ML20205H996
| ML20205H996 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 03/31/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-1057, NUREG-1057-S04, NUREG-1057-S4, NUDOCS 8704010184 | |
| Download: ML20205H996 (92) | |
Text
{{#Wiki_filter:.y y . NUREG-1057 i Supplement No. 4 1 Safety Evaluation Report l related to the operation of
- Beaver Valley Power Station,
) Unit 2 Docket No. 50-412 Duquesne Light Company, et al. i U.S. Nuclear Regulatory Commission j Office of Nuclear Reactor Regulation 1 March 1987 (" "%,,
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# A NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources:
- 1. The NRC Public Document Room,1717 H Street, N.W.
Washington, DC 20555
- 2. The Superintendent of Documents, U.S. Government Printing Office, Post Office Box 37082, Washington, DC 20013 7082
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Referenced documents available for inspection and copying for a fee from the NRC Public Docu-ment Room include NRC correspondence and internal NRC memoranda: NRC Office of Inspection and Enforcement bulletins, circulars, information notices, inspection and investigation notices; Licensee Event Reports; vendor reports and correspondence; Commission papers; and applicant and licensee documents and correspondence. The following documents in the NUREG series are available for purchase from the GPO Sales Program: formal NRC staff and contractor reports, NRC-sponsored conference proceedings, and NRC booklets and brochures. Also available are Regulatory Guides, NRC regulations in the Code of Federal Regulations, and Nuclear Regulatory Commission Issuances. Documents available from the National Technical Information Service include NUREG series reports and technical reports prepared by other federal agencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission. Documents available from public and special technical libraries include all open literature items, such as books, journal and periodical articles, and transactions. Federal Register notices, federal and state legislation, and congressional reports can usually be obtained from these libraries. Documents such as theses, dissertations, foreign reports and translations, and non NRC conference proceedings are available for purchase from the organization sponsoring the publication cited. Single copies of NRC draf t reports are available free, to the extent of supply, upon written request to the Division of Technical Information and Document Control, U.S. Nuclear Regulatory Com-mission, Washington, DC 20555. Copies of industry codes and standards used in a substantive manner in the NRC regulatory process are maintained at the NRC Library, 7920 Norfolk Avenue, Bethesda, Maryland, and are available there for reference use by the public. Codes and standards are usually copyrighted and may be purchased from the originating organization or, if they arr. American National Standards, from the American National Standards Institute,1430 Broadway, New York, NY 10018.
NUREG-1057 - Supplement No. 4 Safety Evaluation Report related to the operation of Beaver Valley Power Station, Unit 2 Docket No. 50412 Duquesne Light Company, et al. U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation March 1987 p =* a., Nhff] i i i
ABSTRACT This report, Supplement No. 4 to the Safety Evaluation Report for the applica-tion filed by the Duquesne Light Company et al. (the applicant) for a license to operate the Beaver Valley Power Station, Unit 2 (Docket No. 50-412), has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission. This supplement reports the status of certain items that had not been resolved when the Safety Evaluation Report and its Supplements 1, 2, and 3 were published. I i l l Beaver Valley 2 SSER 4 iii 1
1 l TABLE OF CONTENTS Page l ABSTRACT .......................................................... iii 1 INTRODUCTION AND GENERAL DISCUSSION .......................... 1-1 1.1 Introduction ............................................ 1-1 1 2 SITE CHARACTERISTICS ......................................... 2-1 2.4 Hydrologic Engineering .................................. 2-1 2.4.3 Probable Maximum Flood on Streams and Rivers ..... 2-1 { 2.4.3.1 Ohio River Flood ........................ 2-1 3 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS ..... 3-1 3.6 Protection Against Dynamic Effects Associated With the Postulated Rupture of Piping ............................ 3-1 j 3.6.3 Deterministic " Leak-Before-Break" Evaluation To Eliminate Postulated Breaks as a Desi
..................gn Basis for High-Energy Piping ............. 3-1 3.10 Seismic and Dynamic Qualification of Seismic Category I Mechanical and Electrical Equipment .................... 3-10 3.10.1 Seismic and Dynamic Qualification of Electrical and Mechanical Equipment........................ 3-10 3.10.1.1 Introduction.......................... 3-10 3.10.1.2 Discussion............................ 3-10 l 3.10.1.3 Generic Items......................... 3-11 4 ' 3.10.1.4 Equipment-Specific Issues............. 3-12 3.10.1.5 Conclusions........................... 3-12 7 INSTRUMENTATION AND CONTROLS ................................. 7-1 f
4
- 7. 5 Information Systems Important to Safety ................. 7-1 7.5.2 Specific Findings ................................ 7-1 7.5.2.1 Emergency Response Capability, RG 1.97, Revision 2 Requirements ................. 7-1 I
Beaver Valley 2 SSER 4 v
1 TABLE OF C0r:4 TENTS (Continued) a P_a!Le 10 STEAM AND POWER CONVERSION SYSTEM .................. ......... 10-1 ,
- 10. 4 O the r Fe atu re s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-1 10.4.9 Auxiliary Feedwater System ...................... 10-1 13 CONDUCT OF OPERATIONS ........................................ 13-1 ,
13.2 Training ................................................ 13-1 13.2.1 Licensed Operator Training Program ....... . . .. 13-1 13.2.1.2 Beaver Valley Operator Cross-Training ' Program ........................... ...- 13-1 13.3 Eme rge ncy P l a nni ng . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13-2 13.3.2 Evaluation' of the Applicant.'s Onsite Emergency P1an ........................................... 13-2 13.3.2.12 Medical and Public Health Support .... 13-2 15 ACCIDENT ANALYSIS ............................................ 15-1 15.8 Anticipated Transients'Without Scram ......... ......... 15-1 APPENDICES A CONTINUATION OF CHRON0 LOGY OF NRC ST/MF RADIOLOGICAL REVIEW 0F BEAVER VALLEY POWER STATION, UNIT 2 B BIBLIOGRAPHY ' , D ACRONYMS AND INITIALISMS. s i E NRC STAFF CONTRIBUTORS AND' C0l45ULTANT
%s L STAFF SAFETY EVALUATION REPORT ON CONFORMANCE TO GENERIC LETTER 83-28, ITEM 2.1 (PART 2), EQllIPMENT CLASSIFICATION (REACTOR TRIP SYSTEM COMPONENTS) AND VENDOR INTERFACE N STAFF SAFETY EVALOATION REPORT ON CCNFORMANCE TO GEMRIC ~
LETTER 83-28, IT6114.5.2 (REACTOR, TRIP SYSTIM RELIABILITY,, i ' ON-LINE TESTING) , , O STAFF SAFETY EVALUATION. REPORT ON CONFORMANCE TO GENERIC LETTER 83-28, ITEM 3.1 (POST-MAINTENANCE TESTING. REACTOR =, TRIP SYSTEM COMP 0NENTS), ITEM"3.2 (POST-MAINTEhAF.E TESTING, ALL OTHER SAFETY-RELATED COMP 0NENT O , AND ITEM 4.5.1.(REACTOR TRIP SYSTEM RELIABILITY,; SYSTEM FUNCTIONAL TESTING) 's - Beaver Valley 2 SSER 4 , vi s ' l Y t 4\. , - - - , . , , - -
.s I ..
TABLE OF CONTENTS (Continued) o l APPENDICES (Continued) P ERRATA Q BEAVER VALLEY 2 SQRT REPORT TABLES Page 1.2 Open issues ................................................. 1-3 1.3 Backfit issues .............................................. 1-4 1.4 Co n f i rmato ry i s s ue s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 1.5 License condition issues .................................... 1-9 3.1 Summary of high-energy piping in the Beaver Valley Unit 2
" l ea k- be fo re-b rea k" eva l ua t i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-13 3.2 Standard Review Plans (SRPs) that address mitigation of flooding, missiles, and component support failures. . . . . . . . . . . . 3-13 3.3 Equipment audited by the SQRT ............................... 3-14 1
p l t s s 4 Li..
. Beaver Valley 2 SSER 4 vii "s,
i
E f 1 1 INTRODUCTION AND GENERAL DISCUSSION 1.1 Introduction u, The Nuclear Regula' tory Commission (NRC) Safety Evaluation Report (NUREG-1057) l (SER) on the application of the Duquesne Light Company (DLC or the applicant) l for a license to operate the Beaver Valley Power Station, Unit 2, was issued 4 in" October 1985. Supplementss 1, 2, and 3 were issued in May, August, and November, respectively, of 1986. This is the fourth supplement to the SER. The purpose of this fourth Supplemental Safety Evaluation Report (SSER 4) is to j , revise the SER by providing the results of the staff's review of new informa-tion subsequently submitted by the applicant. The information provided in letters referenced in this SSER must be acceptably documented in amendments J to the Beaver Valley Unit 2 Final Safety Analysis Report (FSAR) by the applicant before the unit is licensed. Each section or appendix of this SSER is designated and titled so that it cor-responds to the section or appendix of the SER that has been affected by the staff's additional evaluation. Except where specifically noted, the SSER does , not replace the corresponding SER section or appendix. Appendix A is a contin- , uation of the chronology *of events, including correspondence, leading to the ! publication of this SSER. Appendix B lists the references used. Appendix D is
- a list of abbreviations used in this supplement. Appendix E is a list of the principal contributors to this SSER. Appendix L, which was published in SSER 2, is being expanded in this supplement. Appendices N, 0, P, and Q are added to the SER in this supplement. Appendix P, " Errata," corrects errors in the SER and in SSER 1 through 3. No changes were made to Appendices C, F, G, H, I, J, K, and M.
u' j Tables 1.2, 1.3, 1.4, and 1.5, all corresponding to tables of the same numbers i in the SER and previous supplements, provide summaries of the status of open, backfit, confirmatory, and license condition issues, respectively. If the 1 status of an issue has changed sinc.e issuance of the SER, details of the change i are documented in this supplement. Actica items that resulted from the Three Mile IrMr< Unit 2 (TMI-2) accident have been addressed in the SER: Table 1.1 of thg SER )rovided cross-references of various items to sections in the 93 THI-2 action items that were not fully closed out in the SER have b w. 1 % q fled as open or confirma- l tory issues in the SER or its supplements. Closeout status of open or confirma-tory TMI-2 issues may be obtained by reviewing Tables 1.2 and 1.4 of this l supplement. , l Action items that res'.ited from the Salem anticipated-transient-without-scram (ATWS) event hav3 been and will be addressed in various SER supplements. .Close-out status of these items is presented in Section 15.8 of this supplement. ; Copiec M this ~SER are available for public inspection ,iMthe NRC Public Docu-
~
ment Rooni at 1717 il 5treet~ H.W. , Washington, D.C. , and et(the B. F. Jones Beaver Valley 2 SSER 4 1-1 3' .,
' ? A. -- , - -.
d Memorial Library, 663 Franklin Ave. , Aliquippa, Pa. Copies of this SSER are also available for purchase from the sources indicated on the inside front cover of this report. The NRC Project Manager is Peter S. Tam. He was assisted by Mr. Jaime Guillen, Project Engineer. Mr. Tam and Mr. Guillen may be contacted by calling (301) 492-9409 or by writing to the following address: Division of PWR Licensing-A U.S. Nuclear Regulatory Commission Washington, D.C. 20555 The next supplement (SSER 5) is expected to be issued in April 1987. [ (- l l l Beaver Valley 2 SSER 4 1-2
i Table 1.2 Open issues l Issue Status SER section l (1) Preservice/ inservice testing program , (a) PST Closed in SSER 3 3.9.6 (b) IST Under review 3.9.6 (2) Pump and valve leak testing Closed in SSER 3 3.9.6 (3) Inadequate core cooling instrumenta- Closed in SSER 2 4.4.7 tion (Item II.F.2 of NUREG-0737) (4) Preservice/ inservice inspection program (a) PSI Under review 5.2.4.1, 5.2.4.3, 5.4.2.2 (b) ISI Updated in SSER 1, 6.6 remains open (5) Safe and alternate shutdown Under review 9.5.1 (6) Management and organization Under review 13.1, 13.4, 13.5.1 (7) Cross-training program Closed in SSER 1 13.2.1.2 (8) Emergency preparedness plan Updated in SSER 4, 13.3.3 remains open (9) Initial test program Closed in SSER 3 14 I (10) Control room design review Updated in SSER 1, 18.1 remains open (11) Safety parameter display system Updated in SSER 1, 18.2 remains open i P Beaver Valley 2 SSER 4 1-3
Table 1.3 Backfit issues Issue Status
- SER section (1) Snow and ice load C 2.3.1 (2) Underestimation of atmospheric dispersion C 2.3.4, 15.4.8 conditions (X/Q) at exclusion area boundary and consequences of radioactive release (3) Potential for flooding from probable maximum C 2.4.2, 2.4.10 precipitation and Peggs Run (4) Steam generator level control and protection C2 7.3.3.12 (5) Motor-operated accumulator isolation valve C 8.3.1.12 (6) Spent fuel pool maximum heat load C 9.1.3 (7) Fire suppression in the cable spreading room A 9.5.1.6 (8) Class 1E power for lighting and communication C 9.5.2.1 systems (9) Application of GDC 5 to communication systems C 9.5.2.1 (10) Application of GDC 2 and 4 to communication C 9.5.2 systems (11) Application of GDC 4 to lighting systems C 9.5.3 (12) Illumination levels in excess of SRP criteria C 9.5.3 (13) Application of RG 1.26 to areas excluded by C 9.5.4-9.5.8 RG 1.26 (14) Air dryers for emergency diesel generator C 9.5.6 (15) Alarm for rocker arm lube oil reserve C 9.5.7 (16) Diesel lube oil fill procedure C 9.5.7 .
*A - Issue was discussed in appeal meeting, and partial resolution was ad-dressed in the SER (October 1985). Status updated in SSER 3.
