65 Az 90 2 ICW, up high, cracked and delaminated 66 Az 105-135 120 400 ftWarea, floor coating between the hatch open area and the inner concrete walls is 30% cracked and abraded, coating is not tight and chips easily -CAP029629, WO 0212790 67 Az 115 1 ICW, delamination 68 Az 120 10 ICW, cracked and peeling paint, WTUP, orange touchup on the southwest wall of 1HXIB 69 Az 265 1 ICW delaminated 70 Az 275 20 ICW, delaminating and peeling coating, especially around the embeds Subtotal 164 Total 996 Page 70 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com ENCLOSURE 2 ATTACHMENT 1 Unit I Containment Elevations Showinci Degraded Or Nonconforminq Coatinqs A7TACHW~NT I PACE 4 z 1u61! I1 COMWU~Isde EL. SW'O*-CONr I out- .rIfn W2J1,f2C-,5 I :U/:32 PM Page 71 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com ENCLOSURE 2 ATTACHMENT 1 Unit I Containment Elevations Showina Degraded Or Nonconforminq Coatinqs ATTACHMENT I PACE 5 UNIT 1 CONJTA INENT EL. 21'-O-E?3t% 1253t- IJqfn 2i~b2C---

1 :tiS:34 p1.Page 72 of 84 PDF created with pdfFactory Pro trial version www.pdffactory.com ENCLOSURE 2 ATTACHMENT I Unit I Containment Elevations Showinq Decraded Or Nonconforminq Coatinqs A7TACHMENT I PAGE 6 z _,, LO! I CCOTA10"1 C.ON 1441- .rirn 2 I I. I:i I:U.' VM.Page 73 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com ENCLOSURE 2 ATTACHMENT I Unit I Containment Elevations Showing Degraded Or Nonconforminq Coatings ATTACHPENT 1 PAGE 7 UITI I CONfTAINW~10 IL. "'-O't~3N 165* i.tIqn 1J~C. I:1-j:3S PM Page 74 of 84 PDF created with pdfFactory Pro trial version www.Ddffactorv.com ENCLOSURE 2 ATTACHMENT 2 Unit 2 Delaminating Qualified Coatings Page 1 of 3 Map Location Area (ft 2) Description Keyway I Access shaft 40 LP, rust dripping and loose joint material at the horizontal transition joint, concrete to the LP 2 Floor (entire) 190 Approximate 380 sf area, 50% delaminated 3 Floor (entire) 100 LP, debris strewn. Require cleaning for proper inspection 4 Base, Sump A 100 LP, standing water with dirt/debris, condition of the floor was inaccessible.

No obvious evidence of rust on floor.5 Tunnel 100 LP, southwest, floor, debris pile at the kick plate separating the tunnel from the access shaft. The kick plate is not sealed. Debris has paint chips in it.6 Reactor room 30 Tunnel opening,S outheast wall.C oncrete top coat delaminating 7 Reactor room 30 Concrete wall. Delaminating concrete coating at a construction joint Subtotal 590 8' Elevation, General Area 8 Az 149, El 20 4 LP, service water penetration P08, light rust 9 Az 240, El 16 1 East face of Sump A shaft, delamination over steel embed 10 Az 270 15 Entry to SG cubicle, delamination of white touch-up.Subtotal 20 A Steam Generator Cubicle 11 Entryway, El 12 20 East of the East wall, tape residue and degraded concrete coating on the Reactor wall and the East wall 12 Base 185 Walls, along perimeter Delaminating and cracked coating distributed on all walls 13 2nd L 70 Walls, cracked and delaminated coating 14 3rd L 270 Walls, cracked and delaminating coating and touch-up Subtotal 545 A Reactor Coolant Pump Cubicle 15 Base, El 20 110 Walls, along a perimeter.

Delaminating and cracked coating distributed on all walls 16 2nd L 70 Walls concrete coating, delaminated, and white touch up 17 3rd L 210 All walls, distributed.

Concrete coating delaminated, and touch up -no adhesion of base coat 18 3rd L 55 RCP, top portion of the bottom half. Degraded coating, easily removed, apparently noti nsulated 19 3rd L 10 RCP, bottom flange of the top half, flange perimeter is degraded 20 3rd L 50 East wall, cracks and delamination 21 4th L 270 Walls, concrete coating, delaminated, and touch up Subtotal 775 .-Page 75 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com ENCLOSURE 2 ATTACHMENT 2 Unit 2 Delaminating Qualified Coatings Page 2 of 3 Map# Location Area (ft 2) Description B Steam Generator Cubicle 22 Base 165 All Walls (not as bad as 'A" cubicles)23 Base 2 2 large columns with hairline cracking and delamination at top corners 24 2nd Level 110 All walls, delamination 25 3rd Level 195 EastW all 26 3rd Level 60 South Wall 27 3rd Level 10 West Wall 28 4th Level 200 North wall coating in very poor condition 29 4th Level 500 East and notch wall coating in very poor condition 30 4th Level 70 South & West walls Subtotal 1312 B Reactor Coolant Pump Cubicle 31 Base 2 2 large columns with hairline cracking and delamination at top corners 32 2nd Level 105 All walls, delamination 33 4th Level 60 All walls 34 Top L 35 Concrete wall coating, delaminated top coat Subtotal 202 Pressurizer Cubicle 35 Top Level 3 Spalled concrete and degraded coating 6 to 7 feet below access opening 36 Top Level 1 Degraded coating on wall at top of ladder 37 Mid Level 20 Wall coating, delaminating Small platforms 38 Bottom Level 50 Wall touch up, grout holes and delaminations 39 Bottom Level 3 Fire damage near door to RCP Cubicle Subtotal 77 21' Elevation General Area 40 Az 0 to 30 10 ICW, cracked and delaminating coating 41 Az 12 35 Head laydown stand. Steel coating is severely degraded.Concrete coating appears tight 67 Az 90,E 129 2 ICW, (north wall), 4 sf grout holes & 2 sf degraded concrete 42 Az 145, El 28 10 LP, 6 of the 36 penetrations have light to medium rust and/or degraded coating. CAP064095, WOs 0501976, 0501977, and 0501978 43 Az 148 to 153 5 LP, SW pipe through 2CPP-45&46 has heavy rust bleeding through the coating (originally

-18 sf, 11 sf removed)44 Az 250, El 25 20 Wall, coating, delaminating 45 Az 260, El 29 10 Wall coating, delaminating Page 76 of 84 PDF created with pdfFactory Pro trial version www.pdffactory.com ENCLOSURE 2 ATTACHMENT 2 Unit 2 Delaminating Qualified Coatings Page 3of 3 Map Location Area (ft) Description 211 Elevation General Area (continued) 46 Az 266, El 26 15 By LP,c avity cooling valve area. Most of the coating appears tight, some areas have rust bleeding through coating CAP051481,W 0 0309879 Subtotal 107 46' Elevation General Area 47 Az 0 to 10 6 Floor area abraded and delaminating 48 Az 110, El 60 10 ICW, face of B RCP East wall: Wall, grouth oles, degraded coating, degraded supports 49 Az 115 20 By LP, floor delaminating.

Failure of the concrete itself, not just the coating 50 Az 135, El 58 1 ICW, delaminated concrete top coat 51 Az 148, El 63 2 ICW, floor to ceiling line, degraded (light rust) floor________penetration 52 Az 158, El 51 10 ICW, face of B SG Southeast wall: Degraded coating on 2AC12, 2AC13 & grout holes 53 Az 225 to 269 60 Floor area 305 sf, 20% abraded -POOR ADHESION 54 Az 255, El 46 4 Floor and steel surrounding insulated HB-1 riser; poor adhesion, chipping off 55 Az 300 to 320 10 ICW, long horizontal crack with delamination 56 Az 309, El 51 6 ICW, delaminating concrete coating, with tiny cracks 57 Az 310, El 58 10 ICW, degraded concrete coating 58 Az 310, El 48 2 ICW, 2 large circular areas delaminating Subtotal 141 66' Elevation, General Area 59 Az 0, El 74 1 ICW 60 Az 70,E I 99 10 LP, penetration V02. Control equipment 61 Az 89,E 1103 2 By LP,C rane access platform, Crane rail girder support by platform 62 Az 90,E I 99 10 LP, penetration V01, control, (1AT402)63 Az 240 20 EastW all of 'A" SG 64 Az 293, El 97 120 Top of South A SG wall, delaminating concrete 65 Az 325 2 1CW 66 Az 300, El 115 6 LP, few large blisters over globs of grease or dirt Subtotal 171 j Total 3940 Page 77 of 84 PDF created with pdfFactory Pro trial version www.pdffactory.com ENCLOSURE 2 ATTACHMENT 2 Unit 2 Containment Elevations Showing Degraded Or Nonconforming Coatings ATTACHMENT 2 PACE 4 z EL- 8" O tCON I BI- .vn w1iQCn l :Um:d' PM Page 78 of 84 PDF created with pdfFactory Pro trial version wwv/.pdffactory.com ENCLOSURE 2 ATTACHMENT 2 Unit 2 Containment Elevations Showinq Degraded Or Nonconforming Coatings ATTACHMENT 2 PAGE 5 UNIT 2 CONTA I NMENT EL. 21'-O" C.ON I2~1- .rirln 201 I.51:L1.:9 PM Page 79 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com ENCLOSURE 2 ATTACHMENT 2 Unit 2 Containment Elevations Showing Deqraded Or Nonconforminq Coatinms ATTACHM~ENT 2 PAGE 6 z _,6 EL. AV-00 Utyx 1441-.Ifl t lj 2Cn I:1=:2 PM.Page 80 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com ENCLOSURE 2 ATTACHMENT 2 Unit 2 Containment Elevations Showinq Degraded Or Nonconforming Coatings ATTACHMENT 2 PAGE 7 NImI 7 CcWNTAINaCNT EL. 66f -0 U01, I 65t- .dnn WI21~1 2C~I:1hC R.1 Page 81 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com ENCLOSURE 2 ATTACHMENT 3 SIS DRAINS -ELEVATION©~C I1..'44 E;* I- '-!S1& LT P:1 ~-': VALVE 5r,, FXA ,-v.?.'YL I:*~AUXILIARY El .(-)10'.4-VI~ CR.AINS-E-LEVATIZN Page 82 of 84 PDF created with pdfFactory Pro trial version www.pdffactorv.com ENCLOSURE 2 ATTACHMENT 4 SIS DRAINS -PLAN AUXILIARY BUI LDING Q CONTAlNM.ENT ,FINE SCREEN (1/8" MESS1)COARSE SCREEN I CDNTAINKhiEE DRXAS -MAN Page 83 of 84 PDF created with pdfFactory Pro trial version www.pdffactory.com ENCLOSURE 2 ATTACHMENT 5 1(2)SI-850A/B Valve El. 8'-O" Valve Operator (Located In PAS)Valve Cylinder Page 84 of 84 PDF created with pdfFactory Pro trial version www.Ddffactorv.com ENCLOSURE 3 CD-ROM CONTAINING REQUESTED RECORDS This enclosure provides an index to the CD-ROM which contains the records that were requested in the Request for Additional Information (RAI), as well as additional records identified by NMC that supportr eview of the RAI: 1. Gibbs & Hill, Inc., "Evaluation of Paint and Insulation Debris Effects on ContainmentE mergency Sump Performance," (Unit 1), dated May 18, 1989.2. Gibbs & Hill, Inc., "Evaluation of Paint and Insulation Debris Effects on ContainmentE mergency Sump Performance," (Unit 2), dated August 1, 1990.3. Sargent & Lundy Calculation M-09334-345-RH-1, "Containment Sump Blockage Due to Failure of Unqualified/Undocumented Coatings (Unit 1)," Revision 0, approved June 4, 1998,a nd Revision 1, issued January 21, 1999.4. Sargent & Lundy Calculation M-09334-431-RH-1, "Containment Sump Blockage Due to Failure of Unqualified/Undocumented Coatings (Unit 2)," Revision 0, issued January 1, 1999.5. Operability Recommendation OPR 161, "Containment Coatings Not Maintained Within Analyzed Limits," Revision 1.6. Operability Recommendation OPR 162, "Question with the Ability of ECCS Sump Screens To Pass Required Flow," Revision 2.7. Operability Recommendation OPR 164, "Wax Found On Two Elevations On Floors In Unit 2 Containment," Revision 0.8. Operability Recommendation OPR 165, "Tendon Gallery Inspection Results," Revision 0.9. Operability Recommendation OPR 170, "Design Basis Leakage Detection Capability May Have Been Defeated," Revision 1.10. Operability Recommendation OPR 171, "Safety Function for Containment Sump B Isolation Valves," Revision 0.11. Engineering Evaluation 2005-0024, "Evaluation of ContainmentS ump Screen Debris Buildup Based on EPRI (Electric Power Research Institute)

Technical Report and Current Degraded Epoxy Inventories," Revision 1, dated November 9, 2005.Page 1 of 2 PDF created with pdfFactory Pro trial version www.rdffactorv.com ENCLOSURE 3 CD-ROM CONTAINING REQUESTED RECORDS 12. Engineering Evaluation 2006-0003, 'SI-850 Valve Closure Forces and O-Ring Leakage," Revision 0, dated February 7,2 006.13. PointB each Nuclear Plant CAP068442, "Generic Letter 98-04 Commitments;" linked to CAP 068373, "ContainmentC oatings Not Maintained within Analyzed Limits." Root Cause Evaluation 294 is pending as of February 16, 2006.14. EPRI Technical Reportl 011753, "Design Basis Accident Testing of Pressurized Water Reactor Unqualified Original Equipment Manufacturer Coatings." 15. 10 CFR 50.59 Screening, SCR 2005-0260, "Revisions to Procedures EOP-1, EOP-1.3 and EOP-1.4 to Change RHR Flow Limit During Sump Recirculation." Page 2 of 2 PDF created with pdfFactory Pro trial version www.pdffactorv.com Nuclear Management Company Page I of 6 State Change History Initiate by OMILLIAN, MICHAEL AR Pre-Screen 10127/2005 11:03:40 PM Owner (None)Submit to Screening Team by BELTZ, BOYD AR Screening Que 10131/2005 9:59:29 PM Owner PBNP CAP Admin Screening Update by KREIL, JULIE AR Screening Que 11/1/2005 2:48:54 PM Owner BENNETT, KEVIN Section 1 Activity Request Id: Activity Type: One Line

Description:

Detailed

Description:

CAP068373 CAP Submit Date: 10/27/2005 11:03:40 PM Containment Coatings not maintained within analyzed limits 10/27/2005 11:03:40 PM -OMILLIAN, MICHAEL: While researching the current status of containment coatings on both units, it was determined that they have not been maintained within the envelope evaluated by the analysis of record (S&L Calculations M-09334-345-RH.1 and M-09334-431 for units 1 and 2 respectively).

When degraded or non-conforming coatings in excess of those evaluated in the analyses have been identified and not remediated, the applicable portions of the existing analysis were re-performed to determine the potential impact on ECCS sump performance.

However, this re-performance has been informal, is not documented, lacked an independent review, and prevented the non-conforming conditions from being recognized in the corrective action program.The current situation for Unit 2 has been reviewed to determine whether an immediate challenge to operability exists. While outside of the previously analyzed limits for the quantity of coatings debris, it was found by the initiator of this CAP that the operability of the Unit 2 ECCS sumps is not challenged by this new information.

Further details are summarized in the attached WORD document.10/29/2005 5:51:53 PM -OMILLIAN, MICHAEL: Based on information obtained during consultations on 10/28 and 10/29/05, it has become apparent that the head loss correllation used in the subject calculations has not been validated for particles alone (without fibers), and not for particles the size of the assumed paint chips.Because of this, the conclusions of the calculations as they pertain to head loss due to the assumed paint chips are considered discredited and not valid.However, concurrent discovery has also found that unqualified coatings (i.e. alkyd based)determined that such coatings have been found to fail to particles of -10 to -100 microns (i.e.microscopic "dust" or "fines").

Such small particulates, while completely transportable, would not cause any blockage of the screen holes.The balance of the unconforming coatings are previously qualified epoxies that have been damaged, abraded, chipped, etc. None of these located within the tranporable envelope ("Zone of Influence")

are debonding, and all are located on the floors. Such damage is not expected to result in the production of chips, and even if chips did result, they would either be too small to cause screen blockage (i.e. less than 1/8"), or would be too large to transport (i.e. the flow velocities are too low to move them across the floor). The ZOI for epoxy is for coatings that are located above the floor and could therefore be moved to the sump screens as they sink through the moving water column.Based on this new information, the deficiencies noted in the S&L calculation methodology presents no challenge to the functioning of the ECCS sumps. A full, formal Engineering Evaluation is being prepared to address these issues more fully, and will include such additional factors as other miscellaneous debris (specifically labels that may be present), and potential downstream effects (i.e. potential impacts on pumps, valves, core internals, etc.).However, con KENDALL, THOMAS Initiator Department:

Initiator:

Date/Time of Discovery:

Identified By: Equipment#

(1 st): Equipment

1. (2nd): Equipment

1. (3rd): Site/Unit:

10/27/2005 9:25:00 PM NRC (None)(None)(None)Point Beach -Common Date/Time of Occurrence:

System: Equipment Name (1st): Equipment Name (2nd): Equipment Name (3rd): EP Engineering Programs PB 10/27/2005 9:25:00 PM CONT PB (None)(None)(None)http://enwsO2/tmtrackltmtrack.dll?lssuePage&Template=printitem&recordid=897686&tab

... 2/15/2006 Nuclear Management Company Page 2 of 6 Why did this occur?: 10/27/2005 11:03:40 PM -OMILLIAN, MICHAEL: Undetermined at this time.Immediate Action Taken: 10/27/2005 11:03:40 PM -OMILLIAN, MICHAEL: Immediately informed Engineering Management and collected the most recent data from Unit 2 (the operating unit). Independently re-performed the ECCS sump performance calculations to verify that the sump screens are not jeopardized by the identified increases in unqualified coatings.A similar effort is being performed for Unit 1 to determine the extent of remedial actions necessary prior to unit re-start.Recommendations:

10/27/2005 11:03:40 PM -OMILLIAN, MICHAEL: 1) Perform a formal OPR to more fully document the basis for operability of Unit 2 2) Implement corrective actions to remediate the coatings in Unit 1 prior to unit restart.3) Perform an ACE or RCE (as deemed appropriate) to determine the cause of and appropriate corrective actions for this event.4) Remediate coatings to the degree necessary to restore the units within their analyzed limits and/or revise the analysis.

