RNP-RA/03-0034, Revision to Inservice Testing Program Relief Request IST-RR-3 for Containment Spray Pump Comprehensive Pump Test Requirements

Revision to Inservice Testing Program Relief Request IST-RR-3 for Containment Spray Pump Comprehensive Pump Test Requirements

ML031140369
Person / Time
Site:	Robinson
Issue date:	04/15/2003
From:	Baucom C Progress Energy Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RNP-RA/03-0034
Download: ML031140369 (32)

Text

10 CFR 5fl55a C; Progress Energy Serial: RNP-RA/03-0034 APR 1 5 2003 United States Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261/LICENSE NO. DPR-23 REVISION TO INSERVICE TESTING PROGRAM RELIEF REQUEST IST-RR-3 FOR CONTAINMENT SPRAY PUMP COMPREHENSIVE PUMP TEST REQUIREMENTS Ladies and Gentlemen:

By letter dated August 24, 2001, and pursuant to 10 CFR 50.55a(f)(5)(i), Carolina Power and Light Company (now doing business as Progress Energy Carolinas, Inc.) submitted the "Inservice Testing Program Plan - Fourth Interval" for the H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2.

The HBRSEP, Unit No. 2, Fourth Ten-Year Interval began on February 19, 2002.

A request for relief from the comprehensive pump test requirements of the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code), 1995 Edition with 1996 Addenda, for the HBRSEP, Unit No. 2, Containment Spray pumps was included in the August 24, 2001 submittal. The proposed relief request was identified as Relief Request IST-RR-3. The NRC review of the proposed relief request, as provided in a letter dated June 27, 2002 (TAC No. MB2798), states that pursuant to 10 CFR 50.55a(f)(6), the alternative to perform a reduced flow, more comprehensive quarterly test, is authorized for an interim period until the end of Refueling Outage (RO)-22 on the bases that the Code-required test is impractical to perform without significant plant modification, the interim alternative otherwise meets the criteria of 10 CFR 50.55a(f)(6)(i), and that the interim relief will allow time for the licensee to explore other alternatives, make the necessary plant modification for performing the required test, or submit a revised relief request.

Revision 1 to Relief Request IST-RR-3 is provided as Attachment I to this letter. The revised relief request incorporates guidance provided by NRC staff personnel at the Seventh NRC/ASME Symposium on Valve and Pump Testing, held in Washington, DC, during July 15 through 18, 2002, and documented in NUREG/CP-0152.

Progress Energy Carolinas, In;.

3581 West Entrance Road Hartsville, SC29550

United States Nuclear Regulatory Commission Serial: RNP-RA/03-0034 Page 2 of 2 In order to support implementation of the Fourth Ten-Year Interval Inservice Testing Program, and to facilitate preparations for RO-22, it is requested that NRC complete the review of this revised relief request by August 15, 2003.

If you have any questions regarding this matter, please contact me.

Sincerely, C. T. Baucom Supervisor - Licensing/Regulatory Programs CAC/cac Attachments:

I. 10 CFR 50.55a Relief Request Number IST-RR-3, Revision 1 II. Containment Spray System Description c: Mr. L. A. Reyes, NRC, Region II Mr. C. P. Patel, NRC, NRR NRC Resident Inspectors

United States Nuclear Regulatory Commission Attachment I to Serial: RNP-RA/03-0034 Page 1 of 4 CAROLINA POWER AND LIGHT COMPANY (PROGRESS ENERGY CAROLINAS, INC.)

H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 10 CFR 50.55a RELIEF REQUEST NUMBER IST-RR-3, REVISION 1 RELIEF REQUESTED IN ACCORDANCE WITH 10 CFR 50.55a(f)(5)(iii)

--INSERVICE TESTING IMPRACTICALITY--

United States Nuclear Regulatory Commission Attachment I to Serial: RNP-RA/03-0034 Page 2 of 4

1. ASME Code Components Affected Containment Spray Pump A Containment Spray Pump B These pumps are single suction, single stage centrifugal pumps fabricated by the Ingersoll-Rand Corporation, Cameron Pump Division, in 1969. A system description of the Containment Spray (CS) system is provided in Attachment II. These pumps are classified as Group B pumps.

2. Applicable Code Edition and Addenda

ASME Code 1995, Code for Operation and Maintenance of Nuclear Power Plants, OM Part 6, 1996 Addenda.