C - Closed in SER (October 1985). C2 - Closed in SSER 2 (August 1986). I i s j Beaver Valley 2 SSER 4 1-4 I
Table 1.4 Confirmatory issues Issue Status SER section (1) Operating procedures for continuous Closed in SSER 3 2.2.2 communication links (2) Differential settlements of buried pipes Under review 2.5.4.3.3, 2.5.4.5 (3) Internally generated missiles (outside Unchanged from SER 3.5.1.1 containment) (4) Internally generated missiles (inside Unchanged from SER 3. 5.1. 2 containment) (5) Turbine missiles Unchanged from SER 3. 5.1. 3 (6) Analysis of pipe-break protection Unchanged from SER 3.6.1 outside containment (7) FSAR drawings of break locations Unchanged from SER 3.6.2 (8) Results of jet impingement effects Unchanged from SER 3.6.2 (9) Soil-structure interaction analysis Closed in SSER 1 3.7.3 (10) Design documentation of ASME Code Closed in SSER 2 3.9.3.1 components (11) Item I1.0.1 of NUREG-0737, pressure / Under review 3.9.3.2 relief valves (12) Seismic and dynamic qualification of Updated in SSER 4, 3.10.1 mechanical and electrical equipment remains open (SQRT) (13) Pump and valve operability assurance Under review 3.10.2 l (PVORT) (14) Environmental qualification of Under review - 3.11 ; mechanical and electrical equipment I (EQRT) (15) Peak pellet design basis Closed in SSER 1 4.2.1 (16) Discrepancies in.the FSAR Closed in SSER 1 4.2.2 (17) Rod bowing analysis Closed in SSER 1 4.2.3.1(6) (18) Fuel rod internal pressure Closed in SSER 1 4.2.3.1(8) (19) Predicted cladding collapse time Closed in SSER 1 4.2.3.2(2) Beaver Valley 2 SSER 4 1-5
Table 1.4 (Continued) Issue Status SER section (20) Use of the square-root-of-the-sum-of- Closed in SSER 1 -4.2.3.3(4) the-squares method for seismic and loss-of-coolant-accident load calculation (21) Analysis of combined loss-of-coolant- Under review 4.2.3.3(4) accident and seismic loads (MULTIFLEX) (22) Natural circulation test Updated in SSER 1; 5.4.7.5 remains open (23) Reactor coolant system high point vents Closed in SSER 3 5.4.12 (24) Blowdown mass and energy release Under review 6.2.1.3 analysis methodology (25) Containment sump 50% blockage assumption Updated in SSER 2; 6.2.2 remains open (26) Design modification of automatic reactor Under review 7.2.2.3 trip using shunt coil trip attachment (27) Automatic opening of service water Closed in SSER 1 7.3.3.10
- system valves M0V113C and 113D 1
(28) IE Bulletin 80-06 concerns Unchanged from SER 7.3.3.13 (29) NUREG-0737, Item II.F.1, accident Closed in SSER 1 7.5.2.2 monitoring instrumentation positions (30) Bypass and inoperative status panel Under review 7.5.2.4. (31) Revision of the FSAR--cold leg accumu- Closed in SSER 3 7.6.2.4 lator motor-operated valve position indication (32) Control sy' stem failure caused by Under review 7.7.2.3 malfunctions of common power source or . instrument line (33) Confirmatory site visit (a) Independence of offsite power Closed in SSER 1 8.2.2.3
- between the switchyard and Class IE j system t
! (b) Confirmation of the protective Closed in SSER 1 8.3.1.2 bypass Beaver Valley 2 SSER 4 1-6
Table 1.4 (Continued) Issue Status SER section (33) Confirmatory site visit (Continued) (c) Verification of DG start and load Closed in SSER 1 8.3.1.8 bypass (d) DG load capability qualification Closed in SSER 1 8.3.1.9 test (e) Margin qualification test Closed in SSER 1 8.3.1.10 (f) Electrical interconnection between Closed in SSER 1 8.3.1.13 redundant Class 1E buses (g) Verification of electrical Closed in SSER 1 8.3.3.5 independence between power supplies to controls in control room and remote locations (34) Voltage analysis--verification of test Unchanged from SER 8.3.1.1 results (35) Documentation of description and analysis Unchanged from SER 8.3.3.7.1 of compliance with GDC 50 (36) Completion of plant-specific core damage Unchanged from SER 9.3.2.2 estimate procedure before fuel load (37) Training program for the operation and Unchanged from SER 9.5.4.1 maintenance of the diesel generators (38) Vibration of instruments and controls on Unchanged from SER 9.5.4.1 diesel generator (39) Surveillance of lube oil level in the Closed in SSER 2 9.5.7 diesel generator rocker arm lube oil reservoir (40) Solid waste process control program Unchanged from SER 11.4.2 (41) TMI Action Plan items (a) III.D.1.1, postaccident reactor Under review 13.5.2 coolant leakage outside containment (b) II.K.1.5 and II.K.1.10, IE Bulletins Under review 15.9.2 on measures to mitigate small-break 15.9.3 LOCAs and loss of feedwater (c) II.K.3.5, automatic reactor Under review 15.9.9 coolant pump trip during LOCA l Beaver Valley 2 SSER 4 1-7 l l
Table 1.4 (Continued) Issue Status SER section (41) TMI Action Plan items (continued) (d) II.K.3.17, report on ECCS outage Under review 15.9.11 II.K.3.31, compliance with Closed in SSER 3 15.9.14 (e) 10 CFR 50.46 (42) Plant-specific dropped rod analysis Closed in SSER 2 15.4.2 (43) Steam generator tube rupture Under review 15.6.3 (44) Quality assurance program Closed in SSER 1 17.4 (45) Cross-training of Unit 1 & 2 operators Closed in SSER 4 13.2.1.1 (46) Control room isolation on high radiation Under review 7.3.3.9 signal (47) Review of procedures generation package Unchanged from 13.5.2 SSER 1 (48) Fire protection: Amendment 12 review and
> site visit (a) Amendment 12 review Closed in SSER 3 9.5.1 (b) Site visit Completed on 1/30/87 9.5.1 <
(c) Safety-related system Under review 9.5.1 fire-barrier deviations (49) Steam generator high-level trip as non- Unchanged from 7.3, protection system SSER 2 15.1.2 l (50) Implementation letter of ICCI system Unchanged from 4.4.7 SSER 2 (51) Superheated steam in valve house Opened in SSER 3; 3.6.1 due to steamline break under review (52) Initial testing t Unchanged from 14 (a) Accumulator isolation valves SSER 3 (b) 50V, P0, IST tests Under review 14 4 Beaver Valley 2 SSER 4 1-8
Table 1.4 (Continued) l Issue Status SER section l (52) Initial testing (continued) (c) Plant performance after MSIV Unchanged from 14 closure SSER 3 (d) Steam extraction system and Unchanged from 14 process computer SSER 3 Table 1.5 License condition issues License condition Status SER section (1) Emergency response capability, Specifics provided 7.5.2.1 RG 1.97, Rev. 2 in SSER 1 (2) Fire protection Opened in SSER 3 9.5.1 Beaver Valley 2 SSER 4 1-9
i ! 2 SITE CHARACTERISTICS I 2.4 Hydrologic Engineering l 2.4.3 Probable Maximum Floods on Streams and Rivers 2.4.3.1 Ohio River Flood The following four paragraphs replace the second paragraph, which addresses
" flood alert" in SSER 3.
In the SER, the staff stated that the Technical Specifications will require a
" plant flood alert" be issued when the Ohio River reaches a level of 690 feet msl [mean sea level]. By letter dated August 12, 1986, the applicant requested that Section 2.4.3.1 of the SER be revised to delete the proposed requirements of " flood alert."
The applicant's request that the SER be revised to delete the proposed Technical Specification requirement for " flood alert" is based on: (1) the absence of specific wording in the Unit 1 Technical Specification referring to a flood alert; (2) an interpretation, based on informal communications with the NRC staff, that a plant " flood alert" requires formal NRC notification; and (3) the Unit 1 Emergency Preparedness Plan (EPP) which provides for NRC notification for an emergency action level of " Alert" or higher initiated by a flood level exceeding 705 feet msl. The proposed Unit 2 Technical Specification does not contain specific wording requiring a " flood alert" criterion. However, Section 4.7.6.1 of both the pro-posed Unit 2 and current Unit 1 Technical Specifications contain surveillance requirements. These requirements specify that water level measurements at the intake structure be taken at least once per 24 hours with the Ohio River water level below 690 feet msl, and when the water level is at or above 690 feet msl, the measurements are to be taken at least once every 2 hours. The increased , l frequency of measurements constitutes a plant " flood alert" because operating personnel are on notice that a Technical Specification for a limiting condition for operation (LCO) may be exceeded. Therefore, both the current Unit 1 Tech-nical Specification and proposed Unit 2 Technical Specification meet the intent j of the " flood alert" requirement. Furthermore, the staff noted that the applicant will classify a 690-foot msl condition as an " Unusual Event" in the Emergency Preparedness Plan. Review of the EPP is addressed in Chapter 13 of the SER and its supplements. Beaver Valley 2 SSER 4 2-1
I i 3' DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT AND SYSTEMS 3.6 Protection Against Dynamic Effects Associated With the Postulated Rupture of Piping 3.6.3 - Deterministic " Leak-Before-Break" Evaluation To Eliminate Postulated Breaks as a Design Basis for High-Energy Piping By letter dated February 2,1987, Duquesne Light Company (the applicant) { submitted a report, "WHIPJET Program Final Report," prepared by Robert L. Cloud j and Associates on the technical bases-for eliminating postulated pipe ruptures
- as a design basis for high-energy piping. . Additional information was received j from the applicant in a letter dated February 13, 1987. The submittals were made to provide technical justification for the applicant in support of a re-quest for an exemption to General Design Criterion 4 (GDC 4) of Appendix A to.
l 10 CFR 50 regarding the need for protection against dynamic effects from postu-j lated pipe breaks. j The applicant's initial proposed program was submitted in letters dated-i September 6 and October 10, 1985. The staff began its review of the proposed { program shortly after receiving the submittals. The staff's initial response to the applicant's proposal is documented in a letter dated March 5, 1986. 1 i Subsequent to this response, the applicant, staff, and consultants had seven meetings (see Appendix A in SSER 1 through 4) to discuss the applicant's program. By means of deterministic fracture mechanics analyses, the applicant contends l that postulated double-ended guillotine breaks (DEGBs) of the high-energy fluid i piping identified in Table 3.1 will not occur'in Beaver Valley Unit 2 and there-fore need not be considered as a design basis for installing protective devices i i such as pipe whip restraints and jet impingement barriers to guard against the dynamic effects associated with such postulated breaks. No other changes in design requirements are addressed within the scope of the referenced reports; , i i.e., no changes to the definition of a loss of-coolant accident (LOCA) nor its i relationship to the regulations addressing design requirements for emergency 1 core cooling system (ECCS) (10 CFR 50.46), containment (GDC 16, 50) other engineered safety features, and the conditions for environmental qualification ; i of equipment (10 CFR 50.49). ' t j The Commission's regulations require provision of protective measures against j the dynamic effects of postulated pipe breaks in high-energy fluid system piping. Protective measures include physical isolation from postulated pipe rupture j locations if feasible, or the installation of pipe whip restraints, jet impinge-
; ment shields, or compartments. However, recent research performed by the NRC and industry, coupled with operating experience, have indicated that safety can be negatively impacted by the placement of protective devices such as pipe whip restraints. Studies completed by Lawrence Livermore National Laboratory under i contract to the NRC indicate that adverse safety implications can result from j requiring protective devices to resist the dynamic effects associated with j postulated pipe rupture (NUREG/CR-4263). The placement of pipe whip restraints l
1
; Beaver Valley 2 SSER 4 3-1 l
l 4 _ _ _ _ . . _ . _ _ - - - - - - - - - ~ - - - - -
would degrade plant safety if thermal growth is inadvertently restricted, reduces the accessibility for and effectiveness of inservice inspections, increases inservice inspection radiation dosages, and adversely affects construction and maintenance economics. An alternative to providing protective devices against the dynamic loads resulting from postulated pipe ruptures is made possible by the development of advanced fracture mechanics technology. These advanced fracture mechanics techniques deal with relatively small flaws in piping components (either postu-lated or real) and examine their behavior under various pipe loads. The objec-tive is to demonstrate, by deterministic analyses, that the detection of small flaws by either inservice inspection or leakage monitoring systems is assured long before the flaws can grow to critical or unstable sizes which could lead to large break areas such as the DEGB or its equivalent. The concept underlying such analyses is referred to as " leak-before-break." There is no implication that piping failures cannot occur, but rather that improved knowledge of the failure modes of piping systems and the application of appropriate leakage detection can reduce the probability of catastrophic failure to an insignificant value. On April 11, 1986, a final rule was published (51 FR 12502), effective May 12, 1986, amending 10 CFR 50, Appendix A, GDC 4 to allow the use of analyses to eliminate from the design basis the dynamic effects of postulated pipe ruptures of primary coolant loop piping for pressurized-water reactors (PWRs). Accept-able technical procedures and criteria for the leak-before-break technology are defined in NUREG-1061, Volume 3. This limited scope modification to GDC 4 was made by the Commission because safety and economic benefits could be quickly realized without extensive and time-consuming review and discussion if the scope were initially limited to the primary main loop piping of PWRs. Substantial evidence had already been developed to show that the leak-before-break concept was valid for primary main coolant loops of PWRs. The Commission decided not to defer the limited application of leak-before-break technology while the detailed provisions of the proposed acceptance criteria were being reviewed and approved. Many near-term operating license (NT0L) nuclear power plant units, including Beaver Valley Unit 2, and operating nuclear power plant units had requested and received exemptions for eliminating the dynamic effects of postu-lated pipe rupture from the requirements of GDC 4. On July 23, 1986, a proposed rule was published (51 FR 26393) to further amend GDC 4 to broaden the scope to cover all high-energy pfping in all nuclear power plants. The proposed amendment to GDC 4 allows exclusion from the design basis of dynamic effects associated with high-energy pipe rupture by application of leak-before-break technology. Only high-energy piping that meets rigorous ac-ceptance criteria in nuclear power units is covered. High-energy piping is defined as those systems having pressures exceeding 275 psig or temperatures exceeding 200 F. The proposed general acceptance criteria are based on NUREG-1061, Volume 3. Leak-Before-Break Evaluation Parameters Table 3.1 identifies the pipe lines, sizes, materials, and weld processes used to fabricate the high-energy lines that were evaluated for " leak-before-break." All of the evaluated piping is inside containment. In its review of the Beaver Valley 2 SSER 4 3-2
submittal, the staff evaluated the applicant's analyses and materials data with regard to the limiting location (s) determined by stresses in the piping, associated with the combined loads from normal operation and the safe shutdown earth-quake (SSE) and the fracture toughness properties of austenitic steel piping and associated weld materials potential loss of load-bearing capacity by mechanisms such as crack formation (by fatigue or stress-corrosion) or wall-thinning (by erosion or
- erosion-corrosion)
P size of postulated through-wall cracks that would leak at a detectable t rate under normal loads and pressure stability of a " leakage-size crack" under normal plus SSE loads and the expected margin in terms of load margin based on crack size
- leak-Before-Break Evaluation Criteria The NRC staff's criteria for evaluating the above parameters are delineated in NUREG-1061, Volume 3. These criteria are identical to those accepted in the limited scope modification of GDC 4 and to those offered in the proposed broad i scope modification of GDC 4. These criteria are enumerated in Chapter 5.0 of i NUREG-0679 and are as follows
(1) The loading conditions should include the static forces and moments (pres-sure, deadweight, and thermal expansion) due to normal operation, and the forces and moments associated with SSE. These forces and moments and the base metal and weld tensile and toughness properties are to be used to 3 define the locations that have the smallest margins against pipe rupture l in the pipe run (anchor to anchor).