Note that at the end of the current fuel cycle for each unit, these analyses will be superseeded by pending licensing and design efforts to resolve issues under GSI-191.10/28/2005 4:34:05 AM -BELTZ, BOYD: CAP changed to SRO Yes, due to potential concern with unit 2 and need for an OPR.SRO Review Required?:

Y Section 2 Operability Status: Basis for Operability:

Operable but Non-Conforming Compensatory Actions: N 10/27/2005 11:35:57 PM -BELTZ, BOYD: Based on discussion with Engineering and attached evaluation of current status of coatings within unit 2 containment.

Unit 2 Sump B screen is capable of performing required function.

The original analysis did not account for the lower cavity debris screen and surplus of RWST available.

taking into account these two items overall headloss decreases, allowing the ECCS Train to operate further from a limit.Based on this an OPR is being requested for Unit 2, to fully address and formalize (document) coatings concern within containment and potential impact on unit 2 ECCS Train.OPR is not required for unit 1 as a mode restriction is in place and issue will be resolved prior to resuming power operation.

OPR is due 10130/2005 at 2300 with T. Kendall assigned.10/29/2005 11:37:10 PM -BELTZ, BOYD: Based on more accurate information obtained during consultations with G. Zigler of Alion Science, and J. Cavallo of Corrosion Control Laboratories, the initial prompt operability concerning the coatings issue in Unit 2 containment needs to be evaluated.

The initial determination and information supplied by engineering assumed the unqualified coatings would cause blockage and a debris pile and therefore affect the head-loss across the debris screen and subsequently ECCS Train operability concerns The new information reveals that the particle size would be such that they would become completely transportable ( i.e. pass thru the debris screen and therefore no blockage ). The balance of the un-conforming coatings are previously qualified epoxies that have been damaged, abraded, chipped, etc. None of these located within the transportable envelope ("Zone of Influence')

are debonding, and all are located on the floors. Such damage is not expected to result in the production of chips, and even if chips did result, they would either be too small to cause screen blockage (i.e. less than 1/8"), or would be too large to transport (i.e. the flow velocities are too low to move them across the floor). The new information removes concern with fouling / clogging of the debris screen.Due to the information showing screen fouling is no longer the major issue of concern the operability of components down stream of the debris needs to be addressed due to the particle size and transportability of the particles.

The flow path in question is large diameter piping, containing no filters or in-line strainer devices ( other than debris screen ) that may become fouled. The piping construction is also of a material that soft materials would not adversely impact or degrade operation.

Based on engineering review and input concerning industry experience with wear products in primary systems, the soft pliable particles that may be generated due to Sump B coatings becoming displaced would not affect the ability of the ECCS Train to operate and perform its required function during accident conditions.

The pumps it is reasonable to assert that a 0.13% "slurry" would have insigificant hydraulic effects on the ECCS pumps. The remaining consideration is affect on heat transfer and disposition of particles on surfaces, based on engineering input the amount of debris in solution would not impede the heat exchangers from http://enwsO2/tmtrack/tmtrack.dll?lssuePage&Template=printitem&recordid=897686&tab

... 2/15/2006 N, uclear Management Company Page 3 of 6 removing the required decay heat.Based on this new information the operability call of operable remains unchanged even though new information and concerns have been identified, however the OPR is still needed to address these issues more fully, and will include such additional factors as other miscellaneous debris (specifically labels that may be present), and potential downstream effects (i.e. potential impacts on pumps, valves, core internals, etc.).The OPR-161 due date remains unchanged 10/30/05 at 2300.10/30/2005 8:10:10 PM -BELTZ, BOYD: Based on further discussion and complexity of issue along with volume of information required to assemble in a logical manner the OPR due date is being extended 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 10/31/05 at 2300.Research to this point has not identified any new issues to question the prompt operability call associated with this CAP.10/31/2005 9:59:29 PM -BELTZ, BOYD: Based on review and approval of OPR-161 which also addressed the label issue. PBNP Containment Sump B Screens fully capable of performing their safety function, however we are non-conforming to our license basis due to methdology and communication to the NRC in regards to GL 98-04. reference engineering eval 2005-0024.

Unplanned TSAC Entry: N External Notification:

N Section 3 Screened?:

Y Significance Level: A INPO OE Reqd?: N Potential MRFF?: N QA/Nuclear Oversight?:

N Licensing Review?: N Good Catch/Well Doc'd?: NA Section 4 Inappropriate Action: 12/27/2005 1:17:05 PM -BENNETT, KEVIN: None.Process: N/A -Not Applicable Activity:

N/A -Not Applicable Human Error Type: N/A -Not Applicable Human Perf Fall Mode: N/A -Not Applicable Equip Failure Mode: (None) Process Fail Mode: N/A -Not Applicable Org/Mgt Failure Mode: N/A -Not Applicable Group Causing Prob: (None)Hot Buttons: (None)Section 5 CAP Admin: Project: Active/inactive:

Owner: Last Modifier: Last State Changer: NUTRK ID:# of Children: BENNETT, KEVIN CAP Owner: Corrective Actn Program (CAP) AR State: Active Submitter:

BENNETT, KEVIN Last Modified Date: REBITZ, EMMY Last State Change Date: BELTZ, BOYD Close Date: SHERWOOD, GARY AR Screening Que OMILLIAN, MICHAEL 2/14/2006 10:00:12 AM 10/31/2005 9:59:29 PM 0

References:

CAP068442 Update: Prescreen Comments: Import Memo Field: OPR Completed?:

N OLDACTIONNUM:

subtsid: 0 originalissueId:

068373 Site: Point Beach Cartridge and Frame: Response: (None)Primary Topic: (None)Secondary Topic: (None)original~projectjid:

32 Primary Attribute: (None)Secondary Attribute: (None)INPO 03-04 Performance Objective: (None)Notes/Comments Additional Conservatisms In S&L Methodology by FISCHER, JEREMY (10/29/2005 12:30:34 PM)Engineering has continued to review the methodology and approach used by the S&L analysis.

The following conservatisms are inherent to the S&L analysis and noted below: http://enwsO2/tmtrack/tmtrack.dll?IssuePage&Template=printitem&recordid=897686&tab...

2/15/2006 Nuclear Management Company Page 4 of 6 l.Water depth: Minimum depth at switchover is actually 42 vs. the calculated 32" (an increase of 31%). The 42 depth does not credit any additional sump level that may occur due to the spilling of RCS and Si accumulator volumes to the containment.

When put into the Unit 2 calculation, all other things being equal, the head loss drops from 1.34' to 0.51', a 62% decrease.

At the same time, the depth of screen available increases from 1.6' (0.8' average) to 2.43' (1.22' average), a 52% increase.

Note that the non-linear beneficial effect of increasing the water depth on head loss.2.The assumption that all unqualified coatings, regardless of type and location, fail to 1/8" diameter disks is highly conservative.

This optimizes transport while maximizing screen blockage potential.

Recent industry testing (documented in NEI-0407) demonstrates that both Alkyd and Epoxy coatings fail to particles in the range of 10-100 μm. Particles this small would not lodge in the 1/8" mesh screen, and would not result in any head loss above that of a clean screen.3.An implicit assumption that the debris pile is formed from a constant ZOI at the minimum water depth. When RHR flow initiates, water depth is at a minimum level with a maximum ZOI due to the higher approach velocities.

In reality however, there would be no debris pile accumulation and no debris bed on the active screen at the point of switchover to sump recirculation.

These form over time as the water depth continues to increase, the approach velocity decreases, the ZOI decreases, and the total accumulation of debris would be significantly less than that which is calculated based on a constant low water level.4.An implicit assumption that all coatings that fail do so at the instant of initiating sump recirculation.

The transport analysis assumes that all coatings available for transport are transported if the velocities at the lowest flood level are high enough.In reality, coatings would be expected to fail over a period of time. Those that fail before the initiation of sump recirculation would sink to the floor of containment without being transported to the sump screen surface. If within the zone of influence for transport to the debris pile, they would do so when recirculation is initiated.

However, none would be transported up onto the active screen surface. Additionally, coatings failing significantly after recirculation is begun would fall into a deeper sump with correspondingly lower velocity vectors and therefore a smaller ZOI. The quantity that would be transported would be less than that calculated in the analysis.5.Walk-down data is conservatively rounded up when estimating areas of degraded or unqualified coatings.

Further, there is an additional analytical factor of 15% used when assessing the total quantity of unqualified or degraded coatings within the calculated ZOls.6.0nce a debris pile starts to form, it's presence alone reduces the size of the ZOI for accumulation on the active screen surface. Particles falling at the outer fringes of the calculated (minimum sump depth) ZOI will be deposited on the debris pile rather than on the surface of the screen. The actual ZOI would be reduced as the debris pile grows, and the quantity of material deposited on the active screen would be reduced below that calculated.

7.The current design basis precludes the use of containment spray while on sump recirculation.

The maximum available RHR flow under these conditions is -2100 gpm as opposed to the analysis flow rate of 2435 gpm (a reduction of 14%).When entered into the Unit 2 calculation, all other things being equal, the head loss drops from 1.34' to 1.02' (a reduction of 24% in the head loss; another greater-than-linear beneficial effect).8.A factor of 2 was applied to the head loss calculation results. This was based on engineering judgment at the time, and considered that the published correlation had not been demonstrated to be directly applicable to debris mixes of the type being evaluated.

This factor has been retained.Update- by KENDALL, THOMAS (10/29/2005 10:12:46 PM)ALL PREVIOUS ENTRIES ORIGINATOR ENTRIES ON THIS CAP WERE ENTERED BY T. C. KENDALL (ENGINEERING).

DUE TO A LOG-IN ERROR, M. OMILLIAN IS INCORRECTLY IDENTIFIED AS THE ORIGINATOR.

Additional information for assurance of continuing operability:

1) The sentence fragment at the end of the entry dated the evening of 10/29105 is an artifact that was unintentionally left in the text.2) The consultants involved in determining appropriate assessment of coating particle size and screen performance in the presence of suspended particles were G. Zigler of Alion Science, and J. Cavallo of Corrosion Control Laboratories.

Both have extensive experience in the subjects at hand, and particularly in testing and analyses in support of both BWR and PWR sump blockage concerns.

These individuals have been serving as consultants to both NEI, and PBNP's primary contractors for GSI-191 issues (a consortium of AREVA, PCI, and Alden Laboratories).

3) Owing to the fine nature of the coating decomposition products, their fragmentary

/ friable nature, and their relative softness compared with the wear surfaces of rotating compontents in the RHR and SI pumps, the suspended particles are not considered a challenge to the operability of these components.

Previous industry experience with high wear is attributable primarily to larger and hard abrasive particles (dirt, sand, grinding grit, filings, etc., particularly in the entrapping presence of tramp fibers) that were introduced into the system. When present, particularly with softer (i.e. brass or bronze)wear surfaces, these materials have been observed to significantly score the soft surfaces.

All wear surfaces in the Point http://enwsO2/tmtrack/tmtrack.dll?IssuePage&Template=printitem&recordid=897686&tab

... 2/15/2006 Nuclear Management Company Page 5 of 6 Beach Si and RHR pumps are stainless steel or chrome-nickle steel. Additionally, Point Beach does not have the cyclone separators that have been shown at other facilities to be susceptible to entrained debris.Additional bases for prompt operability consideration by KENDALL, THOMAS (10/29/2005 11:23:15 PM)Based on the Engineering Evaluation that is currently undergoing review, the worst-case volumetric concentration of suspended coating decomposition particles would be -0.13% (based on 27.5 cubic feet of dust-like particles suspended in almost 157,000 gallons of sump water at the start of sump recirculation).

NUREG/CR-2792 shows a 1% loss of hydraulic efficiency with a 1 % solid mass in a slurry. Therefore, it is reasonable to assert that a 0.13% slurry" would have insigificant hydraulic effects on the ECCS pumps.Similarly, the low volumetric concentration indicates a negligible impact on the viscosity, density, thermal capacity, and thermal conductivity of the water. Therefore, thermal and hydraulic performance of the cooling systems would not be adversely affected by the suspended particles.

Finally, the mechanism for coating decomposition is the apparent destruction of the binder adhering the solids to each other and the substrate.

The remaining solids would not expected to be "sticky" and adhere to the smoth heat transfer surfaces in the RHR heat exchangers with turbulent flow. While deposition may occur, it would preferrentially occur in regions of low velocity and/or rough surfaces such as the containment floor. Even if some fouling of the RHR heat exchanger surfaces did occur, these heat exchangers are limited by the decay heat removal requirements of a 100 deg F operational cool down in the normal RHR mode. They can tolerate a significant amount of fouling while remaining capable of removing decay heat after an accident.Attachments and Parent/Child Links Information relevant to prompioperabijy (40960 bytes) by OMILLIAN, MICHAEL (10/27/2005 11:03:16 PM)PEnncipalto OPR001 61: Containment Cpatings not maintained with in analyzed Omits by BELTZ, BOYD (10/27/2005 11:27:17 PM)PErncipa~otRCE0OQ294:

Containment Coatings not maintained within analyzed limits by KREIL, JULIE (11/1/2005 2:51:21 PM)Principal-toOBD00281:-(TRPR Containment Coatings not maintained withinanalyzedlimits by KREIL, JULIE (11/2/2005 7:37:57 AM)___icipai to MREm0535_(TRP Containment Coatings notmaintained within analyzed limits by KREIL, JULIE (11/18/2005 7:33:56 AM)LinkedfromCAP068346:

Potential Weaknessesin Containment Coating Programand Associated Analyses by TAYLOR, DAVID (11/22/2005 10:28:03AM)

Linked from CAP068535: (TRP) Unit 2 enters LCO 3.0.3 and commenced TS required shutdown.

by GADZALA, JACK (11/29/2005 8:58:55 AM)Linked To OBD000302 by admin (12/2/2005 6:06:51 PM)Linked to CAP068527:

Degraded Coating Inventory for Unit 2 May Exceed OPR 161 Analyis Limit by TAYLOR, DAVID (12/9/2005 12:51:27 PM)Change History 12127120051:17:05 PM by BENNETT, KEVIN Human Perf Fail Mode Changed From (None) To N/A -Not Applicable Process Fail Mode Changed From (None) To N/A -Not Applicable Org/Mgt Failure Mode Changed From (None) To N/A -Not Applicable Last Modified Date Changed From 12/9/2005 12:51:27 PM To 12/27/20051:17:05 PM Last Modifier Changed From TAYLOR, DAVID To BENNETT, KEVIN 1/1712006 7:58:05 AM by REBITZ, EMMY Last Modified Date Changed From 12/27/20051:17:05 PM To 1/17/2006 7:58:05 AM Last Modifier Changed From BENNETT, KEVIN To REBITZ, EMMY Attachment Updated: Principal to MRE000535: (TRP) Containment Coatings not maintained within analyzed limits 2114/2006 10:00:12 AM by REBITZ, EMMY Last Modified Date Changed From 1/17/2006 7:58:05 AM To 2/14/2006 10:00:12 AM http://enwsO2/tmtrackltmtrack.dll?IssuePage&Template=printitem&recordid=897686&tab

... 2/15/2006 Nuclear Management Company Page 6 of 6 Attachment Updated: Linked from CAP068535: (TRP) Unit 2 enters LCO 3.0.3 and commenced TS required shutdown.http://enwsO2/tmtrack/tmtrack.dll?IssuePage&Template=printitem&recordid=897686&tab

... 2/15/2006 Nuclear Management Company Page 1 of3 State Change History RCE Initiate by KREIL, JULIE Update by BENNETT, KEVIN Assign Work 11/1/2005 2:51:18 PM Owner SHERWOOD, GARY Conduct Work 2/10/2006 2:32:09 PM Owner SHERWOOD, GARY Assign by SHERWOOD, GARY Update by HORVATH, JOSEPH Conduct Work 11/4/2005 6:37:20 PM Owner SHERWOOD, GARY Assign Work 2/14/2006 11:14:42 AM Owner SHERWOOD, GARY Return by SHERWOOD, GARY Assign by SHERWOOD, GARY Assign Work 1/27/2006 11:18:53 AM Owner SHERWOOD, GARY Assign by SHERWOOD, GARY Conduct Work 1/27/2006 11:19:19 AM Owner HOPKINS, ROD Conduct Work 2/15/2006 5:24:27 PM Owner HORVATH, JOSEPH Section 1 Activity Request Id: Activity Type: Site/Unit:

One Line

Description:

Activity Requested:

RCE000294 Root Cause Evaluation Submit Date: 11/11/2005 2:51:18 PM Point Beach -Common Containment Coatings not maintained within analyzed limits Perform a Root Cause Evaluation of CAP068373 in accordance with NMC RCE Manual. 30 DAY DUE DATE. Per Managers Meeting of 11-01-05, Gary Sherwood is the RCE Management Sponsor. This RCE requires CARB review once completed.

