ML20209A943

Confirms Adequacy of Safety,Relief & Block Valves Based on Review of EPRI Generic Test Program Results,Pipe Line & Support Evaluation & Analysis & Equipment Environ Qualification Repts,Per NUREG-0737,Item II.D.1

ML20209A943
Person / Time
Site:	Beaver Valley
Issue date:	04/10/1987
From:	Carey J DUQUESNE LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM 2NRC-7-073, 2NRC-7-73, TAC-62894, TAC-62920, NUDOCS 8704280320
Download: ML20209A943 (10)

Text

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'Af M 2NRC-7-073 Bewer Va o. 2 Unit Project organization Telecopy ( 3 Ext.160 P.O. Box 328 April 10, 1987 Shippingport, PA 15077 United States Nuclear Regulatory Commission ATTN: Document Control Washington, DC 20555

SUBJECT:

Beaver Valley Power Station - Unit No. 2 Docket No. 50-412 NUREG 0737, Item II.D.1 Safety / Relief Valves (SER Confirmatory iten #11)

Gentlemen:

In accordance with the requirenents of NtREG 0737, Iten ll.D.1, this con-finns the adequacy of the safety, relief and block valves based on our review of the EPRI generic test progran results, pipe line and supports evaluation and analysis, and equipment environnental qualification reports. Duquesne Light Company has actively pursued resolution of these issues through direct partici-pation in the Electric Power Research Institute (EPRI) valve test progran.

The primary objective of the test progran was to provide full-scale test d ata confirming the functional capability of pr imary systen power operated relief valves (PORVs) and safety valves for expected operating and accident conditions. The second objective was to obtain sufficient piping thermal hydraulic load data to permit confirmation of models which may be utilized for plant unique analysis of safety and relief valve discharge piping systens. The relief and safety valve tests were conpleted in August 1981 and Decenber 1981, respectively.

Valves and piping configuraton similar to those at BVPS-2 have been tested within the EPRI safety and relief valve test progran. This progran was conduc-ted in response to NUREG 0737, Iten ll.D.1 (USNRC 1980) . The results of the EPRI test progran denonstrate the acceptability of the BVPS-2 design.

The Crosby safety valves (6M16) and Garrett (Crosby) PORVs (3 in. x 6 in.)

currently used at BVPS-2 are enveloped by those in the EPRI test. In addition, the fluid conditions and valve opening times used in calculating the flow tran-sient loads for the BVPS-2 wer e derived fran the EPRI test. In addi tion, the fluid conditions and valve opening times used in calculating the flow transient loads for BVPS-2 were derived from the EPR1 test progran. The EPRI test condi-tions envelope the BVPS-2 plant-specific design conditions for both anticipated operational occurrences and accident conditions, and the piping and pipe supports are designed to wi thst and the resul ting calcul ated loads in 870428A 0 12 k PDR PDR ,

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United States Nuclear Regulatory Commission Mr. Peter Tam, Project Manager Safety / Relief Valves Page 2 accordance with the applicable code requirements. The differences between the EPRI generic test loop piping and the BVPS-2 piping are accounted for by analyses using Stone and Webster Engineering Corporation (SWEC) computer pro-grams which have been bench 11arked against the EPRI test results as described in Appendix 3A to the BVPS-2 Final Safety Analyses Report. The above analyses, supported by EPRI test results, demonstrate the functionability of the valves and the piping system's ability to withstand the imposed loadings from expected flow conditions.

The format of the rest of this letter is divided into the following sec-tions:

Test Program Reports - Safety and Relief Valves Test Program Reports - Block Valves Analysis Mechanical and Electrical Environmental Qualification - Safety, Relief and Block Valves Pertinent FSAR Information Attachments Test Program Reports - Safety and Relief Valves The reports listed below were reviewed and it was concluded that the valves adequately represented the safety and relief valve designs and the con-ditions for BVPS-2. These renorts were utilized to perform the final pl ant specific evaluation. {

The following reports, ex$3pt for 2(b) and 6 are included with this submittal in Attachment A. j

1. "EPRI PWR Safety a$d Relief Valve Test Program Valve Selection /Se-lection/Justificat wn Report", LPRI NP-2292, Final Repo r t, December 1982, documenting that the selected valves represent al l participating PWR plant safety and relief valves.

2. (a) "EPRI PWR Safety and Relief Valve Test Program Test Condition Justification Report", EPRI NP-2460-SR, Special Repor t, December 1982 and (b) " Westinghouse Plant Condition Justification Report" documenting the basis and Justificaton of the valve test conditions '

fqr alI participating PWR plants.