3. Applicable Code Requirement

The ASME OM Code, 1995 Edition/1996 Addenda, ISTB 4.3, Reference Values, in paragraph (e)(l), requires reference values shall be established within 20% (80% - 120%) of the design flow rate for the Comprehensive Pump Test (CPT). The frequency for comprehensive testing of Group B pumps is biennially, in accordance with Table ISTB 5.1-1.

Based on the performance curve provided by the manufacturer (example performance curves for these pumps are provided in Attachment II), the design flow rate for these pumps is approximately 1200 gpm at 475 feet of total developed head. Therefore, the Code requirement implies that for the CPT the flowrate of these pumps should be in the range of 960 gpm to 1440 gpm. Based on the installed system configuration, originally specified design requirements, and accident analysis minimum flow requirements, the design flow of the system has been determined to be 970 gpm, which results in a Code-required flow test band of 776 gpm to 1164 gpm.

4. Impracticality of Compliance H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2, requests relief from the requirements of ISTB 4.3(e)(1), such that the CPT reference values may be established with the pump operating at less than 80% of the design flow rate.

The CS pumps are tested using a test line that circulates back to the Refueling Water Storage Tank (see Attachment II for a CS system flow diagram). This test line produces a flow rate of approximately 25% of the design flow (i.e., approximately 240 gpm). Additional flow is also circulated from the pump discharge to the pump suction through an eductor. The flow rate through this part of the system is estimated at 80 gpm. Therefore, the total flow rate through the pump is approximately 320 gpm, which corresponds to approximately 33% of the pump design flow rate.

An additional flow path is available, but this flow path uses the same piping that is already being used, therefore, it is expected that the additional flow achieved would be minimal. Use of this

United States Nuclear Regulatory Commission Attachment I to Serial: RNP-RA/03-0034 Page 3 of 4 test flow path would also require manipulation of additional valves that are normally locked in position to maintain operability of the system, which introduces the possibility that these valves could be mispositioned during plant operation.

The only CS system flow path available that can produce the required flow rate would spray containment with borated water. This would require an extensive cleanup and would be detrimental to carbon steel material and non-qualified electrical circuits. Therefore, this method is not acceptable.

A similar relief request had not been submitted in the previous inservice testing intervals because the ASME and OM Code versions applicable during the preceding intervals did not require the pump performance reference to be based on design flow rate conditions. The Statements of Consideration associated with the changes to 10 CFR 50.55a that incorporated OM Code 1995 Edition/1996 Addenda stated the following, "A utility (commenter three) stated that changes to the OM Code that appear in the 1995 Edition with the 1996 Addenda would require their facilities to modify the test loop piping for demonstrating pump design flow rate.

The NRC is aware that some licensees may have difficulty fully implementing these tests and in certain cases, due to the impracticality of implementation, a request for relief under Sec.

50.55a(f)(5) would be appropriate. However, the OM committees developed these provisions in an effort to improve functional testing of pumps because present pump testing programs may not be capable of fully demonstrating that pumps are performing as designed. Some licensees have preoperational test loops which may be used to demonstrate full flow for this testing. Hence, the NRC has concluded that current regulatory requirements address this issue and a modification to the final rule in response to this comment is not required."

5. Burden Caused by Compliance To establish the capability to test at the Code-required flow rate would require a substantial plant modification and re-design of the CS system. The cost of a permanent modification and associated activities has been estimated at $220,000.

6. Proposed Alternative and Basis for Use These pumps are classified as Group B pumps, and are subject to quarterly inservice tests where differential pressure and flow rate are required to be monitored.

As an alternative to the Code-required CPT, the quarterly inservice tests will monitor the following parameters: differential pressure, [test line] flow rate, and vibration.

The vibration data will be collected and analyzed as part of each quarterly test in accordance with OM Code requirements for comprehensive pump testing. The vibration acceptance criteria will be limited as specified in Table ISTB 5.2.1-1 of the OM Code, 1995 Edition/1996 Addenda.

Approximately 10 years of vibration data is provided in Attachment H.

Additionally, the quarterly inservice tests will utilize a more accurate discharge pressure instrument than currently installed and used (2 psi accuracy verses 6 psi accuracy). The

United States Nuclear Regulatory Commission Attachment I to Serial: RNP-RA/03-0034 Page 4 of 4 increased accuracy will increase the likelihood that actual pump performance degradation will be detected. More accurate flow rate measurement is not considered necessary, because the tests utilize a fixed resistance system and there are two flow paths that would need to be measured, one of which is not currently instrumented for measurement of flow rate.