Contact S. J. Nikolai to schedule this review.Also see attached RCE Prep Checklist and Extent of Condition for use in completing the RCE.Ensure the findings of PBSA-ENG-05-06 as identified in CAP068399, CAP068400, CAP068402, and CAP068403 are addressed as part of the closeout of this item.N Mode Change Restraint: (None)KENDALL, THOMAS Initiator Department:

EP Engineering Programs PB CATPR: Initiator:

Responsible Group Code: EP Engineering Programs Responsible Department:

PB Activity Supervisor:

SHERWOOD, GARY Activity Performer:

Section 2 Engineering HORVATH, JOSEPH Priority:

2 Due Date: 2/17/2006 Management Exception From PI?: N QAINuclear Oversight?:

N Licensing Review?: N NRC Commitment?:

N NRC Commitment Date: Significance Level: A Section 3 Activity Completed:

11/4/2005 6:37:20 PM -SHERWOOD, GARY: Based on available resources and the current outage in progress, this RCE will not be done until January 2006.11/22/2005 1:50:30 PM -KREIL, JULIE: RCE000293 was closed to this RCE (RCE000294).(None)Hot Buttons: Section 4 QA Supervisor: (N Root Causes & Contributing Causes: 0 Compensatory or Interim Corrective Actions: 0 Safety Sign.,EOC,Maint.

Rule Considerations:

0 Effectiveness Review Plan: 0 Section 5 lone) Licensing Supervisor: (None)Proposed Corrective Actions,CA Focus: 0 Previous Similar Events: 0 Operating Experience:

0 Project: Root Cause Evaluation (RCE) State: Conduct Work http://enws02/tmtrackltmtrack.dI1?lssuePage&Recordld=898436&Tableld=1 000&Templat

... 2/15/2006 Nuclear Management Company Page 2 of 3 Active/inactive:

Active Submitter:

KREIL, JULIE Last Modified Date: 2/15/2006 5:24:27 PM Last State Change Date: 2/15/2006 5:24:27 PM Close Date: NUTRK ID: Child Number: 0

References:

CAP068442 RCE000293 Owner: Assigned Date: Last Modifier: Last State Changer: HORVATH, JOSEPH 2/15/2006 SHERWOOD, GARY SHERWOOD, GARY Update: Import Memo Field: CAP Admin: OLD ACTIONNUM:

Cartridge and Frame: Response: Primary Topic: Secondary Topic: (None)Site: Point Beach (None)(None)(None)Primary Attribute: (None)Secondary Attribute: (None)INPO 03-04 Performance Objective: (None)Notes/Comments Note created during 'Return'transition by SHERWOOD, GARY (1/27/2006 11:18:53 AM)Extension request approved by plant manager for a new due date of 2/17/2006.

Extension request attached.Attachments and Parent/Child Links Subtask from CAP068373:

Containment Coatings not maintained within analyzed limits by KREIL, JULIE (1111/2005 2:51:21 PM)RCEprepchecklist.doc (52736 bytes) by KREIL, JULIE (11/1/2005 2:53:15 PM)RCE 294 Charter (267235 bytes) by PAWLITZKY, TINA (11/18/2005 8:57:22 AM)Linked to RCE000293:

Potential Weaknesses in Containment Coating Program and Associated Analyses by SHERWOOD, GARY (11/19/2005 9:04:00 PM)Linked fromOBD000283: (TRP) .Question with-the ability of ECCS sump screens to-pass required flow by PETERSON, LARRY (12/2/2005 6:04:09 PM)Linked from OBD000281: (TRP) Containment Coatings not maintained within analyzed limits by PETERSON, LARRY (12/2/2005 6:08:59 PM)Linked from CE0 6574: Independent Industy ConsuLtant Reporton PBNP ContainmentCoatings by TAYLOR, DAVID (1 2/22/2005 3:14:23 PM)Linked fromCA064725:

OE023369, Part 2lNotification-Review Coating Program Procedures by TAYLOR, DAVID (1/16/2006 2:53:53 PM)Extension Request Approval (175180 bytes) by HOPKINS, ROD (1/27/2006 9:20:48 AM)Linked from CAP068535: (TRP) Unit 2 enters LCO 3.0.3 and commenced TS required shutdown.LOYDE (2/7/2006 8:15:46 AM)by HAWKI, Revised RCE294 Charter.odf (298490 bytes) by HORVATH, JOSEPH (2/15/2006 4:29:21 PM)Change History 2/15/2006 4:29:22 PM by HORVATH, JOSEPH Last Modified Date Changed From 211412006 11:14:42 AM To 2115/2006 4:29:22 PM Attachment Added: Revised RCE294 Charter.pdf 2/1512006 5:24:27 PM by SHERWOOD, GARY Activity Performer Changed From (None) To HORVATH, JOSEPH State Changed From Assign Work To Conduct Work Via Transition:

Assign http://enwsO2/tmtrack/tmtrack.dll?IssuePage&Recordld=898436&Tableld=1 000&Templat...

2/15/2006 Nuclear Management Company Page 3 of 3 Owner Changed From SHERWOOD, GARY To HORVATH, JOSEPH Assigned Date Changed From 1/27/2006 To 2/15/2006 Last Modified Date Changed From 2115/2006 4:29:22 PM To 211512006 5:24:27 PM Last Modifier Changed From HORVATH, JOSEPH To SHERWOOD, GARY Last State Change Date Changed From 2/14/2006 11:14:42 AM To 2/15/2006 5:24:27 PM Last State Changer Changed From HORVATH, JOSEPH To SHERWOOD, GARY http://enws02/tmtrack/tmtrack.dIl?lssuePage&Recordld=898436&TableId=I O00&Templat

... 2/15/2006 Root Causc Evaluation Charter CAP068373 and CAP068346 RCE 294 Sponsor: Gary Shenvood Problem Statement:

During performance of containment coating inspections and evaluations in UI R29 the following deficiencies were identified.

• Thc as found inventory of unqualified and degraded coatings was above the limits assumed in the analysis of record.* Coating deficiencies were not being documented consistcntly nor wvere corrective actions to correct deficiencies being taken in a timely manner.* Formal engineering evaluations were not performed to ensure the inventory of degraded and unqualified coatings remained within the limits of the analysis of record.* The head loss correlation used for the analysis of record was not valid for the assumed RHR pump suction conditions.

• The impact of sump blockage fromn debris was not correctly modeled in the fluid dynamics model used to calculate available NPSH.Investigation Scope: The purpose of this investigation is to identify the vulnerabilities in the Containment Coating Program and the implementation of the program, which resulted in operation of the plant outside of its design and licensing basis. The investigation should review the adequacy of the current containment coating program, supporting analysis and implcmenting procedures.

The investigation should assess %whether the program and its associated implementing procedures are adequate, if properly implemented, to maintain the plant within its design and licensing bases. Also, the review should look at organizational and programmatic failure modes.Make recommendations for;* Correcting the problem* Preventing recurrence of the problem* Applicability of the root cause to other areas (extent of condition)

Root Cause Evaluation Charter CAP068373 and CAP068346 RCE 294 Investigation Mcthodolorv:

0 0 0 a Interviews Causal Factor Charting Barrier Analysis Cause and Effect Analysis Stream Analysis (if appropriate)

Team Members: Position .Name Department Team Leader Rod Hopkins J Engineering Programs PBNP RCE Evaluator Joe Horvath J Engineering Design PBNP Team Member Davc Taylor Enginecring Programs PBNP Team Member TBD .Engineering Design Fleet Milestones:

Date Assigned 10/26/2005 Status Update 12/212005 Draft Report 12/09/2005 Final Report 12/16/2005 Communications Plan Communication Responsible Date Present RCE charter to Screening Tcam G Shervood 1111612005 Brief PARB on status of RCE G Sherwood 12/6/2005 Provide written status report on RCE to G Shernvood 11/21/2005 Station Sr Management 11/28/2005 12/5/2005 12/12/2005 12/19/2005 Present final RCE to CARB G Sherwood 1/1512006 Approved Reviewed by: ow 4 zSDatc // I./ooS'Gary SheiWood 5 U Management SpoD16 r 7n g a~ Date "//A/eiY Circle one Nuclear Management Company PagelI of 2 State Change History OBD Initiate by KREIL, JULIE Complete and Close by KREIL, JULIE Assign Work 1112/2005 7:37:56 AM Owner PETERSON, LARRY Assign by PETERSON, LARRY Conduct Work 11/9/2005 5:45:47 PM Owner PETERSON, LARRY Work Complete by PETERSON, LARRY Review &Approval 12/2/2005 6:09:52 PM Owner PETERSON, LARRY Approved by PETERSON, LARRY Quality Check 12/212005 6:10:58 PM Owner PBNP CAP Admin Done 1/20/2006 8:04:58 AM Owner (None)Section 1 Activity Request Id: Activity Type: Site/Unit:

One Line

Description:

Activity Requested:

CATPR: Initiator:

OBD000281 Operable But Degraded Submit Date: 11/2/2005 7:37:56 AM Point Beach -Common (TRP) Containment Coatings not maintained within analyzed limits Review OPROO0161 and document an Action Plan (CAPIan) to address the issue. Create necessary actions to accomplish the Action Plan, per ESG 1.10. (Recommend a 30 day due date from date of the approved/signed OPR.)N Mode Change Restraint: (None)KENDALL, THOMAS Initiator Department:

EP Engineering Programs PB Responsible Group Code: ED Engineering Design Responsible Department:

PB Engineering Activity Supervisor:

Section 2 PETERSON, LARRY Activity Performer:

PETERSON, LARRY Priority:

3 Due Date: 12/2/2005 Management Exception From Pi?: N QA/Nuclear Oversight?:

N Licensing Review?: N NRC Commitment?:

N NRC Commitment Date: Significance Level: A Section 3 Activity Completed:

12/2/2005 6:09:52 PM -PETERSON, LARRY: Cap plan completed and attached as PBF-1553a form.Actions to be followed via RCE294, OBD301, OBD302 12/2/2005 6:10:58 PM -PETERSON, LARRY: Action completed as required and documented in attached PBF1553a.

Follow-on actions linked to this record via RCE294, OBD301, and OBD302.This action completed and can be closed.12/5/2005 1:12:34 PM -KREIL, JULIE: Pending CAP Technical Panel review.1/20/2006 8:04:58 AM -KREIL, JULIE: CLOSED per CAP Technical Review Panel (Tom Kendall, Clay Hill, Terry Guay) meeting of 1-5-06.(None)Hot Buttons: Section 4 QA Supervisor: (None) Licensing Supervisor: (None)Section 5 Project: Active/inactive:

Operable But Degraded (OBD) State: Inactive Owner: Done (None)http://enwsO2/tmtrack/tmtrack.dll?lssuePage&RccordId=89851 7&TableId=I OOO&Templat

... 2/15/2006 Nuclear Management Company Page 2 of 2 Submitter:

KREIL, JULIE Last Modified Date: 1/20/2006 8:04:58 AM Last State Change Date: 1/20/2006 8:04:58 AM Close Date: 1/20/2006 8:04:58 AM NUTRK ID: Child Number: 0 Assigned Date: Last Modifier: Last State Changer: 11/9/2005 KREIL, JULIE KREIL, JULIE

References:

Update: Import Memo Field: CAP Admin: OLD ACTION NUM: Cartridge and Frame: Response: Primary Topic: Secondary Topic: CAP068442 PBNP CAP Admin Site: Point Beach (None)(None)(None)Primary Attribute: (None)Secondary Attribute: (None)INPO 03-04 Performance Objective: (None)Attachments and Parent/Child Links Subtask fromCAP068373:

Containment Coatings-not maintainedwithinanalyzedlimits by KREIL, JULIE (11/2/2005 7:37:57 AM)OBD281 pbf1553a.doc (48128 bytes) by PETERSON, LARRY (1212/2005 6:06:11 PM)Linked to OBD000301:

Question with the ability of ECCSsump screens to pass required flow by PETERSON, LARRY (12/2/2005 6:08:26 PM)Linked to OBD000302:

Containment Coatings not maintained within analyzed limits by PETERSON, LARRY (12/2/2005 6:08:43 PM)Linked to RCE000294:

Containment Coatinas not maintained within analyzed limits by PETERSON, LARRY (12/2/2005 6:08:59 PM)Change History 12/5/2005 1:12:34 PM by KREIL, JULIE Last Modified Date Changed From 12/2/2005 6:10:58 PM To 12/5/20051:12:34 PM Last Modifier Changed From PETERSON, LARRY To KREIL, JULIE 1120/2006 8:04:58 AM by KREIL, JULIE Active/Inactive Changed From Active To Inactive Owner Changed From PBNP CAP Admin To (None)Close Date Changed From Unassigned To 1/20/2006 8:04:58 AM Last Modified Date Changed From 12/5/2005 1:12:34 PM To 1/20/2006 8:04:58 AM Last State Change Date Changed From 12/2/2005 6:10:58 PM To 1/20/2006 8:04:58 AM Last State Changer Changed From PETERSON, LARRY To KREIL, JULIE Activity Completed Changed From '[Original Text]' To '[Appended:]

CLOSED per CAP Technical Review Panel (Tom Kendall, Clay Hill, Terry Guay) meeting of 1-5-06.'State Changed From Quality Check To Done Via Transition:

Complete and Close http://enwsO2/tmtrack/tmtrack.dll?IssuePage&Recordld=89851 7&Tableld=1 OOO&Templat

... 2/15/2006 Nuclear Management Company Page I of 2 State Change History MRE Initiate by KREIL, JULIE Complete and Close by PFAFF, SCOTT Assign Work 11/18/2005 7:33:55 AM Owner HAWKI, LOYDE Assign by HAWKI, LOYDE Conduct Work 11121/2005 5:10:29 PM Owner HOFSTRA, JONATHON Work Complete by HOFSTRA, JONATHON Review &Approval 12/14/2005 11:30:37 AM Owner HAWKI, LOYDE Approved by HAWKI, LOYDE Quality Check 12/15/2005 12:54:01 PM Owner PBNP CAP Admin Done 2/10/2006 9:09:21 AM Owner (None)Section 1 Activity Request Id: Activity Type: Site/Unit:

One Line

Description:

Activity Requested:

MRE000535 Maintenance Rule Evaluation Submit Date: 11/18/2005 7:33:55 AM Point Beach -Common (TRP) Containment Coatings not maintained within analyzed limits Perform a Maintenance Rule Evaluation in accordance with NP 7.7.6. Appendix B provides some overall guidance on what should be considered in the MRE. This MRE item has a 30 day due date.CATPR: Initiator:

N KENDALL, THOMAS Responsible Group Code: EPI Engineering Programs Inspection Services PB Activity Supervisor:

HAWKI, LOYDE Mode Change Restraint:

Initiator Department:

Responsible Department:

Activity Performer: (None)EP Engineering Programs PB Engineering HOFSTRA, JONATHON Section 2 Priority:

3 Due Date: 12/18/2005 Management Exception From PI?: N QA/Nuclear Oversight?:

N Licensing Review?: N NRC Commitment?:

N NRC Commitment Date: Significance Level: A Section 3 Activity Completed:

11/21/2005 5:10:29 PM -HAWKI, LOYDE: Need to evaluate assignment.

Appears this is an ECCS MRE not a containment system MRE.Currently assigned as others are not available due to current extended outage and holiday schedule.12/14/2005 11:30:37AM

-HOFSTRA, JONATHON: In reviewing this MRE it appears that this MRE issue is covered under MRE 000534 (Unit 2 enters LCO 3.0.3 and commenced TS required shutdown) and this MRE can be closed to MRE 000534.I also talked to the Maintenance Rule Coordinator and had this person review these MRE's, to see if I could close this MRE to MRE 000534. This person also came to the same conclusion that I did, both MRE's cover the same issue.Therefore, MRE 000535 (containment coating not maintained within analyzed limit) can be closed to MRE 000534 (Unit 2 enters LCO 3.0.3 and commenced TS required shutdown), because this issue will be covered in the evaluation of MRE 000534.12/15/2005 12:54:01 PM -HAWKI, LOYDE: Thi sitem redundant to MRE000534.