4 i !3. "E?RI PWR Safety and Relief Valve Test Program Safety and Relief p Valve Test Report", EPKI NP-2628-SR, Special Report, December 1982 which provided evidence demonstrating the functional capability of the selected test valves under the selected test conditions for all participating PWR plants.

United States Nuclear Regulatory Commission Mr. Peter Tam, Project Manager Safety / Relief Valves Page 3

4. " Application of RELAP S/ MOD 1 for Calculation of Safety and Relief Valve Discharge Piping Hydrodynamic Loads", EPRI NP-2479, Final Report, December 1982, which presents an analytical model bench-marked against test data that may be used for plant unique analysis of safety and relief valve discharge piping systems.

5. " Valve Inlet Fluid Conditions for Pressurizer Safety and Relief Valves in Westingnouse-Designed P lants", LPHI NP-2296, Final Report, December 1982.

6. "EPRl/Wyle Power-0perated Relief Valve Phase III Test Report Volume 11: Summary of Pnase 111 Iesting of the Garrett Relief Valve",

EPRI NP-2b70-LU, October 1982.

Reference 2 above identified BVPS-2 as not being included for Cold Over-pressurization Events. We have reviewed the test data for valve operation for the Cold Overpressure Protection Systen and have determined that the BVPS-2 systen operates within this test envelope and is therefore represented by the referenced reports.

In addition to the above EPRI reports, the following reports previously submitted by DLC for the BVPS-1 response to NUREG 0737, item II.D.1 are appli-cable to BVPS-2.

1. " Safety Valve Contingency Analysis in Support of the EPRI S/RV Test Program - Volume 3: Westinghouse Systems", EPRI NP-2047-LD, Octo-ber 1981,

2. "EPRI PWR Safety and Relief Valve Test Program Guide for Applica-tion of Valve Test Program Results to Plant-Specific Evaluations",

Revision 2, July 1982.

3. "EPRI-Westinghouse Plant Conditions Justification Report".

Test Program Reports - Block Valves The plant-specific submittal for block valve qualification is represented by several reports. The following reports are enclosed in Attachment B.

1. "EPRI PWR Safety and Relief Valve Test Program PORV Block Valve Information Package", May 1982, this package contains a description of block valves in PWR service including valve types, installation arrr angenents and listing by PWR plants. As attachnents to the information package the following reports are included:

a. "EPRI/ Marshall Electric Motor-Operated Valve (Block Valve),

interim Test Data Repoert," May 31, 1982

b. "EPRI Summary Report: Westinghouse Gate Valve Closure Testing Program", March 31, 1982.

n United States Nuclear Regulatory Commission Mr. Peter Tam, Project Manager Safety / Relief Valves Page 4 The BVPS-2 block valves are represented by the above reports and the EPRI

" Test Condition Justification Report" and the Westinghouse " Plant Condition Justification Report."

The block valve tested at the Marshall facility, 3GM88FNH represents the block valves installed at BVPS-2. The enclosed block valve data package satis-fies NUREG-0737, Item II.D.1.B.

Analysis The pressurizer safety and relief valve piping is made up of four 6 inch Schedule 160 stainless steel piping connections to the top of the presurizer.

Each of three of these penetraticns provide a flow path to one of three spring actuated 6M16 Crosby safety valves. This upstrean piping is routed to provide a stean condensing loop seal which is insulated in a common cavity with the pressurizer vessel in order to maintain the fluid temperature near saturation.

The safety valves are supported at their inlet flanges through a welded plate assembly back to the body of the pressurizer vessel. The fourth pressurizer penetration is piped into three 3 inch piping branches each provided with a motor-operated " block valve" and so lenoid actuated - pilot-operated relief valve in series. This piping is also routed to allow condensing stean to fill a loop seal. Elevated liquid temperatures in this case are maintained by elec-trical heat tracing. The downstrean piping from each of the safety and relief valves is 6 inches and although the design pressure is significantly reduced, Schedule 160 stainless steel is used to provide maximum strength. Each of these branch pipes discharge to a single 12 inch Schedule 160 downcomer which discharges to the pressurizer relief tank on the containment floor.

The upstrean piping is ASME 111 Class 1 and subject to the rules of ASME 111 NB3600. This inicudes all normal, upset, emergency and faulted loading conditions and anong other routine analyses, requires detailed fluid dynanic time histories and demonstrations of satisfactory fatigue life. The downstream piping would normally be governed by the more simple rules of ANSI B31.1.