- (2) For the piping run/ systems under evaluation, all pertinent information l should be provided which demonstrates that degradation or failure of the piping resulting from creep, corrosion, erosion-corrosion, stress-corrosion, fatigue, water hammer or other environmental conditions is not likely.
Relevant operating history should be cited, which includes system opera-tional procedures; system or component modification; water chemistry param-eters, ifmits, and control; resistance of material to various forms of stress corrosion, and performance under cyclic loadings. 2
! (3) Pipe lines are evaluated to determine whether there is a high probability '
of degradation or failure from indirect causes such as fires, missiles, and equipment failures and failures of systems or components in close proximity to the pipe.
) (4) The materials data provided should include types of materials and materials ;' specifications used for base metal, weldments and safe-ends, the materials properties including the J-R curve used in the analyses, and long-term effects such as thermal aging and other limitations to valid data (e.g., J ! Beaver Valley 2 SSER 4 3-3 1
1 I maximum, maximum crack growth). The piping materials must be free from. brittle cleavage-type failure over the full range of the system operating temperature. (5) A through-wall crack should be postulated at the limiting locations
- determined from criterion 1 above. The size of the crack should be large i j enough so that detection of leakage is assured using the installed leak i detection capability when the pipe is subjected to normal operational loads.
i .NUREG-1061, Volume 3, recommends the margin on the magnitude of. leakage be no less than a factor of 10 greater than the capability of the leakage
- detection system.
J (6) It should be demonstrated that the postulated leakage crack is stable under i normal plus SSE loads for long periods of time; that is, crack growth, if
- any, is minimal. The margin, in terms of applied loads, should be-deter-j mined by a crack stability analysis, i.e., that the leakage-size crack I will not experience unstable crack growth even if larger loads (larger j than design loads) are applied. This analysis should demonstrate that
; crack growth is stable and the final crack size is limited, so that a
! double-ended pipe break will not occur. 4 The stability analysis should compare the leakage-size crack to the critical-size crack. Under normal plus SSE loads, it should be demon-i strated that there is a margin of at least 2 between the leakage-size crack and the critical-size crack to account for the uncertainties inherent in the analyses, and leakage detection capability. )
; Staff Evaluation j
The staff, with the assistance of personnel from Energy Technology Engineering Center (ETEC) and Novetech Corporation, has evaluated the information presented i in the applicant's February 2, 1987 letter. The staff finds that the applicant
- has presented an acceptable justification for eliminating pipe whip restraints r in the reactor coolant system (RCS), residual heat removal (RHR) system, and r i safety injection system (SIS) piping at Beaver Valley Unit 2. The following paragraphs in this section present the staff's evaluation and conclusions, i
(1) Loads and Load Combinations Normal operating (ASME Code Service Level M loads, including pressure, dead- . I weight, and thermal expansion, were used to determine leak rate and leakage size l cracks. The stability analyses performed to assess margin against pipe fracture i at postulated faulted load conditions were based on normal plus safe shutdown i earthquake (SSE) loads. The loads used by the applicant for the leak rate and
- stability analyses were determined for the as-constructed piping systems con-figurations at Beaver Valley Unit 2.
In both the leak and stability analyses, the individual normal (pressure, ther- . mal, and deadweight) load components were summed algebraically and the seismic ! loads were added absolutely to obtain the bending moment in each of the principal i coordinate directions and the total axial force. The resultant bending moment I was obtained from the square root of the sum of the squares of the principal ' coordinate moments. i i Beaver Valley 2 SSER 4 3-4
Leak-before-break (LBB) evaluations were performed for the limiting location ! within each distinct pipe segment (i.e., pipe run having constant cross-section dimensions between anchor points). Limiting location was determined using the ) 3 recommendations in NUREG-1061, Volume 3, and is the location predicted to have the smallest margin against pipe rupture based on the material resistance to ductile crack instability and the combined normal plus SSE loads. In some instances one location was used to represent the limiting locations for several pipe segments in the LBB evaluation. This categorization is acceptable to the staff because the nature of the loading was not significantly different in the segments, and the segments were all in the same system and had the same cross-sectional dimensions and materials. As a part of its evaluation of these loads, the staff and its contractor, ETEC, conducted an audit at the offices of Stone and Webster Engineering Corporation (SWEC), the architect / engineer for Beaver Valley Unit 2. This audit consisted of a review of piping stress and thermal transient analysis procedures and cri-teria employed by SWEC and Robert L. Cloud and Associates, Inc., the applicant's consultants, in their analyses for the applicable piping systems in the WHIPJET program. A detailed review of the dynamic seismic analysis of the reactor cool-ant system (2 RCS-003-055-1) was performed by ETEC. On the basis of the information reviewed during this audit, the staff concluded that the procedures employed by the applicant are in accordance with ASME Code Section III requirements for Class 1 piping and are adequate for development of the input to all the stability and leak rate analyses. On the basis of its review, the staff finds that the load definition, method of load combination, and locations for which the LBB evaluations were performed are consistent with the recommendations in NUREG-1061, Volume 3, and provide a basis acceptable to the staff for LBB evaluation at Beaver Valley Unit 2. (2) Pipe Degradation Mechanisms The applicant assessed each LBB candidate line to determine if there is signifi-cant potential for degradation of piping integrity during service. This assess-ment included: comparison of the Beaver Valley Unit 2 candidate lines with lines identified in NUREG-0927, -0679, -0691, and -0582 as having a history of service cracking, and a comparison of the design, fluid, and operational conditions at Beaver Valley Unit 2 with those known to cause cracking or failure in piping. The degradation mechanisms evaluated were: intergranular stress corrosion crack-ing of austenitic steel, potential for dynamic loads (e.g., water hammer), low-cycle thermal fatigue, aging effects on austenitic steel (i.e., low-temperature sensitization), wall thinning from erosion, creep damage, and fatigue from design transients. Comparison with service experience shows that none of the candidate lines (those identified in Table 3.1) has a history of service cracking. Also, eval-uation of the plant-specific design, fluid, and operating conditions indicates: (a) intergranular stress corrosion cracking (IGSCC) in austenitic steel is not likely because the candidate lines were constructed using techniques that pro-duce an acceptable level of resistance to IGSCC, the fluid chemistry will be controlled during operation to minimize contaminants, and the dissolved oxygen Beaver Valley 2 SSER 4 3-5
is at a level that would normally preclude IGSCC; (b) crack initiation from vibratory fatigue is unlikely because there are no identified vibration sources and there are no socket welds, which may act as crack initiators, in the candi-date lines; (c) there is adequate fluid mixing to minimize flow stratification and associated low-cycle thermal fatigue; (d) degradation of austenitic steel from low-temperature aging is not likely because there are no cast pipe or i fittings in the candidate lines; (e) fluid conditions and use of erosion-resistant materials in the candidate lines preclude the likelihood of wall thinning by erosion; and (f) the relatively low temperatures (less than 650*F) of the lines preclude any significant creep damage. With respect to fluid transients, measures to minimize and to mitigate the fre-quency and effects of water hammer were reviewed and found to be acceptable. Provisions for minimizing water hammer effects were evaluated in accordance with the guidelines of NUREG-0927. Table 3-2 of NUREG-0927 indicates that for the piping systems in the WHIPJET program, concerns regarding water hammer are associated with: (a) voiding in the residual heat removal (RHR) and safety injection (SI) systems (as part of the ECCS) and (b) relief valve discharge loads in the reactor coolant pressurizer system. Table 3-2 of NUREG-0927 also indi-cates that preventive measures for water hammer should consist of: (a) appro-priate design and plant operation procedures for voiding in the RHR and SI sys-tems and (b) including the relief valve discharge load in pipe support and com-ponents design basis. A review of Section 4.2.2 of the WHIPJET report (Robert L. Cloud and Associates, 1987) verified that the guidelines of Table 3-2 of NUREG-0927 were adequately addressed. Furthermore, provisions for mitigating the frequency and effects of water hammer were also evaluated in accordance with the guidelines of NUREG-0582. On the basis of review of the information submitted by the applicant and com-parisons with the information in NUREG-0927, -0679, -0691, and -0582, the staff concludes that degradation of piping integrity from the mechanisms discussed above is unlikely in the candidate lines at Beaver Valley Unit 2. An assessment also was made to predict the inservice flaw growth from anticipated and postulated transient loads. The initial flaw size for each line was deter-mined from the Section XI (ASME Code) preservice acceptance standards and was assumed to be present in the line before service. The transients, stress and fracture mechanics analysis methods, and the material crack growth relationship used in the fatigue crack growth evaluation were reviewed and accepted by the staff; the results indicate that the postulated flaws would remain within the end-of-design-life size criteria established by the staff. On the basis of comparison of the design, material, operational and fluid conditions anticipated for the LBB candidate lines at Beaver Valley Unit 2 with prior service experience and conditions that may produce degradation in nuclear pipe, the staff concludes that inservice cracking and significant de-gradation are unlikely. Consequently, the staff concludes that LBB analysis is an appropriate basis to demonstrate the probability of pipe rupture for the candidate lines is acceptably low and to justify not installing protective de- , vices against the dynamic effects of postulated pipe breaks in the candidate pipe lines. Beaver Valley 2 SSER 4 3-6
(3) Indirect Sources of Pipe Rupture Pipe degradation or failure from indirect causes such as fires, missiles, and component support failure is prevented by designing, fabricating, and inspecting reactor compartments, components, and supports to staff criteria that reduce to a low probability the likelihood of the events or its impacting safety related components. The staff has reviewed the applicant's compliance with the criteria in SER Sections 3.4.1, 3.5.1.2, 3.9.3, 3.9.6, and 9.5.1. The staff's criteria are contained in standard review plans that are identified in Table 3.2. (4) Materials Data The reactor coolant, residual heat removal, and safety injection systems at Beaver Valley Unit 2 are constructed from Type 304, 304N, and 316 wrought austenitic steel pipe. Most austenitic welds are shielded metal arc welds (SMAWs); some welds in the RHR, RCS, and SIS are submerged arc welds (SAWS).
; Because archival material generally is not available for the material heats in the LBB candidate lines, the material properties for the leak rate and stability analyses were obtained from data in the literature.