1/17/2006 7:58:01 AM -REBITZ, EMMY: Pending CAP Technical Review 2/10/2006 9:09:21 AM -PFAFF, SCOTT: TRP accepted 1126106.(None)Hot Buttons: Section 4 http://enwsO2/tmtrack/tmtrack.dll?lssuePage&Recordld=90 1 066&Tableld=1 OOO&Templat

... 2/15/2006 Nuclear Management Company Page 2 of 2 QA Supervisor: (None) Licensing Supervisor: (None)Section 5 Project: Maintenance Rule Evaluation(MRE)

Active/Inactive:

Inactive Submitter:

KREIL, JULIE Last Modified Date: 2/10/2006 9:09:21 AM Last State Change Date:2/10/2006 9:09:21 AM Close Date: 2/10/2006 9:09:21 AM NUTRK ID: Child Number: 0

References:

CAP068442 Update: Import Memo Field: CAP Admin: PBNP CAP Admin OLD ACTION NUM: Cartridge and Frame: Response: (None)Primary Topic: (None)Secondary Topic: (None)State: Owner: Assigned Date: Last Modifier: Last State Changer: Done (None)11/21/2005 PFAFF, SCOTT PFAFF, SCOTT Site: Point Beach Primary Attribute: (None)Secondary Attribute: (None)INPO 03-04 Performance Objective: (None)Attachments and Parent/Child Links Subtask fromCAP068373:_ContainmentCoatin not maintained within aaIedlimit byKREIL,JULIE(11/18/2005 7:33:56 AM)Change History 1117/2006 7:58:01 AM by REBITZ, EMMY Last Modifier Changed From HAWKI, LOYDE To REBITZ, EMMY 211012006 9:09:21 AM by PFAFF, SCOTT Activity Completed Changed From '[Original Text]' To '[Appended:]

TRP accepted 1/26/06.'State Changed From Quality Check To Done Via Transition:

Complete and Close Active/Inactive Changed From Active To Inactive Owner Changed From PBNP CAP Admin To (None)Last Modified Date Changed From 1/17/2006 7:58:01 AM To 2/10/2006 9:09:21 AM Last Modifier Changed From REBITZ, EMMY To PFAFF, SCOTT Last State Change Date Changed From 12/15/2005 12:54:01 PM To 2/10/2006 9:09:21 AM Last State Changer Changed From HAWKI, LOYDE To PFAFF, SCOTT Close Date Changed From Unassigned To 2/10/2006 9:09:21 AM http://enwsO2/tmtrack/tmtrack.dll?IssuePage&Recordld=90 1 066&TableId=1 OOO&Templat...

2/15/2006 Nuclear Management Company Pagel of2 State Change History OBD Initiate Assign Work Assign Conduct Work 121212005 6:06:46 PM 12/912005 5:47:05 PM by PETERSON, LARRY Owner PETERSON, LARRY by PETERSON, LARRY Owner MCNAMARA, JOE Section 1 Activity Request Id: Activity Type: Site/Unit:

One Line

Description:

Activity Requested:

CATPR: Initiator:

OBD000302 Operable But Degraded Submit Date: 12/2/2005 6:06:46 PM Point Beach -Common Containment Coatings not maintained within analyzed limits Revise/update GL 98-04 response as necessary.

N Mode Change Restraint: (None)KENDALL, THOMAS Initiator Department:

EP Engineering Programs PB Responsible Group Code: EDM Engineering Design Mechanical Responsible Department:

Engineering PB Activity Supervisor:

MCNAMARA, JOE Activity Performer:

MCNAMARA, JOE Section 2 Priority:

3 Due Date: 8/31/2007 Management Exception From PI?: N QA/Nuclear Oversight?:

N Licensing Review?: N NRC Commitment?:

N NRC Commitment Date: Significance Level: A Section 3 Activity Completed:

12/9/2005 5:47:05 PM -PETERSON, LARRY: Due date established per CAP Plan generated in OBD281. Due date assumes need for completion of GSI 191 resolution for final closure.Hot Buttons: (None)Section 4 QA Supervisor: (None) Licensing Supervisor: (None)Section 5 Project: Operable But Degraded (OBD) State: Active/lnactive:

Active Owner Submitter:

PETERSON, LARRY Assigr Last Modified Date: 12/9/2005 5:47:05 PM Last M Last State Change Date: 12/9/2005 5:47:05 PM Last S Close Date: NUTRK ID: Child Number: 0

References:

CAP068442 Update: Import Memo Field: CAP Admin: (None) Site: OLDACTIONNUM:

Cartridge and Frame: Response: (None) Primar Primary Topic: (None) Seconm Secondary Topic: (None) INPO 0 Attachments and Parent/Child Links Linked From CAP068373 by admin (12/2/2005 6:06:52 PM)ied Date: Codifier: tate Changer: Conduct Work MCNAMARA, JOE 12/9/2005 PETERSON, LARRY PETERSON, LARRY Point Beach y Attribute: (None)dary Attribute: (None)13-04 Performance Objective: (None)Linked to OBD000301:

Question with the ability of.ECCS sumpscreens to pass-required flow by PETERSON, LARRY (12/212005 6:07:25 PM)Linked from OBD000281: (TRP) Containment Coatings not maintainedwithinanalyzed limits by PETERSON, LARRY (12/2/2005 6:08:43 PM)http://enwsO2/tmtrack/tmtrack.dll?IssuePage&RecordId=902046&TableId=I OOO&Templat...

2/15/2006 Nuclear Management Company Page 2 of 2 Change History 1219/2005 5:47:05 PM by PETERSON, LARRY Responsible Department Changed From (None) To Engineering Activity Supervisor Changed From PETERSON, LARRY To MCNAMARA, JOE Activity Performer Changed From (None) To MCNAMARA, JOE Priority Changed From (None) To 3 Due Date Changed From Unassigned To 8/3112007 Activity Completed Changed From " To'[Appended:]

Due date established per CAP Plan generated in OBD281. Due date assumes need for completion of GSI 191 resolution for final closure.'State Changed From Assign Work To Conduct Work Via Transition:

Assign Owner Changed From PETERSON, LARRY To MCNAMARA, JOE Assigned Date Changed From 11/9/2005 To 1219/2005 Last Modified Date Changed From 12/5/2005 1:12:35 PM To 12/9/2005 5:47:05 PM http://enwsO2/tmtrack/tmtrack.dll?IssuePage&Recordld=902046&Tableld=1 OOO&Templat...

2/15/2006

--Nuclear Management Company Pagel of3 State Change History Initiate by BACH, STEVEN AR Pre-Screen 10/30/2005 5:39:40 AM Owner (None)Submit to Screening Team by BELTZ, BOYD AR Screening Que 10/31/2005 9:58:48 PM Owner PBNP CAP Admin Screening Update by KREIL, JULIE Done 11/212005 6:54:03 AM Owner (None)Update by ZIFKO, TRACEY AR Screening Que 2/8/2006 7:17:58 AM Owner BENNETT, KEVIN Section 1 Activity Request Id: Activity Type: One Line

Description:

Detailed

Description:

CAP068442 CAP Submit Date: 10/30/2005 5:39:40 AM GL 98-04 Commitments 10/30/2005 5:39:40 AM -BACH, STEVEN: In the Point Beach response to Generic Letter 98-04, transmitted to NRC in NPL 98-0950, we stated on page 4 of 6 that we had implemented the recommendations from the Gibbs and Hill studies to eliminate near field effects. The NRC has requested we provide details on the completion of these recommendations..

The recommendations in the Gibbs and Hill studies appears to have been addressed through work order 9709208 for unit I and 9706824 for unit 2. However, there are differences between what was recommended by Gibbs and Hill and the work actually performed in the field.These two work orders, which removed adhesive labels and failing coatings in the near field area for the sump B strainers of each unit, were driven by duel-unit re-start issue CR 97-2178. The recommendations of CR 97-2178 are based on a general recommendation from the evaluation of IN 95-006 (Ttrack id LAR046129) that adhesive label use be limited to products that have been specifically evaluated for post-LOCA conditions.

Note that the IN 95-006 evaluation was based upon RG 1.82 guidance and the Gibbs and Hill studies Initiator:

BACH, STEVEN Initiator Department:

EESN Engineering Equipment Systems NSSS Mech PB Date/Time of Discovery:

10/30/2005 5:19:38 Date/Time of Occurrence:

10/30/2005 5:19:38 AM AM Identified By: NRC System: SI PB Equipment

1. (1st): (None) Equipment Name (1st): (None)Equipment

1. (2nd): (None) Equipment Name (2nd): (None)Equipment

1. (3rd): (None) Equipment Name (3rd): (None)Site/Unit:

Point Beach -Common Why did this occur?: 10/3012005 5:39:40 AM -BACH, STEVEN: The actual recommendations from the Gibbs and Hill studies vary slightly between units due to different sump B strainer locations.

For unit 1 it was recommended that all concrete coatings be removed from the access shaft wall (for sump A) adjacent to the strainers, and that either a canopy be installed over the strainer or that all steel coatings within 8 feet of the strainer be removed. Note that in the study these recommendations were based on 'highly conservative assumptions' and were intended to 'preclude any additional sump screen blockage" not accounted for by the far field portion of the analysis.

The unit 2 study was performed after the unit 1 study. The Unit 2 recommendations

'for the elimination of any possibility of sump blockage due to paint failure in the near sump zone' were to remove all concrete coatings within the near field zone and either install a canopy or replace the steel coatings within the near field zone.Based on this initial research, it is believed that the Gibbs and Hill recommendations were felt to be overly conservative, and that the subsequent evaluations justified a less drastic approach.The results of Sargent and Lundy calculation M-09334-345-RH.1 revision 0 also supported the less drastic approach.

Thus, the removal of adhesive labels and un-sound coatings was felt to be an appropriate preventative action.Immediate Action Taken: 10/30/2005 5:39:40 AM -BACH, STEVEN: Researched the basis for the GL response and actions taken.Walked down the unit I affected area.Summarized findings and discussed with supervision.

Recommendations:

10/30/2005 5:39:40 AM -BACH, STEVEN: An engineering evaluation and operability recommendation are already in progress to evaluate plant-impact of the broader containment coatings issue.http://enwsO2/tmtrack/tmtrack.dll?IssuePage&Template=printitem&recordid=89801 2&tab ... 2/15/2006

=' , 4 Nuclear Management Company Page 2 of 3 Walkdown of the unit 1 near field zone has verified that the Gibbs and Hill recommendations have not been strictly followed.

Thus, it is recommended that this issue be entered into the corrective action process, and that the commitment change process be used to revise the commitment to existing practice.

Such revision is believed to be further supported by more recent debris transport and downstream effects analysis performed in support of Generic Letter 2004-2 response.SRO Review Required?:

Y Section 2 Operability Status: Operable but Non-Conforming Compensatory Actions: N Basis for Operability:

10/30/2005 8:15:35 PM -BELTZ, BOYD: Sump B screen and ECCS system is operable.

This issue is being addressed via OPR-161 (CAP 068200 ). The data indicates the concern for sump blockage due to coatings will not affect operability of ECCS Train. This CAP will be dispositioned with the review and approval of OPR-161. As this CAP does not identify any new concerns than those that are presently being evaluated.

Sump B and ECCS Trains are Operable.10/30/2005 8:24:38 PM -BELTZ, BOYD: CAP referenced should be 068373.10/31/2005 9:58:48 PM -BELTZ, BOYD: Based on review and approval of OPR-1 61 which also addressed the label issue. PBNP Containment Sump B Screens fully capable of performing their safety function, however we are non-conforming to our license basis due to methdology and communication to the NRC in regards to GL 98-04. reference engineering eval 2005-0024.

Unplanned TSAC Entry: N External Notification:

N Section 3 Screened?:

Y Significance Level: B INPO OE Reqd?: N Potential MRFF?: N QA/Nuclear Oversight?:

N Licensing Review?: N Good Catch/Well Doc'd?: NA Section 4 Inappropriate Action: 12/28/2005 7:27:31 AM -BENNETT, KEVIN: This is a legacy issue.Process: N/A -Not Applicable Activity:

N/A -Not Applicable Human Error Type: N/A -Not Applicable Human Perf Fail Mode: NIA -Not Applicable Equip Failure Mode: (None) Process Fail Mode: N/A -Not Applicable Org/Mgt Failure Mode: N/A -Not Applicable Group Causing Prob: (None)Hot Buttons: (None)Section 5 CAP Admin: Project: Active/Inactive:

Owner: Last Modifier: Last State Changer: NUTRK ID:# of Children:

References:

Update: BENNETT, KEVIN CAP Owner: Corrective Actn Program (CAP) AR State: Active Submitter:

BENNETT, KEVIN Last Modified Date: ZIFKO, TRACEY Last State Change Date: ZIFKO, TRACEY Close Date: HOPKINS, ROD AR Screening Que BACH, STEVEN 2/8/2006 7:19:34 AM 2/8/2006 7:17:58 AM 0 CAP068373 CLOSED TO CAP068373 PER 11-01-05 MANAGERS MEETING.Prescreen Comments: Import Memo Field: OPR Completed?:

OLD ACTIONNUM:

subtsid: originalIssueid:

Site: Per the 02-03-06 CAP screening team, this CAP is to be reopened and a CE initiated.

N 0 068442 Point Beach original project id: 32 http://enwsO2/tmtrack/tmtrack.dll?IssuePage&Template=printitem&recordid=898012&tab

... 2/15/2006

., INuclear Management Company Page 3 of 3 Cartridge and Frame: Response: (None) Primary Attribute: (None)Primary Topic: (None) Secondary Attribute: (None)Secondary Topic: (None) INPO 03-04 Performance Objective: (None)Attachments and Parent/Child Links Linked to CAP070184:

Inadequate communication leads to in process error by ZIFKO, TRACEY (2/8/2006 7:17:45 AM)Principal to CE016874:

GL 98-04 Commitments by ZIFKO, TRACEY (2/8/20067:19:34 AM)Change History 2/812006 7:17:58 AM by ZIFKO, TRACEY Last Modified Date Changed From 218/2006 7:17:45 AM To 2/8/2006 7:17:58 AM Last State Change Date Changed From 11/2/2005 6:54:03 AM To 2/8/2006 7:17:58 AM Last State Changer Changed From KREIL, JULIE To ZIFKO, TRACEY Close Date Changed From 11/2/2005 6:54:03 AM To Unassigned Update Changed From 'CLOSED TO CAP068373 PER 11-01-05 MANAGERS MEETING.'To

'CLOSED TO CAP068373 PER 11-01-05 MANAGERS MEETING. Per the 02-03-06 CAP screening team, this CAP Is to be reopened and a CE initiated.'

2/8/2006 7:18:03 AM by ZIFKO, TRACEY Last Modified Date Changed From 2/8/2006 7:17:58 AM To 2/8/2006 7:18:03 AM original_project id Changed From 0 To 32 originalissueid Changed From " To '068442'2/812006 7:19:34 AM by ZIFKO, TRACEY Last Modified Date Changed From 2/8/2006 7:18:03 AM To 2/8/2006 7:19:34 AM Attachment Added: Principal to CE016874:

GL 98-04 Commitments http://enwsO2/tmtrackltmtrack.dll?IssuePage&Template=printitem&recordid=8980 1 2&tab ... 2/15/2006

.-Nuclear Management Company Page I of 2 State Change History CE Initiate Assign Work Assign Conduct Work CE Initiat TRCY 2/8/2006 7:19:33 AM bBENTKVN 2/15/2006 3:26:32 PM y , EY Owner HOPKINS, ROD by BENNETT. KEVIN Owrer SHERWOOD, GARY Section 1 Activity Request Id: Activity Type: SitelUnit:

One Line

Description:

Activity Requested:

CATPR: Initiator:

CE016874 Condition Evaluation Submit Date: 2/8/2006 7:19:33 AM Point Beach -Common GL 98-04 Commitments Per the 02-03-06 Managers Meeting, perform a Condition Evaluation of CAP068442 in accordance with NP 5.3.1.NOTE: On 03/09/06 tTrack will be removed from service. Per S. Pfaff email of 02/03/06, due dates will not be assigned between March 9 and March 26. Therefore, this CE's due date is March 27 N Mode Change Restraint: (None)BACH, STEVEN Initiator Department:

Responsible Group Code: EPE Engineering Program Equipment Reliability PB Activity Supervisor:

SHERWOOD, GARY Section 2 Responsible Department:

Activity Performer:

EESN Engineering Equipment Systems NSSS Mech PB Engineering SHERWOOD, GARY Priority: (None) Due Date: 3/27/2006 Management Exception From PI?: N QA/Nuclear Oversight?:

N Licensing Review?: N NRC Commitment?:

N NRC Commitment Date: Significance Level: B Section 3 Condtn Eval: 1. Condition Assessment

/ Issue Summary -describe the present condition.

2. Solution to Implement

-identify chosen solution.Activity Completed:

Hot Buttons: (None)Section 4 QA Supervisor: (None) Licensing Supervisor: (None)Section 5 Project: Condition Evaluation (CE) State: Active/Inactive:

Active Owner: Submitter:

ZIFKO, TRACEY Assigned Date: Last Modified Date: 2/15/2006 3:26:32 PM Last Modifier: Last State Change Date: 2/15/2006 3:26:32 PM Last State Changer: Close Date: NUTRK ID: Child Number: 0

References:

CAP068373 Update: CLOSED TO CAP068373 PER 11-01-05 MANAC Conduct Work SHERWOOD, GARY 2/15/2006 BENNETT, KEVIN BENNETT, KEVIN GERS MEETING.Per the 02-03-06 CAP screening team, this CAP is to be reopened and a CE initiated.

Import Memo Field: CAP Admin: OLDACTIONNUM:

Cartridge and Frame: Response: Primary Topic: Secondary Topic: (None)Site: Point Beach (None)(None)(None)Primary Attribute: (None)Secondary Attribute: (None)INPO 03-04 Performance Objective: (None)http://enwsO2/tmtrack/tmtrack.dll?IssuePage&Recordld=906788&TableId=1 000&Templat...