Additional condition monitoring of the pumps will be conducted, which includes oil sampling and analysis, and spectral analysis of vibration. This condition monitoring will be performed annually. Approximately 7 years of oil sampling data is provided in Attachment II.

The CPT and the quarterly inservice tests will be conducted with the same system configuration.

The proposed alternative provides an acceptable level of quality and safety, because the CS pumps are standby pumps (not normally operating). These pumps are seldom used and they are run only for testing. The total run time is estimated at less than two (2) hours per year. The proposed alternative to the Code-required flow test will continue to provide an acceptable level of quality and safety, because pump degradation is expected to be minimal due to low run hours, the pumps are not subjected to events that could cause degradation (e.g., cavitation and runout),

and there is no observed degradation (such as corrosion or erosion) in the pumps or in the CS system. As stated in the Attachment UI, the materials of construction for the CS system and CS pumps are stainless steel or equivalent corrosion-resistant material.

As stated previously, the plant design did not provide for a "design flow test" or even a "near design flow test." The pump condition monitoring provided by the quarterly measurement of vibration and annual oil sampling and analysis, coupled with the other available pump parameters (flow and pressure), will allow pump degradation to be identified and corrected.

Therefore, testing at higher flow rates is not necessary.

Based on the information presented above, there is reasonable assurance that operational readiness of the CS pumps is maintained. The proposed alternative provides an acceptable level of quality and safety. Additionally, compliance with the Code requirement would result in an unusual hardship without a compensating increase in the level of quality and safety.

7. Duration of the Proposed Alternative If approved, the duration of the proposed relief would be the remainder of the Fourth Ten-Year Inservice Inspection Interval, which ends on February 18, 2012.

United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/03-0034 Page 1 of 26 CAROLINA POWER AND LIGHT COMPANY (PROGRESS ENERGY CAROLINAS, INC.)

H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 INSERVICE TESTING PROGRAM PLAN - FOURTH INTERVAL CONTAINMENT SPRAY SYSTEM DESCRIPTION

United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/03-0034 Page 2 of 26 The H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2, Updated Final Safety Analysis Report (UFSAR), describes the design basis for the Containment Spray system as follows:

The primary purpose of the Containment Spray (CS) system is to spray cool water into the containment atmosphere, when appropriate, in the event of a LOCA and thereby ensure that containment pressure does not exceed its design value which is 42 psig at 263TF (100 percent relative humidity). This protection is afforded for all pipe break sizes up to and including the hypothetical instantaneous circumferential rupture of a reactor coolant pipe. Pressure and temperature transients for LOCA are presented in the HBRSEP, Unit No. 2, UFSAR, Section 6.2.1.1.1.1. Although the water in the core after a LOCA is quickly subcooled by the Safety Injection system, the CS system design is based on the conservative assumption that the core residual heat is released to the containment as steam.

The CS system was designed to spray at least 2322 gpm of borated water into the Containment Building whenever the coincidence of two sets of two out of three (Hi-Hi) containment pressure signals occurs or a manual signal is given. Either of two subsystems containing a pump and associated valving and spray headers is independently capable of delivering one-half of this flow, or 1161 gpm.

The design basis was to provide sufficient heat removal capability to maintain the post-accident containment pressure below the design pressure, assuming that the core residual heat is released to the containment as steam. This requires a heat removal capacity of the subsystem, with either pump operating, at least equivalent to two fan-coolers heat removal capability at the containment design conditions. A second purpose served by the CS system (CSS) is to remove elemental iodine from the containment atmosphere should it be released in the event of a LOCA (refer to UFSAR Section 6.5.2). The analysis showing the system's ability to limit offsite thyroid dose to within 10 CFR 100 limits after a hypothetical LOCA is presented in Section 15.

The spray system was designed to operate over an extended time period, following a primary coolant system failure, as required, to restore and maintain containment conditions at near atmospheric pressure. It has the capability of reducing the containment post-accident pressure and consequent containment leakage, taking into account any reduction due to single failures of active components.

Portions of other systems which share functions and become part of the containment cooling system, when required, are designed to meet the criteria of this section. Any single failure of active components in such systems does not degrade the heat removal capability of containment cooling.

Those portions of the spray system located outside of the containment which are designed to circulate, under post-accident conditions, radioactively contaminated water collected in the containment meet the following requirements:

a) Adequate shielding to maintain radiation levels within the guidelines of 10 CFR 100 b) Collection of discharges from pressure relieving devices into closed systems, and c) Means to limit radioactivity leakage to the environs, consistent with guidelines set forth in 10 CFR 100.