Generally, the data from the literature came from several sources, and included from 3 to 12 test samples and 3 or more heats of material for each material specification. However, for some materials, such as Type 316 aus,tenitic steel, few data were available and data from Type 316 and 304 austenitic steel were combined. A conservative representation from the combined population was then used for the analysis. The applicant employed two techniques to ensure that the data from the litera-ture provided a reasonably conservative representation for materials in the LBB candidate lines at BVPS-2. First, the room temperature yield and ultimate strengths from the BVPS-2 materials were compared to those reported in the literature to ensure the BVPS-2 materials were bounded by the data in the litera-ture. The tensile strength properties were used as an initial screen because specific toughness values are not available for the BVPS-2 materials. In addi-tion, for welds, comparisons of welding parameters were made when available in the literature. Second, because data available to characterize the BVPS-2 mate-rials with respect to those described in the literature were limited, conserva-tive representations of the data in the literature were used for the BVPS-2 evaluation. This conservatism included using the lower bound stress / strain and toughness data for the stability analysis and margin assessment. For the leak rate analysis, the average stress / strain properties were used to obtain realistic leakage size cracks. The staff also compared the lower bound stress / strain and toughness relation-ships developed by the applicant to those developed previously and independently by other investigators. Comparison with data from the NRC Degraded Pipe Program, i Phase II (in progress) indicates that the austenitic steel base metal lower bound stress / strain and toughness relationships used for BVPS-2 are reasonable. Comparison with the information in NP-4690-SR (Electric Power Research Institute, 1986) shows that the BVPS-2 lower bound toughness relationships for austenitic SMAW and SAW and the lower bound stress / strain curve for SMAW are essentially identical to those used in Section XI of the ASME Code. The SAW stress / strain Beaver Valley 2 SSER 4 3-7
curve used in the BVPS-2 evaluation is consistent with the data in the NRC De-graded Pipe Program (in progress) and EPRI Special Report NP-4690-SR (July 1986) when compared to the yield strength measured for the BVPS-2 material, and pro-vides a reasonably conservative basis for the LBB analysis. On the basis of its review of the data submitted by the applicant, the staff concludes that the applicant complies with the recommendations in NUREG-1061, ( Volume 3, for data acquisition and, when limited data were available, made rea- l sonably conservative representations of the material properties in the LBB l candidate lines at BVPS-2. l (5) Leak Rate Criteria and Computations All LBB candidate lines at BVPS-2 are inside the containment. The applicant has proposed a leakage detection criterion that includes a detected leak rate of ! 0.5 gpm and, in accordance with the recommendation in NUREG-1061, Volume 3, a margin of 10 was applied to this leak rate to define the leakage size flaw used in the stability analyses. The basis for the 0.5 gpm leak rate is the presence inside containment of diverse and redundant leakage detection equipment and methods that are in compliance with Regulatory Guide 1.45, operating experience at Beaver Valley Unit 1, and in-containment surveillance requirements and admin-istrative procedures for locating unidentified leakage exceeding 0.5 gpm. The operating experience at Beaver Valley Unit 1, has shown that a mass inventory balance, which is performed at least every 72 hours, has an accuracy of about 0.2 gpm. The 0.5 gpm " threshold" for initiating in-containment surveillance and administrative procedures for locating unidentified leakage has been used successfully at Beaver Valley Unit 1. It is also in keeping with experience at other plants which exhibit a normal background unidentified average leakage rate between 0.1 gpm and 0.3 gpm. The applicant also has demonstrated with pressurized pipe tests that leak rates above 0.1 gpm can be readily detected visually. These experiments plus experience at Beaver Valley Unit 1 indicate that a 0.5 gpm leak rate can be reliably detected and located during s'ervice. The staff has reviewed the experimental results and the surveillance requirements and procedures, and finds they are acceptable and adequate to reliably detect 0.5 gpm unidentified leakage inside containment at BVPS-2, and provide an accept-able basis for defining leakage size flaws used for stability analyses. Detected leaks will be repaired within the system limiting conditions for opera-tion established in either technical specifications or administrative procedures. When leakage is detected in reactor coolant pressure boundary piping, Technical Specification 3.4.6.2 requires that the plant be in hot shutdown within 6 hours and in cold shutdown within the next 30 hours. Repair would be required before restart. When pipe leakage is detected in other portions of the piping in the WHIPJET program, the applicant has Indicated that the system will be isolated and de-clared inoperable and action will then be taken in accordance with the appli-cable technical specifications. The relationship between flaw size and leak rate was determined using computer software that includes an elastic plastic fracture mechanics routine for deter-mining crack opening area and a thermal hydraulics routine to compute mass flow
- through the postulated throughwall flaw.
Beaver Valley 2 SSER 4 3-8
The software has been benchmarked using both laboratory and service data. The , laboratory data included leakage measured through slits, holes, and cracks placed in ferritic and austenitic steel materials. The service data were obtained t from observed leakage through a cracked austenitic steel safe end in the Duane Arnold Nuclear Plant as described in NUREG-0531. Except for small flaws where plugging may occur, the software was judged to predict leak rates within a 125% accuracy. For small flaws (leak rates less than 0.1 gpm) where plugging may occur, the predicted leak rate could be 2 to 4 times larger than the actual leakage rates. The staff has reviewed the benchmark information supplied by the applicant and concludes that the software used to predict leak rate is adequately verified and has acceptable accuracy for application to LBB analysis when used with appro-priate margin on predicted leak rate. i (6) Margins on Load and Flaw Size The applicant has performed stability analyses to determine if the leakage size crack is stable at 1.4 x (normal plus SSE loads) and if the instability flaw size at normal plus SSE load is at least twice as long as the leakage size crack. The stability analyses were performed using elastic plastic fracture mechanics analysis methods, the load combination method described in item 1 of this sec-tion, lower bound stress / strain and toughness relationships discussed in item 4 of this evaluation, and the leakage size cracks determined for each size piping
! using the associated leak-detection criterion.
The elastic plastic fracture mechanics analyses were performed using the defor-mation plasticity failure assessment diagram (DPFAD) method. The computational results were obtained using computer software that has been benchmarked by com-parison with experimental data and other independently developed software. The staff reviewed the benchmark information supplied by the applicant and concludes that it has acceptable accuracy and can be used for the BVPS-2 crack stability
; analysis.
In accordance with the recommendations in NUREG-1061, Volume 3, stability analy-ses were performed for the material conditions associated with each pipe line; l namely, the weld (SAW and/or SMAW) toughness and stress / strain relationships and i the base metal toughness and stress / strain relationships. The limiting of these ! material conditions was used to assess margin on load and flaw size. The results from the stability analyses show that at all limiting locations in the LBB candidate systems a margin against pipe rupture of at least 1.4 exists on the normal plus SSE loads for the leakage size cracks, and that except for the 6-inch SIS lines, the critical throughwall flaw sizes at normal plus SSE loads are at least twice the length of the leakage size cracks. The stability analysis for the 6-inch SIS line indicates that the instability circumferential
, throughwall flaw length was 1.8 rather than 2 times the length of the leakage
- size flaw. The staff's independent stability computation indicates that a mar-gin of 2 exists for this line (see below). This line was accepted by the staff because the required margin on load was demonstrated, the instability flaw size l was determined conservatively as verified by independent calculations, and the i
10% difference in the applicant's calculations is within the computational accuracy of the analysis methods, i Beaver Valley 2 SSER 4 3-9
The staff also performed several independent stability computations to verify the applicant's results. These computations were performed using the "Z-Factor" method which is the basis for developing allowable flaw sizes for austenitic steel piping in Section XI of the ASME Code. This method is described in EPRI Special Report NP-4690-SR and compares favorably with currently available expe-rimental results for SAW austenitic steel piping (see NUREG-0531). The results from this independent analysis indicate that a margin against failure of 1.4 exists on applied normal plus SSE load for the leakage size crack and that the critical throughwall flaw size at normal plus SSE load is at least twice the length of the leakage size crack. Staff Conclusion j On the basis of its review of the information submitted by the applicant and the staff's independent computations, the staff concludes that Duquesne Light Company has provided technical justification for not installing protective de-vices against the dynamic effects of postulated pipe breaks in the Beaver Valley Unit 2 high-energy lines identified in Table 3.1 and illustrated in Figures 3.1 through 3.16 of the WHIPJET Program Final Report (Robert L. Cloud and Associates, 1987). The hardware that is not required to be installed is listed in Table 3.4 of the same report (1987). By letter dated March 3,1986 the staff informed the applicant that to implement the results of WHIPJET (i.e. to eliminate certain piping restraints) would in-volve a technical or schedular exemption. On the basis of the complete resolu-tion of all technical issues, a schedular exemption can be granted and would be issued as part of the operating license. 3.10 Seismic and Dynamic Qualification of Safety-Related Mechanical and Electrical Equipment 3.10.1 Seismic and Dynamic Qualification of Electrical and Mechanical Equipment 3.10.1.1 Introduction Evaluation of the applicant's program for seismic and dynamic qualification of safety related electrical and mechanical equipment consists of: (1) a determi-nation of the acceptability of the procedures used, standards fellowed, and the completeness of the program in general, and (2) an audit of sel wt.ed equipment items to develop a basis for the judgment of the completeness and adequacy of the implementation of the entire seismic and dynamic qualification program. Guidance for the evaluation is provided by the Standard Review Plan (SRP) Section 3.10; Regulatory Guides (RGs) 1.61, 1.89, 1.92, and 1.100; NUREG-0484; and Institute of Electrical and Electronics Engineers (IEEE) Standards 344-1975 and 323-1974. These documents define acceptable methodologies for the seismic qualification of equipment. Conformance with these criteria is sufficient to satisfy the applicable portions of General Design Criteria (GDC) 1, 2, 4, 14, and 30 of Appendix A to 10 CFR 50, Appendix B to 10 CFR 50, and Appendix A to 10 CFR 100. Evaluation of the program for Beaver Valley Unit 2 was performed by a Seismic Qualification Review Team (SQRT) which was assisted by staff con-sultants from the Idaho National Engineering Laboratory and is included in this supplement as Appendix Q. Beaver Valley 2 SSER 4 3-10 1
I 3.10.1.2 Discussion The SQRT reviewed the equipment dynamic qualification information contained in the Final Safety Analysis Report (FSAR) Sections 3.9.2 and 3.10 and made a site visit from September 30 through October 3, 1986. The purpose of the review was to determine the extent to which the qualification of the equipment, as installed at Beaver Valley Unit 2, meets the criteria described above. A representative sample of safety-related electrical and mechanical equipment, as well as instru-mentation, included in both nuclear-steam supply system (NSSS) and balance-of-I plant (B0P) scopes, was selected for the audit. Table 3.3 identifies the equip-ment audited. The site visit consisted of field observation of the actual, final equipment configuration and its installation. This was followed by a review of the corresponding design specifications, test documents, and/or analysis documents which the applicant maintains in its central files. The field installation of the equipment must be observed to verify and validate equipment modeling employed in the qualification program. In addition to the document reviews and equipment inspections, the applicant made a summary presentation of the maintenance, startup testing, and in-service inspection programs. The audit identified both generic and equipment-specific concerns. A summary of the issues and their disposition, if any, is detailed in the following sections and in Table 3.3. 3.10.1.3 Generic Items (1) During the field observation, the staff found that several of the equipment items did not have a model number or a serial number for identification. These items, instead, had a " number" put on by the utility. This made it very difficult to establish a permanent auditable link between the field equipment and the qualification documentation. On inquiry, the applicant indicated that, in the documentation, the applicant has the linkage estab-
' lished between the " mark numbers" and the model or serial numbers. This, however, inserts an extra layer of paperwork between the field item and the qualification documentation and has the potential of losing direct tracea-bility.
Therefore, the utility should resolve the issue of installing the model or serial numbers (as the case may be) provided by the supplier on the equipment in the field for positive identification and traceability. This is to be confirmed. (2) During field observation of the loop stop valve protection cabinet (NSSS-15), it appeared that the clearance between this unit and adjacent cabinets was not adequate. On inquiry, the staff learned that this problem may exist with other cabinets. Therefore, this problem of adequate clearance between adjacent cabinets should be addressed on a generic basis. The response should include examples of resolution for typical cases. (3) The completion of the program of verification of as-built loads for pumps and valves is to be confirmed. (4) The applicant is to inform the staff of the completion of the seismic and dynamic qualification program. The completion and confirmation must occur before fuel load. Beaver Valley 2 SSER 4 3-11
I 1 1 1 1 3.10.1.4 Equipment-Specific Issues I (1) Review of the qualification documentation for the residual heat removal i (RHR) system heat exchanger raised the following concerns: (a) the appropriateness of using Bijlaard analysis for the 24-inch nozzle-shell junction stresses (b) sizing calculations for the support lug welds (c) the use of specified lug dimensions in the support lug-shell junction stress analysis The above concerns must be resolved for the RHR heat exchanger to be qualified. (2) Field observation and review of qualification documents indicated the fol-lowing issues with respect to the alternate shutdown panel: (a) Finite element model used for the analysis was not authenticated. (b) There was neither a list nor qualification of internal Class IE instruments. (c) There was no permanent auditable link between the field item and the documentation. The qualification of this item is pending and requires resolution of the above issues of concern. (3) The review of qualification documentation for the motor-operated damper indicated that there were a substantial number (too many to list here) of anomalies detected. In some cases the acceptance criteria were changed. The applicant made an initial response to the anomalies at the time of the audit. However, most of these responses do not explain the reason for acceptability. The list of anomalies in Wyle Laboratories Report No. 58784 should be resolved and the reason for the acceptability should be discussed in each case. Also, the change of acceptance criteria, if any, should be justified. 3.10.1.5 Conclusions On the basis of observation of field installation, review of the qualification documentt, and responses provided by the applicant to the SQRT's questions dur-ing the audit, the staff found the applicant's seismic and dynamic qualification program to be well defined and adequately implemented. Upon closure of the issues identified in Sections 3.10.1.3 and 3.10.1.4, as well as in Table 3.3, the seismic and dynamic qualification of the safety-related equipment at Beaver Valley Unit 2 will meet the applicable portions of GDC 1, 2, 4, 14, and 30 of Appendix A to 10 CFR 50, Appendix B to 10 CFR 50, and Appendix A to 10 CFR 100. Since issues still need to be resolved, confirmatory issue 12 is kept open to track them. Beaver Valley 2 SSER 4 3-12 l
Table 3.1 Summary of high-energy piping in the Beaver Valley Unit 2
" leak-before-break" evaluation Pipe size Pipe material specifications /
Piping lines * (inches) weld processes ** Reactor coolant loops (A, B, & C) 8 SA 376 Type 304/SMAW bypass lines Safety injection lines into 6 SA 376 Type 316/SMAW loops (A, B, & C) cold legs Safety injection lines into 6 SA 376 Type 316/SMAW loops (A, B, & C) hot legs Accumulator injection lines into 12 SA 376 Type 316, SA 182 loops (A, B, & C) cold legs Type 304N, SMAW, SAW Pressurizer surge line 14 SA 376 Type 304, SA 182 Type 304N, SMAW, SAW Residual heat removal line from 12 SA 376 Type 316/SMAW, SAW loop A hot leg Residual heat removal lines into 10 SA 376 Type 316/SMAW, SAW loops B&C
*The piping lines and their limits are as illustrated in Figures 3.1 through 3.16 of the "WHIPJET Program Final Report"(Robert L. Cloud and Associates,Inc.,1987) **SMAW: Shielded metal arc welding process SAW: Submerged arc weld process Table 3.2 Standard Review Plans (SRPs) that address mitigation of flooding, missiles, and component support failures SRP Section Title 3.9.3 ASME Code Class 1, 2 and 3 Components, Coniponent Supports and Core Support Structure 3.9.6 Inservice Testing of Pumps and Valves 3.4.1 Flood Protection 3.5.1.2 Internally Generated Missiles (Inside Containment) 9.5.1 Fire Protection Program Beaver Valley 2 SSER 4 3-13 f
- __ - . _ _ _ _ _ _ ._m _ . _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ ____._. ________ _ _.__ _ . . _ . .