2/15/2006

,- Nuclear Management Company Page 2 of 2 Attachments and Parent/Child Links Subtask from CAP068442:

GL 98-04 Commitments by ZIFKO, TRACEY (2/8/2006 7:19:34 AM)Change History 2/8/2006 7:19:34 AM by ZIFKO, TRACEY Attachment Added: Subtask from CAP068442:

GL 98-04 Commitments 2/15/2006 3:26:32 PM by BENNETT, KEVIN Activity Supervisor Changed From HOPKINS, ROD To SHERWOOD, GARY Activity Performer Changed From (None) To SHERWOOD, GARY State Changed From Assign Work To Conduct Work Via Transition:

Assign Owner Changed From HOPKINS, ROD To SHERWOOD, GARY Assigned Date Changed From Unassigned To 2/15/2006 Last Modified Date Changed From 2/8/2006 7:19:34 AM To 2/15/2006 3:26:32 PM Last Modifier Changed From ZIFKO, TRACEY To BENNETT, KEVIN Last State Change Date Changed From 2/8/2006 7:19:33 AM To 2/15/2006 3:26:32 PM Last State Changer Changed From ZIFKO, TRACEY To BENNETT, KEVIN http://enwsO2/tmtrack/tmtrack.dll?IssuePage&Recordld=906788&TableId=1 OOO&Templat...

2/15/2006 Point Beach Nuclear Plant ENGINEERING EVALUATION COVER SHEET AND REVIEW FORM Engineerin2 Evaluation Number: Title of Engineering Evaluation:

2005-0024 Evaluation of Containment Sump Screen Debris Buildup Based on EPRI Technical Report and Current Degraded Epoxy Inventories System (CHAMPS Identifier Codes): UNIT: PBO [] PB1 ED PB2 0 Cont, RH, and Si El Original Engineering Evaluation 0 Revised Engineering Evaluation, Revision # I[ Supersedes Engineering E Canceled QA Scope 0D Yes D No Discipline:

0 CIVIL Safety Related 0 Yes El No L ELECTRICAL E INSTRUMENTATION AND CONTROL 19MECHANICAL L NUCLEAR COMPONENTS U PRA El CHEMICAL AND RADIOLOGICAL-El SYSTEMS Paee Inventory:

Page 1 -2 Form PBF-1608b Attachments:

Page j- 11 Form PBF-1649 NPC 2005-00627

Pages Al through Al NPC2005-00628 B! BI John Crane Seals Cl C6 Specific Gravity DI D2 Corrected Specific El E2 Gravity Summary of WO F! F1 0502011 results S&L ZO for Epoxy Coatings GI GI U2 Prepared By: (Qual Required)

Date: Tom Kendall _ _ _ _ _ _ _ _ _ ___Print Name Signature Reviewed By: (Qual Required)

Date: Mark Ralph M J7. .___________

Print Name Signature Approved By: Date: Print Naidented PBF-1 608b Revision 2 08/05/04 RECD NOV 0 9 2005 Page I of 9 Refrernce:

NP72.10 Engineering Eval. No. 2005-0024 Revision I Point Beach Nuclear Plant ENGINEERING EVALUATION Page No. 2 of 11 Initials TCK Date 1119/2005 Comments And Resolution Reviewer Comments: Resolution:

1.Incorporate discussions of additional margin (i.e.increasing sump depth with time)2. Demonstrate additional margin available by further reducing the outlet flow rates to that present at the time of containment sump switchover

(-1560 gpm)3. Remove remnants of discussions on floating tape 1. Incorporated

2. Incorporated

3. Incorporated PBF-1608b Revision 2 08105/04 Page 2 of 10

Reference:

NP7.2.10 Engineering Eval. No. 2005-0024 Point Beach Nuclear Plant Revision I ENGINEERING EVALUATION PageNo. 3 Of 1 Initials TCK Date 11/9/2005 1.0 PURPOSE This evaluation will assess debris buildup on the Unit I and Unit 2 Containment B Sump fine screens relative to experimental data concerning alkyd coatings and current inventories of damaged epoxy in the Zones of Influence in both containments.

The Zones of Influence are the areas extending from the sump screens in which debris could be potentially transported to the screens rather than settling on the floor. These zones are determined in References I and 2.The purpose of this evaluation is to estimate the impact of an error encountered in Calculation M-09334-345-RH.I, Rev. 1"Containment Sump Blockage Due To Failure Of Unqualified/Undocumented Coatings" (Unit 1) and Calculation M-09334-431-RH.I, Rev. 0 "Containment Sump Blockage Due To Failure Of Unqualified/Undocumented Coatings" (Unit 2)(References 1 and 2). The error encountered in both calculations concerns the value of the surface to volume ratio (S,) for Alkyd debris used in the calculation of the head loss across the Containment Sump Screens following a hypothetical Loss of Coolant Accident (LOCA). The impact on the sump screens of tape and adhesive labels present in containment is also evaluated to determine if this potential debris can block flow from reaching the sump screens.Revision I was prepared as the result of follow-up testing and an error in the tabulation of degraded coatings in Unit 2: After issuance of Revision 0, the anomalous specific gravity results for the red and white striped tape (i.e. specific gravity less than 1, contrary to expectations) prompted additional testing. The intent was to determine whether the tape would break down in a hot chemical solution and cause particles of size and density that could block screen passages.

Prior to performing the hot chemical testing, it was determined that the previous measurement of specific gravity contained an error, and that the tape actually has a specific gravity greater than one. While the exact cause of the error was not determined, it is likely that it was due to some adherent air bubbles or trapped air in the tape sample as its displacement was measured.

Revision 1 corrects the erroneous results pertaining to the buoyancy of the subject tape, and hot chemical testing of the tape was not pursued.Additionally.

after the issuance of Revision 0, it was discovered that there was approximately I I square feet of degraded (delaminating) epoxy coating on the 21' elevation of the unit 2 containment that is within the previously established Zone of Influence for impingement on the ECCS sump screens (see CAP 068527). Input #4 (quantity of degraded epoxy coatings in the vicinity of the sump screens) had not reflected this inventory.

Reference

1. 6 (Attachment B) was in error. The accessible portions of the degraded coatings were removed to the extent practical.

However, some of these degraded epoxy coatings remain in place. This revision also addresses whether the remaining coatings could challenge the operability of the Unit 2 ECCS screens.

2.0 BACKGROUND

In 1999, two calculations (Ref. I and 2) were generated for PBNP which assessed coating debris buildup on the Containment Sump Screens during the Recirculation Phase of a Design Basis Loss of Coolant Accident.

At that time, the greatest contributor to sump screen blockage was thought to be from alkyd coating debris particles.

In those calculations, an assumption concerning the size and shape of the alkyd coating debris has since been proven to be incorrect.

The calculations assumed that the coatings failed to flat circular chips with a 1/8 inch diameter and the thickness of the dry coating film. It has since been demonstrated that such coatings fail to very small particulates and not to larger chips.In applying the assumption of chip fragments, an incorrect calculation of the specific surface area (SJ) was also made, resulting in an incorrect value of head loss using the NUREG/CR-6224 correlation.

Finally, since the NUREG/CR-6224 correlation had been established for debris beds of mixed fiber and fine particulates, it was inappropriate to use it when evaluating the head loss through a bed of larger, flat chips with no fiber content.In September 2005, EPRI published a technical report (Ref. 4 ) that documents experimental results that demonstrate that the debris from alkyd coatings is in the form of fine particulates (rather than flat chips) with a size less than 1200 microns. These experimentally determined alkyd debris particulates would be much smaller than the 0.125" (>3000 micron) screen perforation size. Using this revised shape and size information results in the conclusion that the alkyd coating debris would not cause a head loss across the containment sump screens following a design basis event because the particles would pass through the screens unimpeded.

P13F-1649 Revision(0 11/29/00 Refermnce:

NP7.2.10 Engineering Eval. No. 2005-0024 Revision I Point Beach Nuclear Plant ENGINEERING EVALUATION Page No. 4 of 11 Initials TCK Date 11/9/2005 3.0 INPUTS AND ASSUMPTIONS I. Size of alkyd coatings debris particle following a LOCA 1128 microns 2. Size of epoxy coatings debris particle following a LOCA 0.125" 3. Area of degraded epoxy coating in Unit 1 5 f1t 4. Area of degraded epoxy coating in Unit 2 18 ft 2 5. Maximum RWST level prior to sump recirculation 34%6. Size of sump fine screen perforation 0.125" 7. Inside diameter of sump fine screen 13.5" 8. Required open area of sump fine screen 50%o 9. It is assumed that Case RIA/B for a single train of RHR operation is the maximum flow rate for recirculation operation.

This is rounded up from the -1580 gpm in the source document.

Combined RHRISI "piggyback" flow is now procedurally limited to less than this value. 1,600 gpm 10. Minimum Flood Level at Recirculation Switcliover (used in Ref. I & 2) 2.68 ft 11. Containment Sump Total Flowrate (used in Ref. I & 2)2 trains RHRJSI (concurrent operation) plus CS 4,850 gpm 12. <not used>[Ref. 4, Table 4-11[Ref. I and 2J[Ref. 5][Ref. 6)[Ref. 9 and 10)[Ref. 7)(Ref. 7][Ref. 81[Ref. 14, Att. 7)[Ref. I & 2, Sect. 6.2.1][Ref. I & 2, Sect. 4.2.1]13. <not used>14. Unit I Recirculation Flow Velocity Results 15. Unit 2 Recirculazion Flow Velocity Results 16. Specific Gravity of Brady B-500 Labels (yellow and black labels)17. Specific Gravity of Brady B-950 vinyl tape (red and white labels)18. Specific Gravity of Green Duct Tape n/a n/a[Ref. 1, Table 2A][Ref. 2, Table 2A]1.1[Att. D]1.3 1.1[Att. E][Att. DI PBF-1649 Refcrence:

NP72.10 Engineering Eval. No. 2005-0024 Point Beach Nuclear Plant Revision I ENGINEERING EVALUATION Page No. 5 of 11 Initials TCK Date 1119/2005

4.0 REFERENCES

1. Calculation M-09334-345-RH.1, Rev. 1 "Containment Sump Blockage Due To Failure Of Unqualified/Undocumented Coatings" (Unit 1)2. Calculation M-0933443 1-RH.!, Rev. 0 "Containment Sump Blockage Due To Failure Of Unqualified/Undocumented Coatings" (Unit 2)3. Calculation 2000-0044, Rev. 3, "Containment Accident Sump Level as a Function of RWST Draindown.

4. EPRI Technical Report 101 1753, "Design Basis Accident Testing of Pressurized Water Reactor Unqualified Original Equipment Manufacturer Coatings.5. NPC 2005-00627, Unit 1 Post Emergent Work Coating Status (Attachment A. 1 page)6. NPC 2005-00628, Unit 2 Coating Status (Attachment B, 1 page)7. DWG M-276, Rev. 2, Containment Safety Injection Sump Requirements for Screens B. USNRC Regulatory Guide 1.82, Rev 0, June, 1974 9. BG.EOP 1.3, Rev 21 "Transfer to Containment Sump Recirculation" 10. BG-EOP-1.4, Rev. 11, "Transfer to Containment Sump Recirculation -High Head" 11. NPL 2005-0193, Design Information Transmittal to AREVA 12. John Crane, Bulletin No. S-1/lB, Type 1/IB Elastomer Bellows Seals (Attachment C, 6 pages)13. NUREGICR-2792, "An Assessment of Residual Heat Removal and Containment Spray Pump Performance Under Air and Debris Ingesting Conditions".

14. Calculation N-92-086, Rev. 3, "ECCS Pump NPSH Calculation." 15. Reg. Guide 1.82, Rev. 1.16. NLTREG/CR-6773, "GSI-191:

Integrated Debris-Transport Tests in Water Using Simulated Containment Floor Geometries."*

17. T-Track Corrective Action item LAR 046129, "Potential Blockage of Safety -Related Strainers by Material Brought Inside Containment" 18. E-mailed report of Brady B-950 Vinyl tape specific density testing results (Attachment E)19. WO 0502011, "Service Water Piping,21'

& 8' Elev." -summary of activities in Attachment F 20. NAVCO Piping Datalog, I Ph edition. National Valve and Manufacturing Co.21. Bechtel Drawing C-2122, Rev. 1: Liner Plate Plan, Elevation, and Wall Penetration Schedule 22. Bechtel Drawing C-2126, Rev. 3: Container Liner Plate Floor Plan PBF-1649 Revision 0 11/29/O0 Reference; NP72.10 Engineering Eval. No. 2005-0024 Point Beach Nuclear Plant Revision I ENGINEERING EVALUATION Page No. 6 of 11 Initials TCK Date I 19/2005 5.0 EVALUATION AND RESULTS Total Wetted Fine Screen Area The surface area of the fine screen is that of a right cylinder with a diameter of 13.5" (Input 7).Ascreen -I .h = 1h = (3.53 h) ft2 1 2 The minimum height of the immersed screen can be determined by applying the relationship generated in Reference 3 between sump height and RWST volume expended.Lsnmp = 6 2.8 3-0.7 1-LRWST Switchover to containment sump recirculation is directed at an indicated RWST level of 34%. The derivation of this formula already conservatively takes into account instrument uncertainty, so no additional uncertainty need be applied. Therefore:

Ls, =62.83-0.71-LRjysT =62.83-0.71-34

= 38.69inches

=3.22ft Therefore the minimum wetted surface area is: Ascreen (3.53. h) ft 2 = 3.53

• 3.22 = 1 1.37 ft 2 This minimum area would be increased from the time of switchover to sump recirculation as the containment sump continued to fill by containment spray drawing from the RWST down to 9% indicated level. Additionally, the actual sump level would be expected to be higher since the volumes of the spilled RCS and Safety Accumulators have been omitted from the above calculations.

Allowable Blockage of Fine Screen Area Paragraph C.7 of USNRC Regulatory Guidel.82 Revision 0 states that "A vertically mounted fine inner [ECCS] screen should be provided...

The available surface area used in determining the design coolant velocity should be at based on one-half of the free surface area of the fine screen to conservatively account for partial blockage...".

This is consistent with the design philosophy used for the PBNP strainers.

Applying this standard to the total wetted area of 11.37 ft 2 yields a maximum allowable screen blockage at the time of initiating sump recirculation of: A.,,,,,, Bl&,okn = 0.5

• As,,, = 0.5

• 11.37 ft 2 = 5.68ft 2 Particle Sizes of Coating Debris Per Reference 4, the maximum particle size for alkyd coatings debris is 1128 microns. This is much smaller than the screen perforation size of -3000 microns and thus will not contribute to screen blockage.

Therefore, the submergence of 3.22 ft and corresponding wetted area of 11.37 ft 2 are applicable.

Zone of Influence The Zone of Influence

('zor', also termed "Zone of Transport")

was established in References I and 2. The ZOI is that region in where a coating fragment, drifting under the influence of a horizontal velocity field while it sank, could reach the surface of the ECCS screens prior to reaching the floor of containment.

From this definition, it is clear that a first-order variable in determining the size of the ZOI is the velocity of water. Since the velocity at any given point in containment is proportional to the flow rate of water being drawn out of the sump, it is possible to directly scale the size of the ZOI for various withdrawal rates.The difference between the total flow rate used in the Ref. I and 2 calculations and the one train flow rate (Input 9) is used to scale the results from Ref. I and 2 to more accurately reflect the specific scenario.

To determine the ratio, the total flow rate used in Ref. 1 and 2 (4,847.1 gpm, Input 11) is divided into the single train flow rate (1,600 gpm): PBF-1649 Revision 0 11/29/00

Reference:

NP7.2.10 Engineering Eval. No. 2005-0024 Point Beach Nuclear Plant Revision I ENGINEERING EVALUATION Page No. 7 of 11 Initials TCK Date 11/9/2005 SF,,,1,600 gprn -0.330 4,847.1 gp=0 A second order variable is the depth of the sump water. Between the minimum and maximum analyzed depths there was a 131% increase in depth. The following tables, using data extracted from page 31 of References 1 and 2 respectively, illustrate that the variation in ZOI dimensions are much less than linearly variable with depth. As will be discussed later, only epoxy coatings are of concern, so these tables are limited to only epoxy coatings: Unit I ZOI at ZOI at ZOI for Epoxies 2.68' depth 6.18 depth Ch. 1 -Ch.2 -Sump 4.2 3.2 Ch. 5 4 Ch. 4 -Ch. 3 -Sump 4.8 4.8 Ch. 7 4 Ch. 8 -Ch. 3 ->Sump 4.7 4.9 Ch.8 -Ch.6 -Ch.3-)Sump 4.7 4.9 Unit 2 ZOI at ZOI at ZOI for Epoxies 2.68' depth 6.18' depth Ch. 11 i Ch.14 4 Sump 7.3 6.6 Ch. 12 i Ch. 13 -3 Ch. 14 e Sump 6.6 6.6 Ch. 37 -Ch. 38 4 Sump 4.8 5.4 Ch. 12 i Ch. 15 e Ch. 38 e Sump 4.8 5.4 This information will be used in the following section when scaling the S&L determined ZOI to account for lower velocities due to reduced sump flow rates.Degraded "Qualified" Debris Inventory in the Zones or Influence Zones of Influence were determined in References I and 2 and these were used to determine locations to be inspected for degraded epoxy. The results of the inspections can be found in References 5 and 6. As discussed above, alkyd coatings will not contribute to screen blockage.