However, since this piping is in close proximity with the reactor coolant sys-ten code safety valves, structural isolaton between the two piping classes must be assured. Therefore DLC has elected to upgrade the safety class analyses to ASME 111 Class 3 where the effects of seismic acceleration of fluid dynanic loading are considered in a consistent manner with the upstrean piping. This results in one computer mathematical model which includes three safety valves and three relief valves and all piping between the pressurizer and relief tank.

One distinct advantage of this combined model is that the overlapping or compounding effects of multiple relief valve, safety valve, or seismic distur-bance can be accounted for easily.

Attachment C provides a list, in descending order, of ten design documents which substantiate the BVPS-2 presurizer safety and relief valve piping design.

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- United States Nuclear Regulatory Commission Mr. Peter Tam, Project Manager Safety / Relief Valves l Page 5 Water filled loop seal: Natur al condensation of stean in the piping upstrean of eacn valve forms a water loop seal which helps to prevent leakage, by insu-lating undissolved gases from the valve seat. This water tends to produce very large piping loads as it is accelerated through the pipe. EPRI demonstrated this phenomenon and showed that the closer the water temperature was to satura-tion temperature the less the resulting load. This is apparently due to higher percentages of water flashing to stean downstrean of the valves. SWEC developed and benchmarked a computer program "WAT-SLUG" which accurately calcu-lates the results of this water slug discharge phenomenon. "WAT-SLUG" was employed in the analysis of BVPS-2 pressurizer safety and relief valve piping.

Final development of the loop seal requirements is illustrated via the corres-pondence listed in Attachment D.

Discharge loads on downstream piping: In spite of the f act that the loop seal water seal tenperatur e is maintainea to minimize the effects of the water slug, the resulting loads on downstream piping are still more severe than those loads which can be absored by the class of piping required based on pressure rating alone. Therefore, Schedule 160 was substituted for Schedule 80 which would normally be used for the given pressure / temperature rating.

Thermal f atigue of Class 1 piping: Heat transient analyses denonstrated that in the area of the loop seals f atigue damage was a potential problem. Mechani-Cal cyclic loading such as water slug discharge and seismic events compounded the effects of the thermal transient resulting from rapidly filling the " cold" loop seal wi th " hot" pressur izer steam. This effect also was mitigated by maintaining a higher loop seal temperature.

Cold Overpressure Mitigation Systen: Ordinarily one would assume the fluid oynanic loading resulting from this low pressure " cold" water solid discharge event would be enveloped by the much higher pressure water slug events which occur during plant at power conditions. However, our analysis shows that this event produces the limiting loads on the relief valve upstream piping and is therefore an integral part of the overall analysis, i Mechanical and Electrical Environmental Qualification - Safety Relief and Block valves As noted in FSAR Section 3.11 and Tab le 1.7-3, detailed information on the environmental qualification of Class 1E electrical equipment and safety-related mechanical equipment is provided directly to the NRC in separate submittals.

This progranmatic qualification information is contained in the following docu-ments:

1. "BVPS-2 Equi'pment Qualification Report - Environmental Qualifica-tion of Liass it tiectrical tquipment"

2. "BVPS-2 Environmental Qualification of Safety-Related Mechanical Equipment"

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. United States Nuclear Regulatory Commission Mr. Peter Tam, Project Manager Safety / Relief Valves Page 6 Specific conponent qualification methodology and results are contained in the separate canponent-specific qualification packages which have been previ-ously made available to the NRC for review.

The following electrical and nechanical environmental qualifications are submitted for the safety, relief, and block valves and are located in Attach-ment E.

1. " Methodology for Qualifying Westinghouse WRD Supplied NSSS Safety-Related Electrical Equipment", WCAP-8587 "Metnocology," Revision 6-A(hP), March 1983.

2. " Equipment Qualification Data Package Crosby Positon Indication Device", Westinghouse Class 3, EQDP-HE-7, Revision 1, October 1966.

3. " Equipment Qualification Data Package Safety-Related Limitorque Motor Operators (Qualification Group A)", EQDP-HE-1, Revision 3, October 1984.

4. " Equipment Cualification Data Package Garrett (PORV) Solenoid Oper-dted Pilot \ alve and Position Indication Device", Westinghouse Class 3, EQDP-HE-9, Revision 1, January 1985.

5. " Equipment Qualification Test Report Limitorque Motor Operator (Environmental and Seismic Testing Inside Containment) (HELB Environments)", wLAP-666/, bupp. 2-Hula, Revision 1, October 1984

6. " Equipment Qualifiation Test Report Crosby Position Indicating Devic ts (Environmental and Seismic Testing)", WCAP-8687, hupp.

2-'IT074, Revision 1, october 196d.