Tab 12 3.3 Equipment audited by the SQRT j' cx Appiteant ID Status Number Equipment Number Safety function Findings Resolution
'ol N555-2 Steam generator 2M55*FT475 Measures steam generator Qualified l 4 flow transmitter flow rate J ne Reactor vessel 2RCS*HCV250A Vents hydrogen from reactor Qualified j O M555-3 letdown modulat- vessel head and modulates q e M ing valve letdown to pressure i
na relief tank Acts as a PORV block valve 1. Valve and operator qual- 1. A walkdown of To be
!$ N555-4 Motor-operated 2RCS*MOV536 ified but its motor con- electrical systems confirmed Q gate valve as described in trol cabinet was mounted
- in close proximity to Duquesne Light's neighboring cabinet so project procedure that seismic impacting BVM 236 is in prog-could occur. ress and will re- ,
- 2. Piping analysis param- solve such ;
i eters for valve were interferences. [ about 20K in error. 2. Piping audit program documented in 2 BVM-156 Rev. 3 will reconcile this discrepancy. w
- 1. Appropriateness of using 1. Justify use of Bij- To be
$ M555-8 RHR heat exchanger 2RHS*E21A Removes residual heat from reactor core Bijlaard analysis for laard technique for resolved for nozzle and lug these particular
'J> stresses is questioned, situations. Provide the sizing
- 2. No sizing calculations 2.
for weld stresses in calculations for the lugs. No weld size on welds on support > drawing. lugs. Verify weld size in the field as per drawing. N555-11 deactor coolcat 2RCS*P21A Used for power production; Qualified pump assembly -must be able to coast down
.and seals in case of loss of power N555-12 Reactor vessel 2RCS*50V2004 Vents He from reactor As-built loads not recon- To be reconciled by an To be 1etdown isola- vessel head; allows 50 gpa ciled yet. ongoing program. It is confir1med l
tion valve letdown to pressure relief being handled on a gener-7 l tank ic basis. See resolution 1 on N555-4. M555-13 Low-head safety 2515*P12A Provides emergency core Method of load combination Load combination method Qualified I cooling injection during may be nonconservative. It was examined for all injection pump accident was satisfactory for this pumps chosen for this equipment. audit. It was satisfac-
,> tory for all equipment audited.
1 i I
J i l l Table 3.3 (Continued) i s"o Applicant ID Number Equipment Number Safety function Findings Resolution Status 5 h555-14 Plant safety PNL*2RPU-A Provides postaccident Field mounting was not the Stiffness of actual Qualified { < monitoring sys- monitoring of reactor same as test mounting. field mounting was a E tem cabinet vessel level and core judged to be sufficient ' ! - cooling thermocouple to not invalidate the testing. 4
" M555-15 Loop stop valve RK*2W-REL-A Provides loop isolation / Lack of adequate clearance The problem is to be re- To be m protection cabi- block safety injection to prevent interaction. solved on a generic confirmed 2
net for isolated loop; inter-lock system to prevent basis. See resolution on N555-4. a improper reopening of
! loop stop valves and reactor coolant pump
, protection N555-16 Motor-operated 2CHS*LCY1158 Opens path from refueling Leaking significantly. The applicant is com- Qualified j gate valse water storagetank to mitted to repacking and charging pumps' suction stopping the leak. for injection mode M555-19 Nuclear instru- RK*2NUC-INS Provides alarm and indicat- Qualified 1 Y mentation sys- ing function of reactor
! >a tem cabinet status
, w M555-20 Centrifugal 2CH5*P21A Provides safety grade cold Qualified 1 charging / safety shutdown and high head j injection pump safety injection ) N555-21 7300 printed 2RCS*ZT2508 Provides test capability Qualification was briefly Complete circuit card for valve position feed- reviewed to see if the back documentation was complete. ' l l BOP-1 Motor-operated 2CCP*MOV150-2 Provides containment Valve and operator qualified See generic resolution To be l butterfly valve isolation but motor control cabinet on N555-4 confirmed I j mounted with cable tray in i close proximity. 80P-5 Feedwater iso- 2FW5*HYV157C Provides rapid stoppage of Qualified i lation valve feedwater flow in the ! event of a main steamline break 90P-8 Alternate shut- PNL*2ALTSHUTDN Provides alternate shut- 1. Finite element model for 1. To be authenticated. To be down panel down capability analysis not authenti- 2. To be completed. qualified I cated. 3. To be handled on a
- 2. Internals IE instruments generic basis.
i not qualified yet. I'
- 3. No pennanent auditable trail between field and document.
4
Tabla 3.3 (Continued) co Applicant ID Mumber Equipment Number Safety function Findings Resolution Status BCP-9 Central station 2HVR*ACU2078 Removes equipment heat load Qualified @ during DBA A/C units w Suppifes Westinghouse 1. One circuit breaker 1. It was found that a To be - BCP-14 Vital bus dis- PNL*VITBS2-IC sin 11ar 3 pole confirmed 7 tribution panel Class IE loads (Heineman model M board CD3-AO-DU-24fs VAC) breaker with one was not specifically pole being used for m mentioned in test an alarm function $ report. was tested. This is m 2. No manufacturer's model resolved. " number was attached 2. Qualification docu-A to panel. The Beaver mentation referred Valley mark number to mark number was attached. rather than model number. Qualifica-tion package in-cluded a drawing showing mark number. BOP-15 Air-operated 2 SIS *ADV8508 Not safety-related; The Beaver Valley mark Beaver Valley personnel Qualified control valve only needs to maintain number had been changed provided documentation pressure boundary; it is before the audit. Quali- of the changed mark w opened periodically for fication documentation number. e $ the accumulator check referred to old mark number. valve leakage test Manufacturer's model number was not used. BCP-17 Vital bus UPS*VITBS-2 Supplies 120-V ac instru- Two types of integrated cir- These two ampitfiers were Qualified uninterruptable ment, control, and power cuit operational amplifiers found not to perform a power supply for engineered safeguards were not qualified. Class 1E safety function. system protection channels and other Class IE, 120-V ac loads BCP-19 Quench spray 2QSS*P24A Injects sodium into the Qualified chemical injec- quench spray system tion pump BCP-20 Service water 25WS*P21A Supplies cooling water for Qualified pump various purposes through-out plant BCP-23 Motor-operated 2HVC*M00206B Isolate a standby unit Deficiencies reported from All the deficiencies To be damper for control room air the test have not been satis- must be satisfactorily confirmed conditioning factorily resolved yet. disposed of. BCP-24 Fuel oil pres- 2EGF*PS202-2 Starts the backup diesel Qualification was briefly Complete sure switch generator fuel pump if the checked for completeness loss of fuel pressure of qualification package. occurs
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s 3. 7.5 Information Systems Important to Safety " N t ' ' 7.5.2 Specific Findings. V 's \; o 7.5.2.1 Emergancy Response Capability, RG l'07, Revision 2, Requirements l r In Appendix G to Supplement 1 of the SER, the %taff reported that the applicant had identified seconairy system radiation as a type A variable. Type A variables s
; are plant-specific variables that are identified by the_ utility and are those ,'
1, variables that provide the primary information ' required to permit the control
~
i y room operator to take specific, manually controlled' safety actions. In a letter riated Decetter 23, 1986, the applicant provid9d marked-up FSAR sections pnpa'iing to' delete the secondary system' radiation variable as a type A and type.O variable (Regulatory Guide (RG) 1.97 does not list secondary sys- , tem radiatico as a typs B variable), and retain it only as a type E variable. . s The applicant has determined that since the primary purpose of secondary system radiation monitoring is "or monitoring of vented fluids from the steam generator t ? safety valves or atmospheric dump valves,-it thould be classified as type E g < category 2 only. The applicant furtner statts that more direct indications of , 4 steam generator tube rupture are from steatr generator level indications wbjch !' , 1 are qualified Class 1E and are in continuous operation. The steam generator - 1 level indicators are listed as type A variables by the applicant. The staff , finds that these changes are in accordance with the guidancre contained in RG 1.97 and are therefore acceptable. s
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I t J i 10 STEAM AND POWER CONVERSION SYSTEM 10.4 Other Features 10.4.9 Auxiliary Feedwater System i In the SER (p. 10-20) the staff stated that a minimum dedicated volume of water in the primary plant demineralized water storage tank (PPDWST) of 140,000 gallons is reserved for the auxiliary feedwater system (AFWS). This volume would ensure reactor coolant system cooldown to the residual heat removal (RHR) system cut-in temperature of 350*F in 7 hours (3 hours in hot standby and 4 hours for cooldown), i assuming no makeup water to the PPDWST. Because of level instrumentation design limitations, a technical specification requirement of 140,000 gallons is not
, possible. In FSAR Amendment 12, the applicant revised the usable PPDWST volume to be approximately 127,500 gallons, and proposed a Technical Specification limit . of 127,000 gallons. As a result, the design basis for the PPDWST has been changed to be consistent with the Beaver Valley Unit 1 Technical Specification; ! i.e., the PPDWST volume will support 9 hours at hot standby (rather than 3 hours
! in hot standby plus 4 hours in cooldown). The PPDWST has connections to the i 600,000 gallon _ demineralized water storage tank, a non-safety tank, which is used ) for normal makeup. However, the service water system may serve as a safety-i related, long-term, backup source of auxiliary feedwater for the steam generators. j On this basis, the staff concludes the PPDWST inventory of 127,000 gallons to be i acceptable. l
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t 13 CONDUCT OF OPERATIONS 13.2 ' Training
! 13.2.1 Licensed Operator Training Program 13.2.1.2 Beaver Valley Operator Cross-Training Program , In SSER 1,.the staff found the applicant's cross-training program acceptable with two exceptions:
i / (1) reference to dually licensing operators (2) limiting training based on the format of licensing examinations
- By letter dated July 21, 1986, the applicant responded to these two concerns. r Also, by letter dated September 30, 1986, the applicant submitted, under Unit I docket number, revisions to the Nuclear Group Training Administrative Manual, i
which included the cross-training program. The staff's evaluation follows. i The revised cross-training program reflects the possibility of issuance of e either an individual Unit 2 license or a dual (Unit 1/2) license. However, j the; applicant has informed the staff, in a meeting at Region I on January 5, 1987, that there is currently no plan to dual-license any operator. In addi-
) , tion, the applicant's January 27, 1987 submittal on open issue 6 clearly: shows parallel and separate operating organizations for Units 1 and 2. The commit-ment is acceptable to the staff.
t The revised cross-training program has deleted references to the type of ex- : j amination to be administered by NRC Region I in the Scope of. Training, License Review Series Phase, and Simulator Training Phase. This is acceptable to the 4 staff. The revised cross-training program has expanded.the scope of personnel who may enter the program to include " individuals completing the initial. licensed operator training program for Unit 2" as well as individuals who are licensed on Unit 1. Since the objective of-the cross-training program'has been' revised , to " prepare individuals for licensing on Unit 2, either as an individual license I or a dual (Unit 1/2) license," the need for personnel entering the program to be licensed on Unit I has been negated. Completing the initial license training program for Unit 1 is adequate to meet the requirements for licensing at a pre-critical facility (refer to H. R. Denton's letter to all power reactor appli-cants and. licensees, dated March 20,1980) since the requirements for licensing at a' precritical facility are less restricitive than the requirements for li-l censing at an operating facility. The staff finds the expansion of the scope of personnel entering the cross-training program to be acceptable. In summary the staff finds the cross-training program submitted by letter on September 30, 1986 to be acceptable. This closes confirmatory issue 45. Beaver Valley 2 SSER 4 13-1
13.3 Emergency Planning 13.3.2 Evaluation of the Applicant's Onsite Emergency Plan 13.3.2.12 Medical and Public Health Support In the SER, the staff stated that the applicant shall certify annually that the letters of agreement are current. By letter dated December 24, 1985, the applicant commented on the SER and specifically stated that the Emergency Preparedness Plan (Issue 8. Rev.1) has been revised to state that the letters of agreement will be certified to be current on an annual basis. This is responsive to the staff's concern. )
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Beaver Valley 2 SSER 4 13-2
)
15 ACCIDENT ANALYSIS 15.8 Ant $cipated Transients Without Scram Status of Salem ATWS Event Issues On July 3,1983, the NRC issued Generic Letter (GL) 83-28 as a result of the anticipated-transient-without scram (ATWS) events at Salem Nuclear Generating Station. This letter addressed actions to be taken by licensees and applicants to ensure that a comprehensive program of preventive maintenance and surveil-lance testing is implemented for the reactor trip breakers in pressurized-water reactors. The staff has completed its review of parts of the applicant's response to GL 83-28 and will document its results in appendices to the SER. The following list serves as a record of completion of staff review, and shows where individual safety evaluations may be found: Item 1.1, Post-Trip Review (Appendix K, SSER 1) Item 2.1, Equipment Classification and Vendor Interface (Reactor Trip System Components) (Appendix L, SSER 2 and SSER 4) Items 3.1.1 and 3.1.2, Post-Maintenance Testing Reactor Trip System Components (Appendix 0, SSER 4) Items 3.1.3 and 3.2.3, Postmaintenance Testing in Technical Specification That Could Degrade Safety (Appendix M, SSER 2) Items 3.2.1 and 3.2.2, Post-Maintenance Testing - All Other Safety-Related Components (Appendix 0, SSER 4) Item 4.1, Trip System Reliability (Appendix J, SSER 1) Items 4.2.1 and 4.2.2, Preventive Maintenance Program for Reactor Trip Breakers - Maintenance and Trending (Appendix J, SSER 1) Item 4.5.1, Reactor Trip System Reliability - System Functional Testing (Appendix 0, SSER 4) Item 4.5.2, Reactor Trip System Reliability - On-line Testing (Appendix N, SSER 4) Status of Im)1ementation of 10 CFR 50.62, AWS Mitigation System Actuation Circuitry (A4 SAC) In SSER 2, the staff indicated that the applicant will commit to a schedule for meeting the requirements of the subject regulation. By letter dated Beaver Valley 2 SSER 4 15-1
November 10, 1986, the applicant committed to implement the AMSAC design no later than July 1989. The date committed to is consistent with the staff's position and is, therefore, acceptable. The staff has not completed its review of the Beaver Valley Unit 2 AMSAC design for compliance with the ATWS rule, 10 CFR 50.62. However, staff review and approval are not a requirement for plant licensing. As stated in SSER 2, the staff has reviewed and approved the Westinghouse generic AMSAC design, and the staff will review the Beaver Valley Unit 2 plant-specific design. Licensing action TAC 62943 has been opened to track the staff's review of the Beaver Valley Unit 2 design; result of the review will be documented in a future safety evaluation report not associated with the SER or its supplements. On this basis, the staff will take no more prelicensing action on this issue. I Beaver Valley 2 SSER 4 15-2
APPENDIX A CONTINUATION OF CHRONOLOGY OF NRC STAFF RADIOLOGICAL REVIEW 0F BEAVER VALLEY POWER STATION, UNIT 2 September 30 - Seismic Qualification Review Team (SQRT) and Pump and October 3, 1986 Valve Operability Review Team (PVORT) onsite audit (con-firmatory issues 12 and 13). September 30, 1986 Letter from applicant (filed under Beaver Valley Unit 1 docket) addressing, among other things, Unit 1/2 operator cross-training program. October 8, 1986 Letter from applicant submitting Amendments 8 and 9 to Indemnity Agreement No. B-73. October 9, 1986 Letter from applicant providing current status of Safety Evaluation Report (SER) outstanding issues. October 16, 1986 Letter from applicant providing additional information on { control room isolation on high radiation (confirmatory i issue 46) in response to the staff's letter of May 23, 1986. October 20, 1986 Letter from applicant providing information regarding initial operator licensing examination. October 23, 1986 Letter from applicant transmitting final report on the WHIPJET program (use of leak-before-break methodology on balance-of plant piping). October 23, 1986 Letter from applicant providing partial response to confirmatory issue 2, differential settlement of buried pipes. October 23, 1986 Letter to applicant requesting cooperation in onsite review by Equipment Qualification Review Team. October 27, 1986 Letter from applicant providing response to questions raised in the SQRT and PVORT site visit. November 5, 1986 Meeting with applicant on fire protection issues (meeting summary dated November 18, 1986). November 6, 1986 Letter to applicant transmitting first draft of Technical Specifications. November 10, 1986 Letter from applicant submitting schedule to comply with 10 CFR 50.62, mitigation of anticipated-transient-without-scram events. Beaver Valley 2 SSER 4 1 Appendix A
i 4 November 14, 1986 Letter from applicant providing revision on Quality '~ Ass'urance Program (Preliminary Safety Analysis Report (PSAR) Section 17). November 17, 1986 Letter to applicant providing copy of Environmental Assessment and Finding of No Significant. Impact regarding request to extend Construction Permit No. CPPR-105. November 17, 1986 Letter from applicant transmitting Appendix C of the WHIPJET final report. November 18-20, 1986 Equipment Qualification Review Team onsite audit l (confirmatory issue 14). i November 21, 1986 Letter to applicant transmitting typed version of first. . 1 draft of Technical Specifications. November 25, 1986 Letter to applicant transmitting order to extend
- Construction Permit CPPR-105 to December 31, 1987.