Therefore, the debris inventory to be used for the evaluation of screen blockage will comprise of epoxy coatings exhibiting degraded /failed conditions.

Epoxy coatings exhibiting single incident effects such as"dings" and mechanical impact will not be included in the evaluation of screen blockage since these defects are not indicative of bond failure or delamination of the coating. Of the acceptable

("qualified")

coatings, only those exhibiting poor adhesion as evidenced by delamination, blistering, or peeling are considered susceptible to transport to the sump as intact chips or flakes.Reviewing the containment coatings inspection report against these criteria found one area within the Unit 2 ZOI (as established in Reference

2) that contains delaminating qualified coatings.

While much of these were removed under WO 0502011 (Reference 19), a small amount remained above the 21 ' elevation, and an estimated

-6 linear feet of degraded coatings (-3' each on two pipes) below the 21' elevation.

The piping of concern are 2" nominal sized pipe that are the Service Water returns from the cavity cooling coils. Per Reference 20, 2" piping has an outside surface area of 0.62 square feet per linear foot of run. Assuming that 8 linear feet of degraded coatings remain, this would amount to -5 square feet of degraded coatings.It is highly unlikely that if all 5 ft 2 were to fail and transport to the sump screens, that they would form a uniform impervious layer of the same surface area. Random distribution effects would preclude such an occurrence.

Furthermore, since all of the particles formed would be originating from the same location, they would all tend to be deposited in a confined area on the screen surface, if in fact they were capable of being transported there in the first place.Furthermore, the reduced ZOI due to lower velocities than those calculated in References I and 2 places all of the degraded coatings of concern outside the actual ZOI. The piping with the degraded coatings originate at penetrations P45 and P46.From Reference 21, these are located at the 1460 55' 30" and 149° 22' 30 azimuths respectively.

PBF-1649 Revision 0 11129/1o Refercnce:

NP 72.10t I Engineering Eval. No. 2005-0024 Point Beach Nuclear Plant Revision I ENGINEERING EVALUATION PageNo. 8 of 11 Initials TCK Date 1119/2005 Reference 22 depicts the N/S and E/W coordinates of the containment sump strainers relative to the containment center. The line closest to the area of the degraded coatings is located 14' 6" East, and 44' 0" South of the containment geometric center.This gives the outlet radial coordinates of 46' 4" radius on the 1610 48' azimuth.The angular separation between the closest penetration and sump screen is -12.5 degrees. At a radius of 46' 4", the horizontal distance between the penetration at -149° 22' and the closest sump screen is at 161° 48' is -10 feet. From photographs of the piping penetrations and the adjacent drops to the 8' elevation, it is apparent that the short horizontal runs from the penetrations to the drops are less than 3' in the direction of the sump screens. Therefore, the closest horizontal proximity of the degraded coatings to the sump screens is 7 feet or more.From Attachment G (taken from Reference 2), it is apparent that the flow path for the degraded coatings to reach the sump screens would be Channel 11 to Channel 14 to the sump. From the previous table, the maximum ZOI length for this path occurs at minimum water level, and is 7.3 feet.Since the distance transported horizontally while a fragment is falling is directly proportional to the horizontal velocity of the moving fluid, the size of the ZOI due to flow changes can be scaled by the ratio of the expected velocity to the reference velocity.

This ratio or scaling factor was previously determined to be 0.33.Scaling the 7.3 foot reference ZO from the S&L transport analysis by the 0.33 factor results in a ZOI of just 2.4 feet. Based on this information, even accounting for scatter as the postulated fragments drop to the sump. it is clear that none of the fragments would be within the actual Zone of Transport for the ECCS sump screens.Based on this information, no qualified but degraded coating fragments are expected to reach the ECCS sump screens on either unit.Evaluation of Potential Effects of Decomposition Particles from Unqualified Coatings The total inventory of unqualified coatings is bounded by 22,000 square feet per containment (Ref. 5 and 6). While some of this inventory may actually be epoxy coatings lacking sufficient documentation to consider them acceptable, it is conservative to assume that they are alkyd based coatings because it increases the inventory of suspended particulates that must be considered.

If actually formed of epoxy, they would fail as larger chips not available for transport, or would erode under a steam jet to a size comparable to alkyd particles and still be small suspended particulates.

It has been previously established that alkyd coatings have a thickness of 0.015" (Ref. I and 2). Therefore, the total volume of fine particulate debris that may be generated by all unqualified coatings is bounded by: Vcoating ~= (0.015 in / 12 ft/in) x 22,000 ft 2 = 27.5 ft 3 As previously established in this evaluation, the minimum level in containment when sump recirculation is established is in excess of 38.69 inches above the 8' elevation.

From Reference 3, this corresponds to a sump volume of: Viump =[8 ft +(38.69 in I 12 in/ft) -7.93 ft] / 2.1E-5 = 156,900 gallons 20,970 ft 3 The volumetric fraction of the suspended debris fines from coatings would therefore be a maximum of: V..tings / Vsnp = 1.31E-3 = 0.13%Considering the very fine nature of the particles, and the dilute concentration, the resulting suspension would be best characterized as an aqueous suspension with negligible changes in hydraulic properties (e.g. viscosity and density) and thermodynamic properties from clean water. This assertion is further supported by information contained in NUREG/CR-2792 showing only a 1% loss in hydraulic efficiency with a 1% mass in a slurry (Ref. 13).PBF-1649 Revision O 11/29/00 Rcrerencc:

NP7.2.10 Engineering Eval. No. 2005-0024 Point Beach Nuclear Plant Revision I ENGINEERING EVALUATION Page No. 9 of 11 Initials TCK Date 1119/2005 The mechanical seals of the RHR and SI pumps are John Crane Types 1 and IB mechanical seals. The design of these seals uses a non-clogging single-coil spring to supply the seal face closing force. According to the manufacturer's literature (Ref.12). this design and arrangement is not affected by the build-up of solids, and the design is suitable over a wide range of service conditions.

The same literature indicates that Types I and IB seals are used in power generation and other demanding applications such as pulp and paper, petrochemical, food processing, and waste water treatment.

Therefore failure of these seals due to operation with fluids containing suspended coating decomposition particulates is not expected.A review of the other components cited in Reference 11 found that all orifice sizes in the credited flow paths are on the order of inches, and are therefore not subject to blockage.

Additionally, after establishment of containment sump recirculation, no valves in the flow paths are credited with an active function to reposition, all are 2 or more inches in size, and all have stainless steel or harder wearing surfaces.

Based on this, wear, erosion, and blockage of valve components is not considered a concern.Lastly, as documented in Reference I 1, all core and reactor internal flow passages of concern are on the order of fractional inches or larger. They are judged to not be in jeopardy of fouling by the sub-millimeter sized decomposition particles.

General Approach to Adhesive Label I Tape Transport Evaluation The general approach used to determine if adhesive labels/tape can be transported to the containment sump screen is to use the information in Ref. 15 with the results determined in Ref. 1 and 2 (as scaled by the results determined in this evaluation and in Ref. 14). The basis for scaling the results is that Ref. I and 2 used a conservatively high containment sump total flowrate (2 trains of RHRJSI plus Containment Spray), which is a configuration that is not currently allowed at Point Beach. Additionally, the results in Ref. I and 2 used a conservatively low water level for recirculation switchover.

A more appropriate switchover level is determined in this evaluation.

As stated on Page 1.82-3 of Ref. 15, if the velocity of water traveling to the screen is at or below approximately 0.2 ft/sec, debris with a specific gravity of 1.05 or more is likely to settle before reaching the screen surface and would not present potential clogging of sump B screen. The basis for the 0.2 ft/sec is further supported by Spherical Tracer Movement Tests presented in Appendix B of Ref. 16. Table B-2 of Ref. 16 indicates that for three types of spheres, there was no movement seen until the flow velocity exceeded 0.2 ft/sec.The specific gravity of the majority of the labels in containment (Brady B-500 (also referred to as yellow/black conduit/tray labels] is larger than 1.05 (Input 16). The lowest specific gravity of duct tape was determined to be larger than 1.05 (Input 18).While the uncertainty of the measurements documented in Attachment D for the B-500 labels encroach on a specific gravity of 1.05 (as low as 0.99), the mid-span is well above 1.05. Further, previous testing (documented in Ref. 17) determined a specific gravity for these same labels of 1.15. Finally, discussion with the technician confirmed that the tested labels sank, and that the span of uncertainty was due to the resolution of the graduations on the container used to measure the liquid displacement.

Initial testing of the Brady B-950 tape [also referred to as red/white conduit markers])

indicated that its specific gravity was less than 1.0 (See Attachment D). However, follow-up testing demonstrated that it is actually -1.3 (Input 17).Since all labels and tape are expected to sink, and have specific gravities greater than the 1.05 threshold described in Reference 15, it is necessary to determine whether the flow velocities in containment are similarly bounded by the 0.2 fps threshold of the same reference.

The flow velocities that would exist in containment during sump recirculation conditions will now be evaluated to see if they too are within the screening bases of Reference 15..Calculated Sump Water Level From above, the minimum calculated sump water level for switchover to recirculation operation is 3.22 ft.Scaling Factor For Sump Water Level The difference between the sump water level used in the Ref. I and 2 calculations and the sump water level determined in this evaluation is used to scale the results from Ref. I and 2 to more accurately reflect the specific scenario.

To determine the ratio, the sump water level used in Ref. I and 2 (2.68 ft. Input 10) is divided by the sump water level determined in this evaluation (3.22 ft): PBF-1649 Rcevision 0 11129/00 Referencc:

NP7.2.10 Engineering Eval. No. 2005-0024 Point Beach Nuclear Plant Revision I ENGINEERING EVALUATION Page No. 10 of 11 Initials TCK Date 11/9/2005 SF Waler ln-e = 2.68 ft = 0.832 jump 3.22 ft Scaling Factor For Total Flow Rate The difference between the total flowrate used in the Rcf. I and 2 calculations and the one train flowrate (Input 9) is used to scale the results from Ref. I and 2 to more accurately reflect the specific scenario.

To determine the ratio, the total flowrate used in Ref. I and 2 (4,847.1 gpm, Input 11) is divided into the single train flowrate (1,600 gpm): SF -ialflowrtc 1,600 gpnl = 0.330 anE lo rtc 4,847.1 gpm Combination of Scaling Factors The two scaling factors determined above for sump water level and total flow rate can be combined to create a single scaling factor to be used to adjust the results from Ref. I and 2 to more accurately reflect the specific scenario.

These factors are multiplied together to obtain a single scaling factor as follows: SF m = wer let-el

• SF ,, =0.832

• 0.330 = 0.275 Scaling of Recirculation Flow Velocities The limiting Recirculation Flow Velocity for Unit I (Input 14) are in channels 5 and 10 with 0.29 fps each. Applying the above scaling factor, the limiting flows are reduced to 0.08 fps.Similarly, the limiting Recirculation Flow Velocity for Unit 2 (Input 15) is in channel 14 with 0.33 fps. Applying the above scaling factor, this flow becomes 0.09 fps.In both containments, the highest velocities are considerably below the 0.2 fps horizontal velocity threshold established in Reference 15 for the transport of debris. Therefore it is concluded that tape, labels, and similar miscellaneous debris are not transportable to the ECCS sump screens.

6.0 CONCLUSION

S AND RECOMMENDATIONS The total potential debris coating of the Unit 1 and Unit 2 sump fine screens originating from failed coatings is 0 ft 2.Additionally, the coating decomposition particulates would not interact with any film-like debris to form a denser or more impervious layer than just the film itself (i.e. no compounding effect).The suspended coating decomposition particulates do not pose a challenge to the long term functional capabilities of the Emergency Core Cooling strainers, downstream components, or continued cooling of the core after a design basis LOCA event.The existence of various potential adhesive debris (labels/tape) in the Unit I and Unit 2 containment will not threaten the operation of the containment sump screen. This conclusion is based on the fact that the specific gravity of the debris is such that with the expected flow velocities around the containment sump during recirculation operation, the debris from labels and/or tape will settle out to the floor of containment prior to reaching the sump screen.Based on the above considerations, the containment sump screens are considered fully capable of providing the design basis flows to the RHR pumps without an excessive head loss or a risk of air ingestion.

PBF-1649 Rcvision O 11/29/00 Referencc:

NP7.2.10 Point Beach Nuclear Plant ENGINEERING EVALUATION Engineering Eval. No. 2005-0024 Revision I Page No. 11 of 11 Initials TCK Date 11/912005 This evaluation has demonstrated analytically that the horizontal flow velocities in the containment at the time of switch-over to containment sump recirculation would be too small to physically transport any other than very fine particulate debris to the screens. Such debris would necessarily be too small to present a challenge of screen blockage.This should not, however, be taken as an absolute determination of operability.

While there is considerable margin incorporated in (and demonstrated by) this analyses, the safety significance of being able to ensure continued recirculation flow can be maintained dictates that there be analytical defense in depth.As a result, a companion OPR (OPR 162, revision 1) has been prepared to demonstrate that the ECCS sump screens can function acceptably with an assumed limiting debris loading.PBF-1649 Revision 0 t tr-9100 MP 71 2-cvaw .ar .....

0 NW i. INTERNAL.CORRESPONDENCE Com~mkirtdioNnclotEcellenceI

• EA~IzK, Cp I.-IQ~e ( N NPC 2005-00627 To: Toni Kendall From: David Taylor i I1 Date: October 30, 2005

Subject:

Unit 1 Post Enmergent Work Coating Status Copy To: Jeremy Fischer, Scott Manthei, Chuck Richardson, Gary Shenvood, Joe Loor, Rod Hopkins, Mark Ralph, Felicia Hennessy, File The following is a list providing the current inventory for the degraded epoxy coatings in the Unit I containment within the Zone of Influenice (ZOI) defined in Sargent & Lundy (S&L) Calculation M-09334-345-RII.

1, Figure 10, "Zone of Influence for Epoxies to Impact Sump Screen Before Settling to Floor".This list was originally derived from the UIR28 C6ntainmcnt Coating Assessment (reference the report attached to WO 0301039).

The list was subsequently transferred into an Excel spread sheet, verified correct, and adjusted according to worl performed during UIR29..!Elevatioh/Cubicde'

-.; Azimuth> i'ft i .; ; ' .' ti-Containment Elevation 8 Az 20 2 Delaminated coating on k-eyway wall, removed.Liner plate, delarninated top coat of 1/4 inch diametdr, Containment Elevation 8 Az 20. El 1)2 1 removed,.

no rust.Containment Elcvation 66 Az 10, l 2 Dines 6 n liner plate.; Total.i It should be noted that while there is a s'mall quantity of abraded-epoxy floor coatings in close proximity to the sump screens, the abraded epoxy floor coatings are not considered to be within the ZOI for epoxies to impact the sump screens before settling to the floor. This is because the floor coatings are equivalent to epoxy coatings that have settled out on ihe floor, and based on the S&L calculation, epoxy coatings wvill not slide along the floor or othenvise betransported to accumulate at the screens.Based on a review of the U2R27 Contaihiment Coating Assessment (NPM 200570340) and the UIR29 walkdow%,ns conducted per WO 0415370, a bounding value for either unit for the total amount of unqualified coatings [steel substrate itenmis have unqualified alkyd coating per S&L Calculation M-09334-345-RH. 1, Attachment A, Page A2) within each containment is 22,000 square feet. This is based on the actual limiting value of 21,825.5 square feet from the U2R27 Assessment, and an estimated 20,000 square feet from the UIR29 walkdowns.-Gil io/38/oF Prepared:

David Tay Veri'fied:

Joe Loor'7,('Approved:

Felicia Hennessy~t/ CIY5/c~j C o_ INTERNAL etc 4 Committed to Muclow Erccfln~cc

} CORRESPONDENCEif B NPC 2005-00628 To: Gary Shenvood From: David Taylor Date: October 30, 2005

Subject:

Unit 2 Coating Status Copy To: Jeremy Fischer, Scott Manthei, Chuck Richardson, Tom Kendall, Rod Hopkins, Mark Ralph, Felicia Hennessy, File The following is a list providing the cujrrent inventory for the degraded epoxy coatings in the Unit 2 containment within the Zone of Influence (ZOI) defined in Sargent & Lundy (S&L) Calculation M-09334431 -RH.1, Figure 8, "Zone of Influence for Epoxies to Impact Sump Screen Before Settling to Floor".This list was originally derived from the U2R27 Containment Coating Assessment (reference NPM 2005-0340). The list wvas subsequently transferred into an Excel spread sheet and verified correct.*ai.C *. :. Azimuth .t- 't

• f a:. "'- .ot : .: Containment Elevation 8 Az 161, El 13 3 Liner plate. IO dings (distributed).