United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/03-0034 Page 3 of 26 System active components are redundant. System piping located within the containment is redundant and separable in arrangement unless fully protected from damage that may follow any primary coolant system failure. System isolation valves relied upon to operate for containment cooling are redundant, with automatic actuation or manual actuation. All portions of the system located within containment were designed to withstand, without loss of functional performance, the post-accident containment environment and operate without benefit of maintenance for the duration of time to restore and maintain containment conditions at near atmospheric pressure.

Adequate containment cooling and iodine removal are provided by the CSS shown in UFSAR Figure 6.2.2-1, whose components operate in sequential modes. These modes are:

a) Spray from the Refueling Water Storage Tank into the entire containment atmosphere using the containment spray pumps. During this mode, the contents of the spray additive tank (sodium hydroxide) are mixed into the spray stream to provide adequate iodine removal from the containment atmosphere by a washing action.

b) Recirculation of water from the containment sump is provided by the diversion of a portion of the recirculation flow from the discharge of the residual heat removal (RHR) heat exchangers to the suction of the spray pumps after injection from the refueling water storage tank has been terminated.

The principal components of the CSS which provides containment cooling and iodine removal following a LOCA consist of two pumps, one spray additive tank, spray ring headers and nozzles, and the necessary piping and valves. The containment spray pumps and the spray additive tank are located in the Auxiliary Building and the spray pumps take suction directly from the refueling water storage tank.

Permanent test lines for all containment spray loops were located so that the system, up to the isolation valves at the spray header, can be tested. These isolation valves can be checked separately.

The containment spray pumps were designed in accordance with the specifications discussed in UFSAR Section 6.1.1.1.1 for the pumps in the SI system. The materials of construction are stainless steel or equivalent corrosion-resistant material.

UFSAR Section 6.1.1.1.1.1 states:

The pressure-containing parts of the pumps were constructed of castings which conformed to American Society for Testing and Materials (ASTM) A-351 Grade CF8 or CF8M specifications. Stainless steel forgings were procured per ASTM A-182 Grade F304 or F316 or ASTM A-336, Class F8 or F8M, and stainless plate was constructed to ASTM A-240 Type 304 or 316 specifications. All bolting material conformed to ASTM A-193. Materials such as weld-deposited Stellite or Colmonoy were used at points of close running clearances in the pumps to prevent galling and to assure continued performance capability in high velocity areas subject to erosion.

United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/03-0034 Page 4 of 26 All pressure-containing parts of the pumps were chemically and physically analyzed and the results checked to ensure conformance with the applicable ASTM specification. In addition, all pressure-containing parts of the pump were liquid penetrant inspected in accordance with Appendix VIII of Section VIII of the ASME Boiler and Pressure Vessel (B&PV) Code. The acceptance standard for the liquid penetrant test was USAS B31.1, Code for Pressure Piping, Case N-10.

Where welding of pressure-containing parts was necessary, a welding procedure including joint detail was submitted for review and approval by Westinghouse.

The procedure included evidence of qualification necessary for compliance with Section IX of the ASME Code, Welding Qualifications. This requirement also applied to any repair welding performed on pressure-containing parts.

The pressure-containing parts of the pump were assembled and hydrostatically tested to 1.5 times the design pressure for 30 minutes.

Each pump was given a complete shop performance test in accordance with Hydraulic Institute Standards. The pumps were run at design flow and head, shut-off head, and three additional points to verify performance characteristics.

Where NPSH is critical, this value was established at design flow by means of adjusting suction pressure.

UFSAR Section 6.3.2.2.3, Net Positive Suction Head (NPSH) Requirements, states the following:

During the safety injection phase, the worst case conditions for determining NPSH requirements occur with the single failure of a high head pump resulting in the following:

1. 1 high head pump at 661 gpm,

2. 2 low head pumps at 4572 gpm total, and

3. 2 containment spray pumps at 2467 gpm total.

A quantitative analysis of the available and required NPSH for the SI, RHR, and containment spray pumps for both the safety injection phase (with suction from RWST) and the recirculation phase (suction from the containment sump) shows:

1. During the safety injection phase with suction from the RWST (at the RWST low level setpoint), operating as described above, the following applies:

Pump Required NPSH. ft Available. ft High head, 1 pump (most limiting) 32 35.8 Low head (RHR), 2 pumps 10 Approx. 54 Containment spray, 2 pumps 20 Approx. 35

United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/03-0034 Page 5 of 26 From this it can be seen that the high head pump is the controlling component for NPSH. The safety injection phase will be terminated just before the RWST level decreases to the point at which the available NPSH is reduced to the required NPSH of 32 ft at the runout flow of 661 gpm. Transition to recirculation from the containment sump will commence at this point.