November 26, 1986 Letter to applicant requesting additional information on inservice testing of pumps and valves (open issue 1 (b)). November 28, 1986 Letter from applicant commenting on draft. Technical j Specifications the staff transmitted on November 6, 1986. ! December 1, 1986 Letter to applicant providing 20 copies of SER Supple- : 1 ment 3 (SSER 3). December 3, 1986 Letter to applicant transmitting draft safety evaluation on procedures generation package (PGP), confirmatory l
- issue 47.
1 December 3, 1986 Letter from applicant submitting updated Environmental Qualification Program for Safety-Related Mechanical , Equipment, in response to staff questions 270.2 and 270.3. j i December 9-11, 1986 First meeting with applicant at the site to discuss i issues in the staff's draft Technical Specifications i (Summary dated January 7, 1987). I December 12, 1986 Letter from applicant requesting approval to use American
;. Society of Mechanica1' Engineers.(ASME) Code Case.
N-253/N253-1. 4 December 19, 1986 Letter from applicant providing additional information on i effect of superheated steam to equipment inside the main steam isolation valve (MSIV) house. December 23, 1986 Letter from applicant informing staff of reclassification j of secondary system radiation monitors to E2. This is a Regulatory Guide 1.97 issue. 1 4 Beaver Valley 2 SSER 4 2 Appendix A l _ _ _ _ _ _ _ _ . , ,. ....._m.e-. r -. - . . _ . - . _ , . _ _ _ . - -~.-_-- , _
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December 23, 1986 Letter from applicant informing staff of the applicant's understanding of status of SER outstanding issues. December 31, 1986 Letter from applicant discussing the use of Appendix I to 10 CFR 50 versus the use of the document RM-50-2 in developing the Offsite Dose Calculation Manual. January 16, 1987 Meeting with applicant to communicate staff concerns as a result of review of the WHIPJET program report. (Summary dated February 11, 1987.) February 2,1987 Letter from applicant transmitting information requested by the staff on WHIPJET in the January 16, 1987 meeting. Letter also requested schedular exemption from General Design Criterion 4 be granted for piping included in the WHIPJET program. February 13, 1987 Letter from applicant transmitting revised information on WHIPJET. Beaver Valley 2 SSER 4 3 Appendix A
APPENDIX B BIBLIOGRAPHY Duquesne Light Company, March 30, 1984 letter from E. J. Woolever (DLC) to D. G. Eisenhut (NRC), " Response to Generic Letter 83-23." Electric Power Research Institute, " Evaluation of Flaws in Austenitic Steel Piping," prepared by the Section XI Task Group for Piping Flaw Evaluation, ! Special Report NP-4690-SR, July 1986. l l U.S. Nuclear Regulatory Commission, " Degraded Pipe Program, Phase II, in ! process.
-- , Generic Letter 83-28 from D. G. Eisenhut (NRC) to all licensees of operating reactors, applicants for operating license, and holders of construction permits, " Required Actions Based on Generic Implications of Salem ATWS Events," July 8, 1983. -- , NUREG-0484, " Methodology for Combining Dynamic Responses," September 1978; Rev. 1, May 1980. -- , NUREG-0531, " Investigation and Evaluation of Stress-Corrosion Cracking in Piping of Light Water Reactor Plants," U.S. NRC Pipe Crack Study Group, February 1979. -- , NUREG-0582, " Water Hammer in Nuclear Power Plants," July 1979. -- , NUREG-0679, " Pipe Cracking Experience in Light Water Reactors," August 1980. -- , NUREG-0691, " Investigation and Evaluation of Cracking Incidents in Piping i in Pressurized Water Reactors," September 1980. -- , NUREG-0927, " Evaluation of Water Hammer in Nuclear Power Plants,"
March 1984.
-- , NUREG-1000, " Generic Implications of ATWS Events at the Salem Nuclear Power Plant," Vol.1, Generic 1mplications, April 1983; Vol. 2, Licensee and Staff Actions, August 1983. -- , NUREG-1061, Volume 3, " Report of the U.S. Nuclear Regulatory Commission Piping Review Committee, Evaluation of Potential for Pipe Breaks," !
November 1984. !
-- , NUREG/CR-4263, " Reliability Analysis of Stiff Versus Flexible Piping, Final Project Report," May 1985.
Wyle Laboratories, Report No. 58784, October 1983 [ appears as Appendix A to ITT General Controls Engineering Report 730.1.140, " Test Report for Requalification of ITT GC NJ 90 Series Hydramotor Actuators," April 24,1984]. Beaver Valley 2 SSER 4 1 Appendix B
INDUSTRY STANDARDS Institute of Electrical and Electronic Engineers Standard 323-1974, "IEEE Standard for Qualifying Class IE [IE] Equipment for Nuclear Power Generating Stations." Standard 344-1975, "IEEE Recommended Practices for Seismic Qualification of Class IE Equipment for Nuclear Power Generating Stations." a i 1 Beaver Valley 2 SSER 4 2 Appendix B
APPENDIX D ACRONYMS AND ABBREVIATIONS ac alternating current A/C air conditioning AFWS auxiliary feedwater system AMSAC ATWS Mitigation System Actuation Circuitry ASME American Society of Mechanical Engineers ATWS anticipated transient without scram B0P balance of plant BVPS-2 Beaver Valley Power Station Unit 2 CFR Code of Federal Regulations DBA design-basis accident DEGB double-ended guillotine break DLC Duquesne Light Company DPFAD deformation plasticity failure assessment diagram i ECCS emergency core cooling system i EDO ! NRC Executive Director for Operations EPP Emergency Preparedness Plan EQR1 Equipment Qualification Review Team ! ETEC l Ene.rgy Technology Engineering Center FSAR Final Safety Analysis Report GDC general design criterion (a) GL generic letter ICCI inadequate core cooling instrumentation IE NRC Office of Inspection and Enforcement IEEE Institute of Electrical and Electronics Engineers IGSCC intergranular stress corrosion cracking INEL Idaho National Engineering Laboratory IS initial startup ISI inservice inspection IST inservice testing LBB leak before break LCO limiting condition for operation LOCA loss-of-coolant accident MSIV main steam isolation valve ms1 mean sea level MWR maintenance work request Bewer Valley 2 SSER 4 1 Appendix D
i NRC U.S. Nuclear Regulatory Commission NSSS nuclear steam supply system NT0L near-term operating license PGP procedures generation package P0 preoperational PORV power-operated relief valve PPDWST primary plant demineralized water storage tank PSAR preliminary safety analysis report PSI preservice inspection PST preservice testing PVORT Pump and Valve Operability Review Team ! RCS reactor coolant system regulatory guide RG RHR residual heat removal RTS reactor trip system i SAW submerged arc weld SER Safety Evaluation Report SIS safety injection system SMAW shielded metal arc weld 50V system operability verification SQRT Seismic Qualification Review Team SRP Standard Review Plan
- SSE safe shutdown earthquake l SSER Supplemental Safety Evaluation Report SWEC Stone and Webster Engineering Corporation 1
1 TMI-2 Three Mile Island Unit 2 TS Technical Specifications 0 4 Beaver Valley 2 SSER 4 2 Appendix D
APPENDIX E NRC STAFF CONTRIBUTORS AND CONSULTANT Principal staff reviewers and the consultants who contributed to this supplement are: Staff Reviewer Title Review Branch G. Bagchi Section Leader Engineering
- H. L. Brammer Senior Mechanical Engineer Engineering
- N. Dudley Lead Reactor Engineer (Examiner) Reactor Projects No. 1**
B. J. Elliot Materials Engineer Engineering
- R. Goel Mechanical Engineer Plant Systems
- S. Hou Senior Mechanical Engineer Engineering
- J. G. Hunter Reactor Engineer Reactor Operations **
D. Lasher Electrical Engineer Electrical Instrumentation and Control Systems
- J. Lazevnick Electrical Engineer Electrical Instrumentation and Control Systems
- S. S. Lee Materials Engineer Engineering
- W. T. Lefave Senior Mechanical Engineer Plant Systems
- J. Mauck Electrical Engineer Electrical Instrumentation and Control Systems
- N. Romney Structural Engineer Engineering
- E. J. Sullivan, Jr. Section Leader Engineering
- Editor R. Sanders Editor Policy and Publications Managementti j Administrative
{ D. Miller Licensing Assistant PWR Project Directorate No. 2*
- Division of PWR Licensing A, Office of Nuclear Reactor Regulation.
** Division of Reactor Safety, Region I ttDivision of Publications Services, Office of Administration. l Beaver Valley 2 SSER 4 1 Appendix E
CONSULTANTS W. P. Chen, Energy Technology Engineering Center (ETEC) F. G. Farmer,~EG&G Idaho, Inc. R. Gamble, Novetech R. Harris, EG&G Idaho, Inc. l_ J. Singh, EG&G Idaho, Inc. . G. Thinnes, EG&G Idaho, Inc. 1 i i 4 i a N l i l l
-Beaver Valley 2 SSER 4 2 ' Appendix E
APPENDIX L* STAFF SAFETY EVALUATION REPORT ON CONFORMANCE TO GENERIC LETTER 83-28, ITEM 2.1 (PART 2), EQUIPMENT CLASSIFICATION (REACTOR TRIP SYSTEM COMPONENTS) AND VENDOR INTERFACE INTRODUCTION AND
SUMMARY
On February 25, 1983, both scram circuit breakers at Unit 1 of the Salem Nuclear Generating Station (Salem) failed to open upon an automatic reactor trip signal from the reactor prctection system. This incident was terminated manually by the operator about 30 seconds after the initiation of the automatic trip signal. The failure of the circuit breakers was determined to be related to the sticking of the undervoltage tr o attachment. Before this incident, on February 22, 1983, i at Salem Unit 1, an automatic trip signal was generated based on steam generator low-low level during plant startup. In this case, the reactor was tripped manually by the operator almost coincidentally with the automatic trip. Following these incidenti, on February 28, 1983, the NRC Executive Director for Operations (ED0) directed the staff to investigate and report on the generic implications of these occurrences at Salem Unit 1. The results of the staff's inquiry into the generic implications of the Salem Unit 1 incidents are reported in NUREG-1000 " Generic Implications of the ATWS Events at the Salem Nuclear Power Plant.' As a result of this investigation, the Commission (NRC) requested (by Generic Letter 83-28 dated July 8, 1983) all licensees of operating reactors, applicants for operating licenses, and holders of construc-tion permits to respond to generic issues raised by the analyses of these two anticipated-transient without-scram (ATWS) events. This report evaluates the response submitted by Duquesne Light Company, the applicant for Beaver Valley Power Station Unit 2, for Item 2.1 (Part 1) of Generic Letter 83-28. The actual documents reviewed as part of this evaluation are listed in the references. Item 2.1 (Part 2) requires the licensee / applicant to confirm that a contact has been established with the nuclear steam supply system (NSSS) vendor or with the vendors of each of the components of the reactor trip system which includes: periodic communication between the licensee / applicant and the NSSS vendor or the vendors of each of the components of the reactor trip system, and a system of positive feedback which confirms receipt by the licensee / applicant of transmittals of vendor technical information.