Liner plate, mechanical damage to vertical channel Containment Elevation 8 Az 175. El 12 3 .seanm Containment Elevation 66 Az 1.6I 0to 170 10 Liner plate, dings Containment Elevation 66 Az 173. El 71 2 Liner plate. mechanical damage and difigs in area.TotalI S _It should be noted that while there is a small quantity of abraded epoxy floor coatings in close proximity to the sump screens, the abraded epoxy ifloor coatings are not considered to be within the ZOI for epoxies to impact the sump screens before settling to the floor. This is because the floor coatings are equivalent, to --epoxy coatings that have settled out on the floor, and based on the S&L calculation, epoxy coatings will not slide along the floor or otherwise be' transported to accumulate at the screens.Based on a review of the U2R27 Containment Coating Assessment (NPM 2005-0340) and the UIR29 walkdoxvns conducted per WO 0415370, a bounding value for either unit for the total amount of unqualified coatings [steel substrate items have unqualified alkyd coating per S&L Calculation M-09334-431 -RH. 1, Attachment A, Page A21 within each containment is 22,000 square feet. This is based on the actual limiting value of 21,825.5 square 'feet from the U2R27 Assessment, and an estimated 20,000 square feet from the U1R29 walkdowns., v{ 10/3o/3/Os' Prepared:

David Tay , .1 .t Verified:

Tom Kendall Approved:

Felicia ltenesy 2_0QS-O- oLL wN-pc, '~ C-'~AlOFnI E A -Face/Primary Ring B -Spring C -Elastomer Bellows D -Retainer E -Drive Band F -Spring Holder G -Disc FYPE 1/1B lastomer Bellows Seals W~~I.. .'*1'It.-g**t* I.'.?..*.

...f , MI The John Crane Type 1 Elastomer Bellows Seal is widely recognized as the industry workhorse with a proven track record of exceptional performance.

The Type 1 is suitable for a wide range of service conditions:

from water and steam to chemicals and corrosive materials.

• For use in pumps, mixers, blenders, agitators, refrigeration compressors, blowers, fans and other rotary shaft equipment.

• For pulp and paper, petrochemical.

food processing, wastewater treatment, chemical processing, power generation and other demanding applications.

• Temperature:

-40'C to 205'C-40F to 400F (depending on materials used)* Pressure: 1: Up to 29 bar 9/425 psig 1B: Up to 82 bar 9/1200 psig* Speed: See enclosed Speed Limits chart.I aM* Mechanical Drive -Eliminates overstressing of bellows.* Self-Aligning Capability

-Automatic adjustment compensates for abnormal shaft end play runout, primary ring wear and equipment tolerances.

• Special Balancing

-For higher pressure applications and less wear.* Non-Clogging, Single-Coil Spring -Not affected by buildup of solids.* Low Drive Torque -Improves performance and reliability.

LfOQ'\exxl4-Loo !i -D bL4 ?kze- jW I}cC?-=A WA F TYPE 1/1B Elastomer Bellows Seals If

• I I For case or installation, the lead-in edge or shaft or sleeve should be chamrered as shown.a i, i I.M fTI 111I Seal Size/D1 (inches)1.000 1.125*1.250 1.375 1.500 1.625 1.750 1.875 2.000 2.125 2.250 2.375 2.5WO 2.625 2.750 2.875 3.000 3.125 3.250 3.375 3.500 3.625 3.750 3.875 4.000 D3 1.500 1.625 1.812 1.875 2.000 2.250 2.375 2.500 2.625 2.812 2.937 3.062 3.187 3.375 3.500 3.625 3.750 4.000 4.125 4.250 4.375 4.500 4.625 4.750 4 .75 D4 1.750 1.875 2.000 2.125 2.250 2.500 2.625 2.750 2.937 3.125 3.250 3.375 3.500 3.750 3.875 4.000 4.187 4.437 4.562 4.687 4.812 4.937 5.062 5.187 5 312 L 1.562 1.625 1.625 1.687 1.687 2.000 2.000 2.125 2.125 2.375 2.375 2.500 2.500 2.750 2.750 2.875 2.875 3.125 3.125 3.125 3.125 3.250 3.250 3.375 3.375 L38 0.375 0.375 0.375 0.375 0.375 0.375 0.375 0.375 0.375 0.500 0.50o 0.500 0.500 0.500 0.500 0.500 0.5s0 0.500 0.500 0.500 0.500 0.562 0.562 0.562 0 562 7 4 5 o9- Oe(:;LA -.L XDO&- I I,-2_ Rcl>;, / TYPE 1/1 B Elastomer Bellows Seals____/D4 MIN. BORE D MAX.-[03 0 2 MAX.SEAL OD L650 40.l 102'0.125-

'i0.002'SIAFT DIA. _3+0.031'For ease of installation.

the lead-in edge of shaft or sleeve should be charnfered as shown.a .r I i1aM =. I.M T -Seal SizelDl (inches)1.000 1.125 1.20 1.375 1,500 1.625 1.750 1.875 2.000 2.125 2.250 2.375 2.500 2.625 2.750 2.675 3.000 3.125 3.250 3.375 3.500 3.625 3.750 3.875 A nnnf D2 0.875 1.000 1.125 1.250 1.375 1.500 1.625 1.750 1.875 2.000 2.125 2.250 2.375 2.500 2.625 2.750 2.875 2.875 3.000 3.125 3.250 3.375 3.500 3.750 3A .75 D3 1.500 1.625 1.812 1.875 2.000 2.250 2375 2.500 2.625 2.JL 2.937 3.062 3.187 3.375 3.500 3.625 3.750 4.000 4.125 4.250 4.375 4.500 4.625 4.750 A 475 D4 1.750 1.875 2.062 2.125 2.250 2.500 2.625 2.750 2.937 3.125 3.250 3.375 3.500 3.750 3.875 4.000 4.187 4.437 4.562 4.687 4.812 4.937 5.062 5.187' 312 L3 L65 1.875 0.343 1.937 0.343 1.937 0.343 2.000 0.343 2.000 0.343 2.375 0.437 2.375 0.437 2.500 0.437 2.500 0.437 2.750 0.500 2.750 0.500 2.875 0.500 2.875 0.500 3.125 0.562 3.125 0.562 3.250 0.562 3.250 0.562 3.500 0.625 3.500 0.625 3.500 0.625 3.500 0.625 3.625 0.625 3.625 0.625 3.750 0.625 3.750n 0 A2

'Lt~t -O) .L, 1?3 -CA4 ARF TYPE 1/1B Elastomer Bellows Seals M.Pressure (bar g)or ..Pressure (psig)1., nn VJ 69 55 41 28 14-.* :... ..t.. e

• Contact John Crane Engineering

.... ..........

,,. -_. , _: ..3 .a * ; .I. -Q To determine the maximum pressure for the Type 1 or 1B required, multiply the maximum pressure by the Multiplier Factors to obtain the maximum operating pressure.-1000---MMIIII Type sB i Carbon vs. Ceramic Type I B Carbon vs. Silicon Carbide Carbon vs. Silicon a b e ~ 1 .600 400 203 I Carbon vs. Ceramic I I= _-C D I II___ _ _ l _ I _ _ O 0 25 51 76 102 127 152 (mm)I I I I I I I 0 1 2 3 4 5 6 (inches)Seal Size The Basic Pressure Rating is based on a standard Type 1 or Type 1B seal installed according to the criteria given in this data sheet and according to generally accepted industrial practices.

The Basic Pressure Rating assumes stable operation at 1800 rpm in a clean, cool. lubricating, nonvolatile liquid, with an adequate flush rate. When used with the Multiplier Factors, the Basic Pressure Rating can provide a conservative estimate of the dynamic pressure rating.Contact John Crane Engineering for process services outside this range and with more detailed application information in order to obtain the actual dynamic pressure rating.311 M M. .I., Selection Considerations Multiplier Factor Speed 1800rpm x 1.00 Above 1800 rpm_Sealed Fluid Petrol/Gasoline, Kerosene.

or Better x 1.00 Lubricity Water and Aqueous Solutions x 0.75 Flashing Hydrocarbons, .x 0.60 (Specific Gravity <0.65)Sealed Fluid Below 79'C/175-F X 1.00 Temperature From 79 C to 121 C/175¶F to 250'F x 0.90 (for carbon From 121 'C to 1 77'CI250'F to 350'F x 0.80 only) Above 177-C/350'F

• x 0.65 Example for Determining Pressure Rating Limits Seal: 76mm/3 diameter.Type lB Product Water Face Material:

Carbon vs. Silicon Carbide Temperature:

16'C/60'F Shaft Speed: 1800 rpm Using the Basic Pressure Rating Chart. the maximum pressure would be 55 bar g1800 psig.From the Multiplier Factors chart. apply the multipliers for the specific service requirements to determine the maximum dynamic pressure rating for the application.

55 bar g/800 psig x 1 x 0.75 x 1-41 bar g1600 psig At 1800 rpm with the service conditions noted, a 76mm/3' diameter Type IB seal has a dynamic pressure rating of 41 bar g/600 psig. If operating pressure exceeds this dynamic rating. consult your John Crane Sales/Service Engineer.* Multiplier

.1800/new speed Example: If new speed -2700 rpm Multiplier

.1800/2700

.0.67" The ratio of sealed pressure to vapor pressure must be greater than 1.5, oherwise consult John Crane. II the specific gravity is less than 0.60. consult John Crane.II II II III II III V-5 I cex' i 1~j ( ~A k I 0 .2k0OS;-ooz'-I

?, CS1~koe TYPE 1/1 B Elastomer Bellows Seals_I am Mg I MSM For Starting Torque Power Consumption.

consult John Crane Engineering.

a:7 MIM 32W I -Pressure (,psig)2000 TYPE 1B 1500 TYPE 1 1000 500 0 Pressure (bar g)Ij [ 120-100 I M 80 ftftt~~ 60 20 50 75 100 125 (mm)2 3 4 5 (inches)0 1 Seal Size VM IMV a I -Speed (rpm)6000 Welded End 5000 =~ Coil Spring -l*iii~ Wel 4000 _End Standard Spring _ _!3000 2000 Cc 100 0 1 2 3 Seal Size I Coil spring requirements for rotating Type 1 and Stationary IB seal heads with standard stuffing box diameter.Seal Head rg and Adapt L I d End pring-4 z 5 6 (inches)RafM F .MI IIFM ShaftlSleeve Limits Surface Finish 1.00' to 3.125' dia. / 63 Ra 3.125' dia. & up /32 Ra Ovality/Out of Roundness 0.051mm/0.002'(Shaft)End PlaylAxial

+/-0.13mmI0.005 Float Allowance II 0I i II II 2.oD~?I- c ec- CasIJ kiF~@nr'TYPE 1/1B Elastomer Bellows Seals& S I M SEAL COMPONENTS MATERIALS Description Standard

• Options Face/Primary Ring Carbon Antimony-impregnated Carbon Tungsten Carbide Silicon Carbide Retainer 18-8 Stainless Steel Monel'Disc Alloy 20 CB-3 SS Drive Band 316 Stainless Steel Spring Holder Bellows Buna-N Aflas'Fluoroelastomer Ethylene Propylene Neoprene' (Chloroprene)

Springs 18-8 Stainless Steel Monel Alloy 20 CB-3 SS 316 Stainless Steel Anas is a registered trademark of Asahi Glass Co.. Ud.Monel Is a registered trademark of Inco Alloys International, Inc.Neoprene is a registered trademark of DuPont.Type 1/1B elastomer bellows seals can be customized for specific installations after review and evaluation by John Crane Engineering.

The following data is needed to evaluate the proposed service: ** Make and Model of Equipment* Shaft or Sleeve OD* Direction of Shaft Rotation Viewed from Drive End m Seal Cavity Dimensions v Speed a Process Fluid* Specific Gravity* Box Pressure* Vapor Pressure* Temperature

• Viscosity Asia Pacific Singapore Tel: G6-8G1-.1288 Fax: 65.862.4117 Europe. Middle East. Atrica Slough. UK TeI: 44-1753-224000 Fax: 44-1753-224224 Latin America Sao Paulo. PBrazil Tel: 55-11-3049.9979 Fax: 55-11.3849-4511 North America Morton Grove. Illinois USA 1.800-SEALING Tel: 1-847.967-2400 Fax: 1-847.9G7-3915 smuths A Dan: Ci Sm.1hs .:rovp sk.For your rmeaiest John Crane lacit:y. please contact one 0! the locn-ions above.It tnu produclas eatred vvill te used in a potxrnrllty da..ngerous ano, hazadous process. ysut Jon Crare repiuseltaOve should k. consulted prio, to th. sadection anr ue.hi the iarest of continuous devebpminen.

Jolhn Crane Conirmpntub resee the righ to sctr sOusigos tNd 5peciricatiu.s vthout prikr nrticc. It is asngcrous to sinke while hsnuting products made hrom P7FE. Old and nww P1FE products must not be InCrisiilded.

2001 John Cranm Pre, SiO1 wwwjohncrane.corn ISO cerfied -5111B Engineering Evaluation 2005-0024 Revision I Attachment D, Page DI Attachment D Results of Specific Gravity Test Performed by Chemistry Department on October 31, 2005 The following page has the results of specific gravity tests performed by the Chemistry Department on various samples of adhesive labels and tape. The green duct tape, red tape, red/white tape (i.e., Brady B-950), and orange/yellow tape were all samples removed from the Unit 1 containment during U1R29. The Brady label (Brady B-500) was a new label that was not previously affixed. The tests were performed with ambient temperature water.PBF-I 649 RevisionO 11/29/00 Reterenme:

NP72.10 Ta.,S~fs,.:,( -ib / 4(CeiL'Point Bciclh Nuclear Plant DATE 1 0

• 3 0 PRIIARY SYSTEMIS SUPPLF.MENTAL l.OG SHEET APPROVED MA , 4cA, , w SAMPLE TIME BY/,Ci, .\)e -.:LZ_ _ _ __ 'I_ -_-. _ _ -le, TIT\L:tg .r~ .\z. J / 9 -Z _ __7 ____o I-7 I -~ c=- ==== __ __ =?BFn'O0: S'-. I Revisicz.

0 '0O:54 CM.'

Engineering Evaluation 2005-0024 Revision 1 Attachment E, Page El Attachment E Results of Specific Gravity Test Performed by Chemistry Department on November 2,2005 The following page has the results of specific gravity measurements performed by the Chemistry Department on red/white tape (i.e., Brady B-950). The measurements were performed with ambient temperature water.PBF-1649 Revision 0 ll129/00 Refcrence:

NP72.10 IFAI. CVA(,CA1oQ USE DZY Pfe4are Arr C -P, Fz, Kendall, Thomas From: Corell, Gary Sent: Wednesday, November 02, 2005 10:12 PM To: Kendall, Thomas Cc: Grossheim, Robert

Subject:

Tape density analysis Tom A sample of red and white tape was delivered to the lab on Wednesday November 2, 2005 by Aaron Guenther.

The tape was cut into small strips and 1.00 g of tape was placed in a graduated cylinder of demineralized water to eliminate air entrapment.

0.8 ml of water was displaced by the mass of 1.00 9 of tape used. The displacement of 0.8 ml by 1.00 g of tape yields a specific gravity of 1.3 g/ml for the tape analyzed.Gary Corell Chemistry Manager PBNP 11/ 712005 vI&¶. a S Observations of the Paint Removal from 4I f6F1 Service Water Piping in Unit 2 Containment Paint was removed from two Service Water lines on the 21' level of Unit 2 Containment under work order 0502011. The lines were 2.375" diameter each, and ran vertically to 7'from the floor. One line then ran horizontally 2' and the other ran 4', with elbows as necessary.

The criteria given was that all paint needed to be removed except the equivalent of 2' of straight run pipe. The elbows and a small section of one line that had a 6" section at 450 from horizontal were estimated to be equivalent to the 2' run. Over 99%of the paint on the straight runs was removed by paint scrapers and putty knives. The remaining minor flecks left on would take days to remove completely with the tools allowed.A nearby Service Air pipe was tested in the same manner as the Service Water piping.Two cuts were made in the paint down to the piping surface, crossing so that they form an "X". Masking tape was applied for approximately 90 seconds, and then ripped off.The resulting removed paint was less than 1/16" as measured perpendicular from either cut with very little removed, which is classified as cateno m 3A or better. This meets the acceptance criteria set by the Containment Coatings engineer.* ;*,, /C I.. .i~w~o ..934.~-HJ '.-,s, , ~~~~~~~~~~i.

-'.............if" ',a*. .........v;*1 'rjcNo- *0933- 4 5@' ' i l; *- *L~~~~~Tage'" ,.I n No :gtF-

• 3-J -:-Zn'fIfunefr ois oInic.;m5 S .e f7? ;i e Settling to Floor

~ii Point Beach Nuclear Plant ENGINEERING EVALUATION COVER SHEET AND REVIEW FORM Engineering Evaluation Number: Title of Engineering2 Evaluation:

2006-0003 ST-850 Valve Closure Forces and 0-Ring Leaklagc.Svstem (CHAMPS ledentifier Codes) UNIT: PB0 01 PBI 3PB2 Rll Original Engineering Evaluation f Rcvised Engineering Evaluation, Revision # _El Supersedes Engineering Evaluation 11 Canceled QA Scope Zl Yes El No Discipline:

El CIVIL Safety Related [El Yes O No El ELECTRICAL EO INSTRUMENTATION AND CONTROL 3 MECHANICAL

.E NUCLEAR El COMPONENTS

.* PRA Fl CHEMICAL AND RADIOLOGICAL El SYSTEMS Pac Inventory:

Page I -2 Form PBF-I 608b Attachments:

None Page 3-7 Form PBF-1649 : Pages through Prepared By: (Qual Required)

Date: Aaron Guenther _ February 6. 2006 Print Namc Signature Reyiewed By: (Qual Required)

Date: Print Name Signature Reviewed By: (Qual Required)

Date: MOrL RLep- OA /7/ 7 C Print Name Signaure Approved By: Date: Print Name 6 Signaturc\

PBF-I 608b Rcvisinn 2 0/105/04 Page i of 7 Rcrerence:

NP 7.2.10 4 Engineering Evaluation:

2006-0003 Revision:

0 Comments And Resolution Reviewer Comments: Resolution:

None.I P'DF-l608b Revision 2 08/05104 Page 2 of _ eN Rcfcrence:

N 7.2. lo Enginecring Eval. No. 2006-000;Point Beach Nuclear Plant Revision 0 ENGINEERING EVALUATION Page No. 3 Of 7 Initials ALG Date 02/06/2006

1.0 PURPOSE

The purpose of this Engineering Evaluation is to demonstrate that the PB 1/2 SI-850A/B, RHR Pump Sump B Suction valves, are capable of hydraulic closure irrequired post accident.