2. During the transition to the recirculation phase, conditions are such that one high head pump and one containment spray pump are operating.

During this period, the worst case NPSH conditions occur at the RWST low-low level setpoint as follows:

Pump Required NPSH. ft Available. ft 1 high head pump at 662 gpm 32 35.7 1 containment spray pump at 1220 gpm 20 Approx. 35

3. During the recirculation phase (from containment sump) the following applies:

a. High head SI pumps - During recirculation via the high head pump, this pump and the RHR pump would be aligned in series, with the RHR pump (which has a design discharge head of 240 ft) boosting the suction of the high head pump. Thus, no NPSH problems would be experienced.

b. Containment spray pump - Same as high head SI pump.

c. RHR (low head) pump - During recirculation from the containment, at 3833 gpm, the minimum available NPSH with 1.5 ft of water on the containment floor, is 19 ft. The required NPSH at 3833 gpm is 15 ft.

The high head recirculation flow path via the high head SI pumps is only required for the range of small break sizes for which the RCS pressure remains in excess of the shut-off head of the RHR pumps at the end of the safety injection phase.

United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/03-0034 Page 6 of 26 Pump ID: CS&A I -7 17-7 T FTT

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United States Nuclear Regulatory Commission Attachment It to Serial: RNP-RA/03-0034 Page 13 of 26 ATTACHMENT 8.3 Page 1of 1 CONTAINMENT SPRAY PUMP VIBRATION DATA POINTS 4 HORIZONTAL IOST352-1 Rev. 16 Page 20 of 22

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CP&L Laboratory Lube Oil Report Plant: HBR SamplePoint: HBRI14 Descrlption: ACONTAMENTSPRAYPMPBRORESV Particle Particle Particle Particle Particle ConnW Counts Countst Conzsf CountsC Vsosiyt Color Sample Dalc 100ls 100 m 1s lals IS I00f 100 mls Mooiture cSt NurlityASM Al Cu Ca Fc Pb Mg P Sn Cr Zn Si Number Sampled 5-10 10.25 25-SO 50-IOD >I100 Wg @40C mgKOWg D-15W ppm ppm ppm Ppm ppm ppm ppm ppm pprm ppm ppm 300439 01121R003 874000 283 13500 1610 144 17.0 152.0 0.0O 3.0 d3 c2 25 4 '3 425 '3 <1 <25 c1 Comments:

201012 02t1912002 549000 286000 40500 4740 108 12.0 152.0 0.08 3.5 d3 42 45c C2 '3 425 '3 <1 C25 <I Comientris 101430 0321/2001 .- 24.0 153.0 0.11 3.5 c3 c2 45 42 3 <25 425 '3 <1 <45 <I Comments:

1I99 OY1912040- 5.0 153.0 0.08 3.5 c3 c2 25 42 ' 345 45 '3 'I c25 <I Comments:

901047 02123/1999 - - - 290 151.0 01I 3.5 d3 42 45 c2 ' 345 25 '3 <1 a5 <I Comments:

SD6459 12M13M1998 - - . - 36.0 1520 0.10 3.5 d3 c2 c25 c2 ' 325 425 '3 <1 45 cI Counenu:

9:04923 09/l~fl198 - 40.0 151.0 0.11 3.5 c3 4 c25 - 'd 25 425 '3 <1 cZ5 <I commnents:

9803385 06/19/1Y9 - - - 190.0 151.0 0.08 3.3 c3 C2 425 3 c3 C25 c25 <3 'I 45 <I Comments:

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9706847 WItl0f1997 - - - 40.0 1510 0.11 d3 42 C25 42 '3 25 425 '3 <1 C25 cI Comments: 3.5 970519 08121/1997 . . . 58.0 151.0 0.08 3.5 c3 425 42 42 ' c25 C25 c3 <1 425 cl Comments:

970314 05/29/1997 - 53.0 30 151.0 0.11 3.0 c3 2 c45 c2 3 45 4C25 <1 45 <I Commen:

9702222 0 104/997 - - 29.0 150.0 0.11 3.5 c3 42 425 C2 ' 325 c25 d <1 45 <I commerims 97D1552 0310 997 - . . 45.0 148.0 .11 3.0 '3 42 c25 c2 ' C25 45 cd <1 45 <1 Comments:

9701074 O2V17/1997 * - - 1600.O 151.0 0.11 3.5 '3 4 C54 2 '3 45 45 d '1 45 di Commentw 960S6590103/1996 - - 69.0 146.0 0.10 3.0 '3 44 C25 4 ' c25 45 d <1 45 <I Comments:

9601973 OU18U1996 - - 50.0 149.0 0.10 3.0 d3 2 C45 4 3 <25 425 d <1 45 c1 Comments

United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/03-0034 Page 25 of 26 CP&L Laboratory Lube Oil Report I'Rant: HBR SamplePoeit: HBRIl5 Descriptln: B CONTAMENT SPRAY PMP BRO RESV Patidce Pafide Paride Padcle Patide Crs Cmts cot CoCns counts/ Vicosity Color Safle Doe lOmls 300 mis lOOn *ao tis 100 100 mls Moistue cSt Neutrlity ASTM Al Cu Ca Fe Pb Mg P Sn Cr Zn Si Number mSaWd 5-10 10-25 25-50 50-100 >100 uDlg G40C mgKOI/g D-1500 ppm ppm ppm ppm pm ppm ppm ppm pm ppm Pm 301s11 04S/0 2500000 938000 21400 744 16 15.0 152.0 0.os 3.5 '3 <25 <2 3 .- 5 '3 c1 25 <1 Comments:

102382 05/CYZ1 . - - 31.0 153.0 0.31 3.5 '3 45 <2 ' 345 45 '3 <1 <25 c1 Comment:

102269 051113001 23.0 152.0 0.08 3.5 '3 5 <25 3 ' 345 d5 d3 -I <2 <I Comments:

2674 05130J2000 - 18.0 155.0 0.08 3.5 d3 <5 2 <3 45 <25 '3 <1 4<5 2 Cooments:

9903020 06FI9/1999 - - 33.0 153.0 0.11 3.5 d3 43 <25 3 34 <5 <25 '3 <1 25 2 Comments:

• 901623 0418/999 . - 40.0 152.0 0.08 3.5 d3 2 d5 42 3 45 <25 '3 'I <25 <1 Coamtrnts:

90511 0111211999 - - 60.0 152.0 0.11 40 d3 3 <25 4 '3 <25 <25 '3 <1 <45 <I Coimments:

9505733 0 I/t998S - 33.0 152.0 0.1Q 3.5 d3 5 45 4d '3 45 45 '3 <25 -I Comments: 'd 9804311 0712S/199S . -

• 53.0 152.0 0.08 3.5 d3 3 <25 4 ' 345 d5 '3 <25 -cI Comments: <1 9802538 05W61998 - - 25.0 151.0 0.11 3.0 d3 5 <25 2 '3 45 <25 '3 <25 -I Comments: <1 980049 02110/199 - - 21.0 151.0 0.11 3.5 d3 8 45 4 3 <25 <25 '3 45 <I Comments: <1 9706921 11121/917 -

• 39.0 1510 0.11 3.5 '3 3 45 2 3 4545 '3 a5s <I

'I 9M05t2 08M12711 Comments:

• 85.0 1510 0.11 3.5 '3 <25 <2 ' 345 <25 '3 45 <I

<1 9703526 06/03/997 - - - 47.0 152.0 0.11 3.0 d3 C2 45 3 ' 35 245 '3 <25 <I

<1 9702223 04104/1997CDMMU: - - 58.0 149.0 0.08 3.5 d3 <2 45 4 ' 345 45 '3 <25 -i Comments: 4

.4 9701614 03/1 /199 64.0 151.0 0.10 3.5 d3 425 2 '3 <25 45 '3 (1 45 d1 c2 9701075 02t171l997 . 43.0 150.0 0.08 3.5 d3 'c5o 4 ' 345 425 d3 <1 <25 -<

Comments:

960566 0913/IYY6 73.0 148.0 0.10 3.0 '3 42 45 2 '3 <25 <25 3 -I <25 <I Comments:

I

United States Nuclear Regulatory Commission Attachment It to Serial: RNP-RA/03-0034 Page 26 of 26

• Revio No. 15 RM.Dwq. CF00PU4379O82 K S. ROSINSON CWolha oumr & UWht Convay I UPsA f FINAL SATY Ar ALYIS REPORT I

FLOW DIAGRAM CONTAINMENT SPRAY FiGURE * -I