- Continued frua SSER 2 Beaver Valley 2 SSER 4 1 Appendix L
EVALUATION The applicant for Beaver Valley 2, provided its response to Item 2.1 (Part 2) of Generic Letter 83-28 on March 30, 1984. In that response, the applicant confirmed that the nuclear steam supply system for Beaver Valley 2 is a Westinghouse design and that the. reactor trip system (RTS) for Beaver Valley 2 is included as a part of the Westinghouse interface program established for the Beaver Valley 2 NSSS. The Westinghouse interface program for the NSSS includes both periodic communi-cation between Westinghouse and licensees / applicants and positive feedback from licensees / applicants in the form of signed receipts for technical information transmitted by Westinghouse. EG&G Idaho, Inc. provided technical assistance in this review. CONCLUSION The staff finds the applicant's confirming statement that Beaver Valley 2 is a participant in the Westinghouse interface program for the RTS meets the staff position on Item 2.1 (Part 2) of Generic Letter 83-28 and is, therefore, acceptable. In addition, since the applicant is already taking part in the Westinghouse program, there is no need for the staff to track any future imple-mentation of this issue. REFERENCES NRC letter, D. G. Eisenhut to all licensees of operating reactors, applicants for operating license, and holders of construction permits, " Required Actions Based on Generic Implications of Salem ATWS Events (Generic Letter 83-28)," July 8, 1983. Duquesne Light Company letter, E. J. Woolever to D. G. Eisenhut, Director, ; Division of Licensing, NRC, " Response to Generic Letter 83-28," March 30,1984. , i l l Beaver Valley 2 SSER 4 2 Appendix L
APPENDIX N STAFF SAFETY EVALUATION REPORT ON CONFORMANCE TO GENERIC LETTER 83-28, ITEM 4.5.2, REACTOR TRIP SYSTEM RELIABILITY--ON-LINE TESTING INTRODUCTION AND
SUMMARY
On February 25, 1983, both scram circuit breakers at Unit 1 of the Salem Nuclear Generating Station (Salem) failed to open upon an automatic reactor trip signal from the reactor protection system. This incident was terminated manually by the operator about 30 seconds after the initiation of the automatic trip signal. The failure of the circuit breakers was determined to be related to the sticking of the undervoltage trip attachment. Before this incident, on February 22, 1983, at Salem Unit 1, an automatic trip signal was generated based on steam generator low-low level during plant startup. In this case, the reactor was tripped manually by the operator almost coincidentally with the automatic trip. Following these incidents, on February 28, 1983, the NRC Executive Director for Operations (EDO) directed the staff to investigate and report on the generic implications of these occurrences at Salem Unit 1. The results of the staff's inquiry into the generic implications of the Salem Unit 1 incidents are reported in NUREG-1000, " Generic Implications of the ATWS Events at the Salem Nuclear Power Plant." As a result of this investigation, the Commission (NRC) requested (by Generic Letter 83-28 dated July 8,1983) all licensees of operating reac-tors, applicants for an operating license, and holders of construction permits to respond to generic issues raised by the analyses of these two anticipated-transient-without scram (ATWS) events. This report evaluates the response submitted by Duquesne Light Company, the applicant for Beaver Valley Unit 2, for Item 4.5.2 of Generic Letter 83-28. The actual documents reviewed as part of this evaluation are listed in the references. Item 4.5 states a staff position which requires on-line functional testing of the reactor trip system, including independent testing of the diverse trip ! features of the reactor trip breakers, for all plants. Item 4.5.2 requires applicants and licensees with plants not currently designed to permit this periodic on-line testing to justify not making modifications to permit such testing. By letter dated March 30, 1984, the applicant, Duquesne Light Company, responded to the staff position regarding Item 4.5.2 of Generic Letter 83-28. EVALUATION The applicant provided its response to Item 4.5.2 of Generic Letter 83-28 on March 30, 1984. In that response, the applicant stated that on-line testing of the reactor trip system (RTS), with the exception of on-line testing of the reactor trip bypass breakers, will be performed. Beaver Valley 2 SSER 4 1 Appendix N
The applicant stated that on-line testing of the bypass breakers during power operation is not justified because only one bypass breaker can be in service at a time (for less than 2 hours per month), and that when a bypass breaker is in service the RTS will initiate a trip signal to one of the trip breakers which is tested bimonthly. Also, the addition of components necessary to eliminate the need to lift leads and install jumpers (currently required to test the bypass breakers) could decrease the reliability of the bypass breaker system. EG&G Idaho, Inc. provided technical assistance in this review. CONCLUSION The staff finds the applicant's statement of the extent to which the applicant will perform on-line testing of the RTS meets the staff position on Item 4.5.2 of Generic Letter 83-28 and is, therefore, acceptable. The staff also finds the applicant's justification for not performing on-line testing of the bypass breakers sufficient and acceptable. In addition, on the basis of the commit-ment made to perform RTS testing, there is no need for the staff to track any future implementation. REFERENCES NRC letter, D. G. Eisenhut to all licensees of operating reactors, applicants for operating license, and holders of construction permits, " Required Actions Based on Generic Implications of Salem ATWS Events (Generic Letter 83-28)," July 8, 1983. Letter, E. J. Woolever, to D. G. Eisenhut, Director, Division of Licensing, NRC, " Response to Generic Letter 83-28," March 30,1984. l Beaver Valley 2 SSER 4 2 Appendix N
APPENDIX 0 STAFF SAFETY EVALUATION REPORT ON CONFORMANCE TO GENERIC LETTER 83-28, ITEM 3.1, POST-MAINTENANCE TESTING, REACTOR TRIP SYSTEM COMPONENTS, ITEM 3.2, POST-MAINTENANCE TESTING, ALL OTHER COMPONENTS, AND ITEM 4.5.1, REACTOR TRIP SYSTEM RELIABILITY, SYSTEM FUNCTIONAL TESTING INTRODUCTION AND
SUMMARY
On February 25, 1983, both scram circuit breakers at Unit 1 of the Salem Nuclear Generating Station (Salem) failed to open upon an automatic reactor trip signal from the reactor protection system. This incident was terminated manually by the operator about 30 seconds after the initiation of the automatic trip signal. The failure of the circuit breakers was determined to be related to the sticking of the undervoltage trip attachment. Before this incident, on February 22, 1983, at Salem Unit 1, an automatic trip signal was generated based on steam generator low-low level during plant startup. In this case, the reactor was tripped manually by the operator almost coincidentally with the automatic trip. Following these incidents, on February 28, 1983, the NRC Executive Director for Operations (ED0) directed the staff to investigate and report on the j generic implications of these occurrences at Salem Unit 1. The results of the ! staff's inquiry into the generic implications of the Salem Unit 1 incidents ! are reported in NUREG-1000, " Generic Implications of the ATWS Events at the l Salem Nuclear Power Plant." As a result of this investigation, the Commission ! (NRC) requested (by Generic Letter 83-28 dated July 8, 1983) all licensees of operating reactors, applicants for an operating license, and holders of construction permits to respond to generic issues raised by the analyses of these two anticipated transient without scram (ATWS) events. This safety evaluation addresses the issues listed above in the title of this appendix. By letter dated March 30, 1984, Duquesne Light Company (DLC or the applicant) described planned actions regarding the above items for Beaver Valley Power Station Unit 2 (BVPS-2). DISCUSSION AND EVALUATION The applicant's requirements for post maintenance operability testing are found in the Beaver Valley Power Station Operating Manual 1/2, Chapter 48, and Site Administrative P ocedure 3D, "The Maintenance Work Request (MWR)," used to control the connuct of maintenance. The procedures require the Operations Group to review the maintenance requests before the start of work and after the completion of work to determine what post-maintenance operability test-ing is required and to verify the satisfactory completion of testing before returning the equipment to service. The post-maintenance operability testing shall be performed using approved written procedures when required to demon-strate operability as defined in the Technical Specifications. Beaver Valley 2 SSER 4 1 Appendix 0
Since all safety-related work is required to be performed using an MWR, there l is assurance that post-maintenance testing will be performed and then reviewed by the Operations Department to verify that each component will perform its intended safety function before being returned to service. The applicant is incorporating vendor and engineering recommendations for safety-related equipment (including the RTS components) into the procedures and Technical Specifications as they are written. The information used in developing the procedures is stated in the reference section of the procedure. Startup Manual Chapter 2.2, " Regulatory / Technical Information Management," provides instructions for processing regulatory and technical information documents to ensure distribution, review and evaluation to determine applicability for incorporation into affected procedures. The applicant, during the operations phase, will implement SAP-24, " Correspondence Control," and various subtier procedures to provide controls for the distribution, evaluation and followup of documents containing safety-related vendor, industry and site generated information. Procedures will also be reviewed every two years to ensure that the appropriate vendor technical and technical specification test guidance has been incorporated. On the basis of the above, the staff concludes that the applicant's actions are consistent with the NRC staff positions for Items 3.1.1, 3.1.2, 3.2.1, and 3.2.2 of Generic Letter 83-28. The applicant has committed to perform bi-monthly on-line functional testing of the reactor protection logic systems including independent testing of the diverse trip features. The applicant will also perform response time testing of each reactor trip breaker at a refueling frequency for actuation: (1) with the undervoltage coil and shunt coil in service; (2) with only the undervoltage coil in service; and (3) with only the shunt coil in service. On this basis, the staff concludes that the applicant's actions are consistent with the NRC staff position for Item 4.5.1 of Generic Letter 83-28. CONCLUSION The applicant's actions and commitments are consistent with the staff's i positions stated for Items 3.1, 3.2, and 4.5.1. These items are considered l closed. In addition, on the basis that procedures to implement the subject requirements are already in place, there is no need for the staff to track any future implementation. REFERENCES NRC letter, D. G. Eisenhut to all licensees of operating reactors, applicants for operating license, and holders of construction permits, " Required Actions Based on Generic Implications of Salem ATWS Events (Generic Letter 83-28)," July 8, 1983. Duquesne Light Company letter, E. J. Woolever to D. G. Eisenhut, Director, Division of Licensing, NRC, " Response to Generic Letter 83-28," March 30,1984. Beaver Valley Unit 2 Final Safety Analysis Report. Beaver Valley 2 SSER 4 2 Appendix 0
Beaver Valley Power Station 1/2 Operating Manual, Chapter 48, Issue 3, Revision 0. SAP-3D, The Maintenance Work Request, Revision 2. SUM-2.2, Regulatory / Technical Information Management, Revision 1. SAP-2.4, Correspondence Control, Revision 0. LCP-10, Updated FSAR, Revision 4. LCP-9, Technical Specification Change, Revision 1. 4 LCP-2, Preparation of Responses to NRC Correspondence, Revision 2. TAG-6.0, Review of INP0 Data and Vendor Technical Bulletins, Revision 7. 2MSP-1.04-1, Reactor Protection Logic System Train "A" Bi-Monthly Test, Revision 0. 2MSP-1.05-1, Reactor Protection Logic System Train "B" Bi-Monthly Test, Revision 0. MSP-1.05, Solid State Protection System Train "B" Bi-Monthly Test, Revision 31.
- MSP-1.04A, Reactor Trip Breaker A (RTA) Bi-Monthly Test, Revision 1.
l i l
*These documents are located at the site and are not in the public domain.
The staff has ready access to them. Beaver Valley 2 SSER 4 3 Appendix 0
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APPENDIX P J 1 ERRATA Errors in SER: 1 Page Location Comment } 1-17 Open issue 4 SER sections included under this issue should be 5.2.4.1, 5.2.4.3, 5.4.2.2, I and 6.6. Error has been corrected in SSER 4. 1-19 Confirmatory SER Section 2.5.4.3.3.should be added. issue 2 Error has been corrected in SSER 4. 3-42 First full paragraph Change "IEEE 323-1974" to "IEEE 323-1971". 3
; 5-11 First paragraph, Should be " Unidentified leakage is also
] ' first.and second detected by a containment airborne gas sentences and particulate radioactive monitor. , This monitor responds to the increase in airborne radioactivity resulting j from leakage." [ 5-11 Second paragraph, After the words " condenser air ejector , l second sentence vent line" add "and steam generator ; j blowdown lines". ' ! 5-19 Second paragraph, Change " Specification ND5-0064 i first sentence (Revision 0)" to " Specification 10080-DMS-002 (Revision 2)". 5-19 Second paragraph, Add "(Revi.aion 2)" after " Specification second sentence 10080-DMS-002". 6-6 Fifth paragraph Change " SATAN-VI computer program" to
" SATAN-V computer program".