It also demonstrates that when closed, they will prevent gross diversion of Containment Sump B water through a passive failure in the associated Containment Sump B recirculation line both in the short and long term. This Engineering Evaluation is associated with the following PBNP components:

PBI SI-00850A P-IOA RHR PUMP SUMP B SUCTION PBI SI-00850B P-IOB RIHR PUMP SUMP B SUCTION PB2 Sl-00850A P-IOA RHR PUMP SUMP B SUCTION PB2 SI-00850B P-IOB RHR PUMP SUMP B SUCTION

2.0 BACKGROUND

The PBI/2 SI-850A1B valves are normally closed, hydraulically operated valves located inside containment in the line leading from containment sump "B" to the suction of the RHR pumps (P-IOA/B).

These valves perform an active safety-related function in the open position.

The SI-850A/B valves must be capable of opening, by remote manual switch actuation, when transitioning from the injection mode of SI to the recirculation mode of SI. When the RWST is effectively depleted following a LOCA, suction for the Si and RHR pumps is switched to the containment sump to provide long-term core cooling.The SI-850A/B vilvc§ performh an activ'e safety-iclated function to isolate a passive failure in the containment surbp*recirculation line to prevent the gross diversion of containment sunip inventory.

If the credible passive failure were to occur post-accident a Sl-850 valve could be closed in order to: 1) Maintain Containment Sump B inventory, 2)Protect the RHR system and pumps from flooding.

The SI-85OA/B valves must be capable of closing, by remote manual switch actuation, to support these functions.

3.0 INPUTS

AND ASSUMPTIONS Critical dimensions associated with the SI-850A(B)

O-ring (RIO).0-ring dovetail groove (narrow portion) inner diameter:

10.518" 0-ring dovetail groove (narrow portion) outer diameter 10.988"* O-ring dovetail groove depth: 0.231 " PBNP P&ID I IOE017 Sheet 1, Safety Injection System, Unit I (rev. 53)PBNP P&ID 1 OE035 Sheet 1, Safety Injection System, Unit 2 (rev. 49)

4.0 REFERENCES

Ri Westinghouse Document, RFS-M-550, Criteria for leakage protection and maintenance of residual heat removal pumps during safety injection recirculation, Dated September 6, 1967.R2 Stearns and Roger Vendor Drawing 8551/4 Sheet 1.R3 American National Standard For Control Valve Seat Leakage, ANSI B16.104-1976, Approved April 12, 1976.R4 IT 531, Leakage Reduction and Preventive Maintenance Program Test of Containment Sump B Suction Line Mode 5, 6, or Defucled Unit 1, revision 15.R5 IT 536, Leakage Reduction and Preventive Maintenance Program Test of Containment Sump B Suction Line Mode 5, 6, or Defueled Unit 2, revision 18.R6 PBNP Calculation 2000-0018, Parameters for testing SI-850A&B, revision I R7 Parker O-ring Seals Catalog ORD 5700 page B37, Parker Seal Group O-ring Division, 9/92.R8 NMAC Valve Application, Maintenance, and Repair Guide section 4.4, December 1996 PBF-l 649 Revision()

IIl9/00 Rcrcrcncc:

NP7.2.10 Engineering Eval. No. 2006-0003 Point Beach Nuclear Plant Revision 0 ENGINEERING EVALUATION Page'No. 4 of 7 Initials ALG Date 0210612006 R9 Parker O-ring Seals Catalog ORD 5700 page A5-14, Parker Seal Group O-ring Division, 9/92.RIO Stearns and Rogcr Vendor Drawing 10862/8 Sheet 2.R11 Parker O-ring Seals Catalog ORD 5700 page A4-7, Parker Seal Group O-ring Division, 9/92.R12 PBNP Calculation 2000-0044, Containment Accident Sump Level as a Function of RWST Drain Down, revision 3.R13 PBNP EOP 1.3, Transfer to Containment Sump Recirculation

-Lowv Head Injection (rev. 36 for both units), and EOP 1.4, Transfer to Containment Sump Recirculation

-High Head Injection (rev. 17 for Unit I and rev 18 for Unit 2).R14 Westinghouse document WEP-97-522,PBNP Unit 1 and 2 Containment Analysis Assuming Reduced Fan Cooler Performance, dated May 29, 1997.R15 Crane Technical Paper No. 410, Flow of Fluids.R16 Parker O-ring Seals Catalog ORD 5700 page A4-11, Parker Seal Group O-ring Division, 9i92.R17 Parker O-ring Seals Catalog ORD 5700 page B46, Parkcr Seal Group O-ring Division, 9192.R18 SCR 2006-0013, Add Closed Safety Function For SI-850 Valves To PBNP Inservice Testing Program, dated 01/22/06.R19 PBNP-IC-02, RWST Level Instrument Uncertainty/Setpoint Calculation, Rev 2. Note: This calculation is on administrative hold. NP 7.2.4, Calculation Preparation, Review, and Approval (rcv. 15), requires that an evaluation be performed on the calculation to determine its effect prior to use. A review of the calculation with the PBNP Calculation Review Group found no significant discrepancies what would adversely affect the conclusions of EE 2006-0003.

J.O EVALUATION AND RESULTS .-PBNP does not perform a seat leakage test on the valves in the direction of the containment sump to the recirculation lines. However, based on the design of the valves and their operating conditions, the valves are expected to limit a passive pressure boundary failure, and prevent a gross diversion of water from the containment sump.While PBNP's licensing/design basis limits the leakage from a passive failure in the Residual Heat Removal System to less than 50 gpm (RI), the design of the SI-850 valves is expected to significantly reduce this leakage rate when they are closed. Steams and Rogers drawing 8551/4-sheet I (R.2) shows that the SI-850 valves are equipped with a resilient (soft) seat. Resilient seats are used to accomplish good seating performance with much lower contact force than is required in metal-to-metal seats. In the case of the SI-850 valves, the resilient scat is formed by an O-ring which provides the primary seating seal with the metal-to-metal closure acting as a secondary seal (R2). This type of valve scat design has very good seat leakage characteristics.

General Design Guidance on Seat Leakage for SI-850A/B Valves: The American National Standard for Control Valve Seat Leakage, ANSI B 16-104 (R3), provides guidance on allowable seat leakage for valves with an 0-ring seating design (Class VI) similar to the Sl-85OA/B valves. This table only includes valves with nominal sizes of S" or less. Based on the allowvable leakage rates (mil/min) provided in ANSI B16-104 and the port area of the valves it can be determined what the allow able leakage rate per square inch of port area is for the valve sizes listed in the ANSI Standard.

As is seen in the table belowv, the greatest allowvablc leakage rate on a (ml/minyin 2 of port size basis is approximately 0.191 (ml/min)/in 2.PBF-1649 Revision 0 11/129/00 Rcercnce:

NP 7.2.10 I --.Engineering Eval. No. 2006-0003 Point Beach Nuclear Plant Revision 0 ENGINEERING EVALUATION PagcNo. °f 7 Initials ALG Date 02/0612006 Table I -Allowable Seat Leakage Based on Nominal Port Diameter Source Nominal Port (in) Port Area (in 2) Allowable Leakage (mI/min) (mlhnin)/in 2 B116.104 1 0.785 0.15 0.191 B 16.104 1.5 1.767 0.3 0.170 B136.104 2 3.142 0.45 0.143 B 16.104 2.5 4.909 0.6 0.122 B116.104 3 7.069 0.9 0.127 B116.104 4 12.566 1.7 0.135 B116.104 6 28.274 4.0 0.141 B 16.104 8 50.265 6.75 0.134 Derived 10 78.540 15 0.191 By multiplying this 0.191 (mil/min)/in 2 value by the nominal port area of a 10 inch valve, the approximate allowable leankagc rate fora 10 inch valve can be derived. Per the above table, this allowable leakage rate is found tobe approximately 15 ml/min at maximum rated differcntial pressure.

Based on this, the S1-850 valves can be expected to limit leakage to a value significantly below 50 gpm when closed.Hydraulic Valve Closure: Based on the design' of the S1-850 valves, with the top of the valve disc being open to containment,'seat leakage testing from the containment sump to the containment sump recirculation lines is not possible.

PBNP informally checks the S1-850A/B scat tightness and O-ring condiiion each refueling outage under IT 531 for Unit I and IT 536 for Unit 2 (R4/5). During this test theleontainmeclt sump recirculation piping is pressurized to 75 to 80 psig with the SI-850 valves closed to determine system leakage. Any scat leak-age through the SI-850 valves is noted, although it is not quantified.

l he S1-850 valves are held in the closed position with a strong back during the performance of this testing as the pressure is under the valve scat and is acting to force the valve to open. Under accident conditions, if the valve were required to close, this opening force would be gone and the containment pressure and the head of the containment sump would actually assist in closing the valve and compressing the 0-ring to limit scat leakage.During each quarterly Inscrvice Test of the S1-850 valves, plant operators record the hydraulic pump's output pressure in both the open and closed direction stroke. During Inservice Testing of the SI-850A1B valves, the hydraulic pressure from the associated pumps is required to be greater than or equal to 1 150 psig and less than or equal to 1500 psig (R18). Based on (his, an operable Sl-850A/B valve will have at least 1 150 psig of hydraulic pressure acting on the hydraulic cylinder when the valve is closed. As is discussed in Calculation 2000-0018, Parameters for Testing S[-850A&B (R6), the effective hydraulic area of the hydraulic ram when closing the valve is 6.28 sq. in. Based on this, the SI-850 hydraulic valve operators would produce a closing force of approximately 7222 lb (1150 psig

• 6.28 sq. in = 7222 lb).The SI-850 valve disc 0-ring (Parker Size 2-450) is nominally 10.5"x1 l"xl/4" (R7) and is made of Ethylene Propylene (R2). Ethylene Prolylene 0-rings are capable of performing in high temperature and high radiation environments (R8); The actual thickness of a nominal 1/4" 0-ring is 0.275"(R9).

By observation it can be shown that the 0-ring is flattened such that its contact with the valve's seat is defined by an area no greater than that equal to the profile of the narrow portion of the 0-ring retention groove in the valve disc. After this, 0-ring crush is limited by metal-to-metal contact. Based on this, the maximum seating surface area of the 0-ring would be 7.94 sq. in. ([3.1416 * (10.988" / 2)2] -[3.1416 * (10.518" / 2)2j). Nominal inner (1 0.518") and outer (1 0.988") groove measurements associated with the valve disc drawing (R10) were used to calculate this area.Based on this, the hydraulic closure force of 7222 lbs would result in maximum initial 0-ring compressive seating pressure of up to 910 psi (7222 lbs /7.94 sq. in. = 910 psi); however, 0-ring crush would be limited by the design PBF-1649 Rc vision 0 1 1129i00 Rc froncc: NP 7.2.10 Engineering Eval. No. 2006-0003 Point Beach Nuclear Plant Revision 0 ENGINEERING EVALUATION PageNo. 6 of 7 Initials ALG Date 02/0612006 ofthe valve disc dovetailed O-ring retaining groove. This groove is 0.23 1 inches deep thus O-ring crush is limited to 0.044 inches (0.275" -0.231").

Aftcr this crush is achieved, the valve seats go metal-to-metal preventing O-ring over-crush.

This limits the O-ring crush or squeeze to 16 percent (100 (0.044"/0.275")).

The Parker O-ring Handbook (RI ) recommends that for a static O-ring seal, squeeze be maintained between a minimum of 0.007 inches to prevent excessive compressive set and 30 percent to prevent early seal deterioration.

The design of the SI-850 valves O-ring meets both of these recommended constraints.

Based on this, the design and operation of the SI-850A1B valves and their associated operators and hydraulic pumps provides assurance that when hydraulically closed they will isolate containment B sump flow into the associated containment recirculation line to prevent gross diversion.

Long Term Valve Closure: If an SI-850 valve were hydraulically closed, even if the hydraulic operator pressure -decayed away, there would be other forces holding the S1-850 valve closed and compressing the O-ring. Post accident, a S1-850 valve placed in the closed position wvould be exposed to a differential pressure across the valve disc. The differential pressure would be the pressure that the top of the valve disc is exposed to (by the head of the water in the containment B sump, the containment pressure, and the wveight of the valve disc and push rod) minus any pressure in the downstream piping. SI-850A/B valve closure is only relied upon to isolate a passive failure in the containment sump B recirculation line. Once closed it can be assumed that the downstream pressure would be vented through the pressure boundary failure caused by the passive failure and would decay away to atmospheric pressure.For the most limiting case (when the O-ring seals at the inner edge), the minimum diameter of the 0-ring groove is 10.503" (10.518"-0.015l where 0.015" isIhellowable tolerance) based on nominal diameter less altowable tolerance (RIO). It is conservative to assume that containment presstire acts only on the area ofthi valve disc inside this radius and that there is no differential pressure on the area of the disc outside of this radius because it minimizes the static closure forces on the valve. As a result, the surface area of the valve disc that is exposed to a differential pressure would be 86.6 in 2 (3.1416 * (I0.503" /2)2). Postaccident the containment sump is flooded with approximately 4.5 feet of water above the top of the SI-850 valves. This is based on the equation devel6ped in Calculation 2000-0044 (R12) Ls,,p= 62.83 -0.7 1LRAST -62.83 -0.71(12)=54.31"or4.53' based onRWST level being at 12 percent (9 percent measured RWST plts 3 percent for instnument inaccuracy (R19)) during switch over to containment sump recirculation (R13). PBNP would not expect to go on containment sump recirculation until 2252 seconds after the accident at which time containment pressure would be reduced to approximately 26 psig and sump temperature would have been reduced to approximately 209 'F (R14). Under these conditions, conservatively, each 2.41 ft of water would produce a I psi pressure on the valve disc (as containment sump would have pressurized water with a density greater than at atmospheric pressure).

Conversion of Water Density into Static Pressure Density of water at 210 'F is 59.86 lb/ft 3 (R15)59.86 (lbln 3) * (l ft/144in 2) = 0.4157 Ib/ (in 2

• ft); therefore, ftl/(0.4157)=

lb/in 2= psi Therefore, 2.41 fit of water produces I psig of pressure As a result, the containment sump head of 4.53 feet would produce a pressure of 1.9 psi (4.53 ft 1 2.41 ft/psi) and a resulting closing force of 165 lbs (1.9 psi

• 86.6 sq. in.). This force would produce a scating pressure of 20.8 psi (165 lb/7.94 sq. in.)on thecO-ring.

Thc ParkcrO-ring landbook (R16) FigurcA4-16 providesa method for correlating the comprcssivc load per linear inch of seal to percent 0-ring compression.

Using the center-line (RIO)of the O-ring grove it can be determined that the 0-ring's circumference is 33.8 in([I0.518"+

10.988"]1

/2=10.753" -2r*( 10.753/2)

= 33.8 in.). Based on this, the compression load per linear inch of seal for the O-ring PBF-I 649 Kevision0 I W)29.V0 Recrernce:

NP72.10 Point Beach Nuclear Plant ENGINEERING EVALUATION Engincering Eval. No.Revision Page No.2006-0003 0 7... or 7 .Initials ALG Date 02/0612006 would be approximately 5 lbs/in (165 lb / 33.8 in). Per the Parker 0-ring Handbook (R16) this results in a 5 to 10 percent compression orthe 0-ring (depending on the hardness of the 0-ring, Parker E5 15-80 Ethylene Propylene has a typical Shorc A hardness of 80 (RI 7)). This translates into a squeeze or compression of at least 0.0137" (0.275"

• 0.05) which maintains the recommended minimum squeeze of 0.007". It should be noted that the weight of the valve disc and stem along with any containment pressure would also provide additional closing force and 0-ring squeeze above what is calculated above. Based on this, once closed to isolate a passive leak in a containment sump B recirculation line, the SI-850 valves would be expected to remain closed and prevent gross leakage.

6.0 CONCLUSION

S AND RECOMMENDATIONS The Sl-85OA/B valves are capable of hydraulic closure if required post accident.

When closed, they will prevent gross diversion of Containment Sump B water to a passive failure in the associated Containmcnt Sump B recirculation line both in the short and long term. This is based on the closure force and 0-ring compression force being maintained within manufacturer's recommendations when the SI-850 valves are placed in the closed position.PBF-1649 Revision0 11129100 Rcdrcnce:

NP72.10