7. " Equi; ment Qualification Test Report Garrett (PORV) Solenoid Oper-ated 'ilot Valve and Position Indication Device (Environmental and Seismic Testing)", WCAP-6b8/, Supp. 2-H09A, Revision 1, January 1985.

8. System Component Evaluation Worksheets Equipment Mark No. EQ Package 1.D.

a. 2RCS*RV551A HE-07

b. 2RCS*RV5518 HE-07

c. 2RCS*RV551C HE-07 i d. 2RCS*PCV455C HE-09

! e. 2RCS*PCV4550 HE-09 i f. 2RCS*PCV456 HE-09

g. 2RCS*MOV535 HE-01 h, 2RCS*MOV536 HE-01

1. 2RCS*MOV537 HE-01

United States Nuclear Regulatory Commission Mr. Peter Tam, Project Manager Safety / Relief Valves Page 7

9. Mechanical Equipment Environmental Qualification Worksheet MEQ File No. Equipment Type 001-05 Motor Operated Gate Valves

. 001-07 Relief Valve 001-11 Power Operated Relief Valve Pertinent FSAR Information FSAR information which is pertinent to a thorough review of the adequacy of the safety and relief valves can be found in the following sections or tables and are included in Attachment F.

FSAR Title Appendix 3A Computer Program for Dynamic and Static Analysis of Seismic Category 1 Structures, Equipment, and Components 5.2 Integrity of Reactor Coolant Pressure Boundary 5.2.2 Overpressure Protection 5.2.2.1 Design Bases 5.2.2.2 Design Evaluation 5.2.2.3 Piping and Instrunentation Diagr ams 5.2.2.4 Equipment and Component Description b.2.2.5 Mounting of Pressuie Relief Devices 5.2.2.6 Applicable Codes and Certifications 5.2.2.7 Material Specification 5.2.2.8 Process Instrunentation 5.2.2.9 System Reliability 5.2.2.10 Testing and Inspection 5.2.2.11 Reactor Coolant System Pressure Control During Low Tenperature Operation 5.2.2.11.1 System Operations

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United States Nuclear Regulatory Commission Mr. Peter Tam, Project Manager Safety / Relief- Valves Page 8 FSAR Title (continued) 5.2.2.11.2 Evaluations of Low Temperature Overpressure Transients 5.2.2.11.3 Operating Basis Earthquake Evaluation-5.2.2.11.4 Administrative Procedures 5.4.13 Safety and Relief Valves 5.4.13.1 Design Bases 5.4.13.2 Design Description 5.4.13.3 Design Evaluation 5.4.13.4 Tests and Inspections Table 5.4-20 Pressurizer Valves Design Parameters 7.6.7 Interlocks for RCS Pressure Control During Low Temperature Operation 7.6.7.1 Analysis of Interlock 7.6.7.2 Pressurizer Pressure Relief System 7.6.7.3 Description of PPR System Interlocks Figure 7.6-7 Functional Diagran for PORV Interlocks for RCS Pressure Control During Low Temperature Operation All reports, FSAR sections and tables mentioned within this letter are enclosed or previously submitted and are intended to satisfy the NUREG-0737 documenta-tion requirements. If necessary, additional documentation to support the analysis will be submitted upon request.

r United States Nuclear Regulatory Commission Mr. Peter Tam, Project Manager Safety / Relief Valves Page 9 Therefore, the safety, relief and block valves installed at BVPS-2 meet NUREG 0737 Item 11.D.1 requirements and are adequate to perform their intended safety functions. If you have any questions related to this submittal, please contact me or members of my staff.

DUQUESNE LIGHT COMPANY 1

By

,l. (dj Carey V '

Sen(iorYicePresident ANU/ijr NR/AND/SR/VLVS Attactunents AR/NAR cc: Mr. P. Tam, Project Manager - w/ attachments Mr. J. Beall, NRC Sr. Resident inspector - w/o/ attachments Mr. L. Prividy, NRC Resident inspector - w/o/ attachments INP0 Records Center - w/o/ attachments NRC Docunent Control Desk - w/o/ attachments

United States Nuclear Regulatory Commission Mr. Peter Tan, Project Manager Safety / Relief Valves Page 10 COMMONWEALTH OF PENNSYLVANIA )

SS:

COUNTY OF BEAVER On this /M M day of /24 // ,

M , before me, a Notary Public in and for said Comdonwealth and County, personally appeared J.

J. Carey, who being duly sworn, deposed and said that (I) he is Senior Vice President of Duquesne Light, (2) he is duly authorized to execute and file the foregoing Submittal on behalf of said Company, and (3) the statenents set forth in the Submittal are true and correct to the best of his knowledge,

/

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