6-7 Fifth paragraph, Change "360 in.2" to "320 in.2u, fourth sentence 6-7 Sixth paragraph, Change "12.9 psid" to "15.08 psid". { third sentence ' 6-8 First paragraph, Change "18.07 psid" to "5.38 psid". first sentence i l Beaver Valley 2 SSER 4 1 Appendix P q l
6-10 Eighth line from Change "with" to "within". bottom of page 6-21 Third paragraph, After the words "two check valves" add second sentence "or three check valves". 7-13 Item (8) in its Should be " service water system pump entirety isolation (a) containment isolation phase A". 8-6 Last paragraph, Should read "using actual or simulated second sentence, load". fourth sentence 8-12 Section 8.3.3.1.2, Change " lighting" to lightning". second paragraph, fourth sentence
; 9-12 Last paragraph, Change " Category I" to " Category I_I".
fifth sentence 9-31 Second paragraph, Change "RG 152" to "RG 1.52". first sentence 9-46 Fourth paragraph, Should read "The turbine building fourth sentence south exterior wall adjacent to the transformers is 2-hour rated". . 9-55 Section 9.5.1.8 Item 3 should be deleted. Insert the following statement: "In addition, cable spreading room fire protection is a backfit issue, as described in Section 9.5.1.6." 9-67 Fourth sentence Should be "...(i.e., greater than on page 24 hours)". i 9-72 Last sentence of Should be deleted. I last paragraph 10-5 Fifth sentence Should read "close on low pressure on page signals in any steamline,". 10-16 Fourth sentence Delete the phrase " motor-operated isolation from bottom of page valve". 10-17 Second full paragraph, Change " Motor-operated" to " Pneumatic sixth sentence hydraulic-operated" i Beaver Valley 2 SSER 4 2 Appendix P
10-17 Second to last "in the PPDWST, are seismic Category sentence of second I". Should be supplemented to read, full paragraph "in the PPDWST, are seismic Category I, except connection which is associated with a non-safety-related transmitter. This instrument and tubing and valves are designed as seismic Category II and will not fail in a manner that will cause a loss of pressure boundary in the event of an SSE." 10-17 Second full paragraph, Change " Category I" to " Category II". last sentence 10-18* Third paragraph, "Although" should be deleted. The third sentence sentence should read, "The separate cubicle enclosure...from the turbine driven pump; by letter dated August 12, 1984, the applicant has provided the results of...." 10-18 Sixth sentence, Change "an air piston-operated valve fourth paragraph that opens" to " solenoid valves that open". 10-18 Eighth sentence, Sentence should be deleted and replaced fourth paragraph with: "The solenoid valves are powered from the emergency busses and fail in the open position on loss of power to the valves." 15-4 Third complete Should read "The safety injection system sentence (SIS) injects borated water (2000 ppm) from the refueling water storage tank into the...." Errors in SSER 2: P_ age Location Comment 4-3 Paragraph after "(3)" should not be there at all; the item d paragraph is an independent one and is not part 3 of anything. Errors in SSER 3: _Pjige Location Comment Appendix D, 7th abbreviation EAP represents " Engineering page 1 Assurance Program". Beaver Valley 2 SSER 4 3 Appendix P
APPENDIX Q BEAVER VALLEY-2 SQRT REPORT l l f Beaver Valley 2 SSER 4 Appendix Q
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4 EGG-EA-7498 i i i 3 BEAVER VALLEY-2 SQRT-REPORT
- J. N. Singh j B. L. Harris
! G. L. Thinnes 1 ; Published December 1986 1
i ) i j EG&G Idaho, Inc' . j Idaho Falls, Idaho 83415 l l Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Under DOE Contract No. DE-AC07-76ID01570 FIN No. A6415 l
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ABSTRACT EG&G Idaho is assisting the Nuclear Regulatory Commission in evaluating Duquesne Light Company's program for the dynamic qualification of safety related electrical and mechanical equipment for the Beaver Valley Electric Generating Plant, Unit 2. Applicants are required to use test or analysis or a combination of both to qualify equipment, such that its safety function will be ensured during and after the dynamic event, and provide documentation. The review, when completed, will indicate whether an appropriate qualification program has been defined and implemented for seismic Category I mechanical and electrical equipment which will provide reasonable assurance that such equipment will function properly during and after the excitation due to vibratory forces of the dynamic event. 1 Beaver Valley 2 SSER 4 ii Appendix Q
. l.
s ., SU R RY A seismic qualification review team (SQRT) consisting of engineers from the Engineering Branch PWR-A of the Nuclear Regulatory Commission and ' Idaho National Engineering Laboratory made a site visit to ,the Beaver Valley Electrical Generating Plant, Unit'2 of Duquesne Light- Company s located de Shippingport, Pennsylvania; They observed the field i installation and reviewed the qualification report; for twenty-four , selected pieces of seismic Category I electrical ar.d mechanical equipmenti and their supporting structures. Four generic and three equipment specif3: concerns were identified for which additional information fis needed in order for the SQRT to complete'the review. *lhese are referred to as open items. Although the open items need to be resolved, the review indicated ~ that the equipment was adequately qualified for the dynamic environment at 5 Beaver. Valley. 1
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Beaver Valley 2 SSER 4 ' iii Appendix Q. m_
2 i CONTENTS t ABSTRACT,.............................................................. ii
SUMMARY
............................................................... iii
- 1. INTRODUCTION ..................................................... 1
- 2. NUCLEAR STEAM SUPPLY SYSTEM (NSSS) EQUIPMENT ..................... 2 2.1 Steam Generator Flow Transmitter.(NSSS-2) .................. 2 2.2 Reactor Vessel Letdown Modulating Valve (NSSS-3) ........... 2 4
2.3 Motor Operated Gate Valve (NSSS-4) ......................... 3 2.4 Residual Heat Removal Heat Exchanger (NSSS-8) .............. 5 Reactor Coolant Pump and Seals (NSSS-11) ................... 2.5 5 2.6 Reactor Vessel Letdown Isolation Valve-8035A (NSSS-12) ..... 7 2.7 Low Head Safety Injection Pump (NSSS-13) ................... 9 2.8 Plant Safety Monitoring System Cabinet (NSSS-14) ........... 10 2.9 Loop Stop Valve Protection Cabinet (NSSS 15) ............... 11 2.10 8 In. Motor Operated Gate Valve (NSSS-16) .................. 12 2.11 Nuclear Instrumentation System Cabinet (NSSS-19) ........... 14 2.12 Centrifugal Charging / Safety Injection Pump (NSSS-20) . ... .. . 14 g 2.13 Position Transmitter - 7300 Printed Circuit Card (NSSS-21) ............................................. 16 ,
- 3. BALANCE OF PLANT (B0P) EQUIPMENT ................................. 17 3.1 Motor Operated Butterfly Valve (BCP-1) . . . . . . . . . . . . . . . . . . . . . 17
\ 3.2 Feedwater Isolation Valve (80P-5) .......................... 17 3.3 Alternate Shutdown Panel (B0P-8) ........................... 18 ; 3.4 Central Station A/C Unit (B0P-9) ........................... 19 3.5 Vital Bus Distribution Panel Board (B0P-14) ................ 20 . 3.6 Air Operated Control Valve (B0P-15) ........................ 21 3.7 Vital Bus Uninterruptable Power Supply System (B0P-17) ..... 22 3.8 Quench Spray Chemical Injection Pump (B0P-19) .............. 23 3.9 Service Water Pump (B0P-20) ................................ 23 3.10 Motor Operated Damper (80P-23) ............................. 24 i ,
3.11 Fuel Oil Pressure Switch (B0P-24) .......................... 25
- 4. FINDINGS AND CONCLUSIONS ......................................... 26 4.1 Generic Issues ............................................. 26 4.2 Equipment Specific Issues .................................. 26 4.3 Conclusion ................................................. 27 4
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, Beaver Valley 2 SSER 4 iv Appendix Q i
A k s
y-lh 4 g, , t BEAVER VALLEY-2 SQRT REPORTL
- 1. INTRODUCTION
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' The Engineering Branch (EB) of the Nuclear Regulatory Commission (NRC) has the lead responsibility in reviewing and evaluating the dynamic qualification of safety related mechanical and electrical equipment. This equipment may be subjected to vibration from earthquakes and/or hydrodynamic-forces. Applicants are required to use' test or analysis or a combination of both to qualify equipment essential to plant safety, such that its function will be ensured during and after the dynamic event.
These pieces of coaipment and how they meet the required criteria are described by tne applicant in a Final-Safety Analysis Report (FSAR). On complation cf the FSAR review, evaluation and approval, the applicant receives an Operating License (OL) for commercial plant operation. A Seismic Qualification Raview Team (SQRT) consisting of engineers from the EB cf-NRC.and Idaho National Engineering Laboratory (INEL), made a site visit to the Beaver Valley Electric Generating Plant, Unit 2 of '. Duquesne Light Company, Shippingport, Pennsylvania, from September 30 through October 3, 1986. The purpose of the visit was to observe the field installation, review the equipment qualification methods, procedures (including modeling technique and adequacy), and documented results for a list of selected seismic Category I mechanical 'and electrical equipment and their supporting structures. This report, containing the review findings, indicates which of the items are qualified and require no additional documentation. It also identifies some equipment and certain general concerns to complete forthe which additional information is needed in order for the SQRT review. These are referred to as-open items. The applicant is to further investigate and provide additional documentation to resolve these issues. l l 6 w s Beaver Valley 2 SSER 4 1 > Appendix Q
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- 2. NUCLEAR STEAM SUPPLY SYSTEM (NSSS) EQUIPMENT 2.1 Steam Generator Flow Transmitter (NSSS-2)
This transmitter (tag No. 2MS9 FT475) was an ITT Barton transmitter, model Number 764. It is located in containment at the 772 ft level and measures steam generator flow rate in the main steam system. The transmitter and flow manifold are mounted by its support brackets to a structural support anchored to a wall. The transmitter, manifold, and bracket were seismically qualified by test at Westinghouse and documented in the report " Equipment Qualification Test Report--Barton Differential Pressure Transmitters--Group A," WCAP 8687, Supplement 2-E03A, Rev. 2, March 1983. Further detail on qualification is described in " Equipment Qualification Data Package--Differential Pressure Transmitters: Qualification Group A," WCAP 8587, EQDP-ESE-3A, Rev. 4, March 1983. A multiaxis sine sweep test was performed for the resonance search from 1-33 Hz. The qualification test was performed by pseudo triaxial sine beat tests which enveloped a 10% margin on the RRS. Five OBE level and twelve SSE level tests were performed demonstrating structural integrity and operability of the transmitter. The reviewer questioned the rigidness of the structural support of the transmitter and manifold but calculations by Duquesne Light indicated the support to be well within the rigid range. Based on the observation of the field installation, review of the qualification documents and the applicant's res'ponse to questions, this item is seismically qualified. 2.2 Peactor Vessel Letdown Modulating Valve (NSSS-3) The reactor vessel letdown modulating valve was supplied by Target Rock (Model 79AB-003). It is located inside the containment at the 742 ft elevation and is used to vent hydrogen from the reactor vessel head and to modulate the letdown to the pressure relief tank. . 1 The valve was qualified by a combination of test and analysis. Thermal, mechanical, vibration, and radiation aging were performed as discussed in Westinghouse report WCAP-8687, Supp. H10C, Rev. 1, Equipment Qualification Test Report Target Rock Modulating Valve (Environmental and
~
Seismic Testing), dated January 1985. The valve was mounted to an electrodynamic test table using a test fixture that clamped around the pipe stubs near the valve body. The valve was line mounted in the plant with pipe supports some distance from the valve. The acceleration levels used in the testing were 3.2 g and 4.0 g in each direction simultaneously applied for OBE and SSE, respectively. These levels are much larger than the maximum accelerations from the piping analysis, 0.39 g for OBE and 0.642 g for SSE. Therefore the difference between field and test mounting was considered satisfactory. The valve was pressurized before testing was begun. A 0.2 g sine sweep resonance search was performed and the lowest natural frequency of the valve was found to be 40 Hz. Qualification Beaver Valley 2 SSER 4 2 Appendix Q
testing consisted of four OBE and two SSE biaxial sine dwell tests at a single frequency of either the natural frequency or 50 Hz, whichever was lower. During each dwell the valve was operated. During one test a wire came loose. This was attributed to excessive handling. The test was rerun after replacement of the wire and the test was successful. After the seismic testing the valve was subjected to a body hydrostatic test, a disc hydrostatic and seal leakage test and a seal weld leak test. The seal leakage was excessive after the seismic testing. The valve was disassembled and examined to determine the cause of the excessive leakage. The seat leakage was attributed to a buildup of scale around the seat area caused by corrosive material in the test apparatus. The scale was removed and the valve was reassembled. The seat leakage was reduced by a factor of 10 to 65 cc in 10 minutes, which was judged acceptable. The valve was also subjected to a high energy line break environment. The valve did not function due to an unsealed electrical connection. An external seal was then added and the valve functioned as required. Stresses in the valve body were found in Target Rock report No. 2586, Seismic Analysis Report for the Target Rock Project 79Z-59, dated July 18, 1984. The lowest natural frequency of the valve was calculated as 121 Hz. A static analysis of the valve was performed using accelerations of 3.2 g and 4.0 g applied simultaneously in all three directions for OBE and SSE, respectively. Deadweight, pressure, and loads from the extended structure were included in the stress calculations. All stresses were below the ASME code allowable stresses. The qualified life of the valve was found to be 9.5 years. The whole valve is to be replaced at that time. Based on the observation of the field installation, review of the qualification documents, and the applicant's response to questions, the reactor vessel letdown modulating valve is adequately qualified for the prescribed loads. 2.3 Motor Operated Gate Valve (NSSS-4) The PORV 3-in. block valve (tag No. 2RCS*MOV536) is a motor operated gate valve in the pressurizer cubicle built by Westinghouse as model Number 03003GM99FNH00G W750010. The valve has a Limitorque operator with ID Number B77Q2439M-VK. The pipe is supported horizontally by a snubber and vertically by a spring hanger next to the valve. The valve was seismically qualified by a combination of testing and analysis. The valve body was qualified by an axisymmetric finite analysis performed by Westinghouse and documented in the Westinghouse report,
" Seismic Analysis Report for Westinghouse Class I Nuclear Valves,"
EM-5689. All natural frequencies of the body were above 163 Hz; therefore a static analysis was performed using 4.5 g loading in three directions. Stresses were highest in the body bonnet bolts: 53,582 psi compared to an ASME code allowable of 53,648 psi. Deflection calculated in the analysis of the bonnet-stem was 0.002 in. compared to an allowable deflection of 0.013 in. Beaver Valley 2 S56R 4 3 Appendix Q
The actuator was qualified by testing performed at Westinghouse. Sine sweep resonance search tests indicated no natural frequencies below 35 Hz. Then single axis sine beat tests were performed in each of three directions using 5 g input at the OBE level and 7.5 g input at the SSE level. Five OBE level and one SSE level tests were performed. This testing was documented in Westinghouse report: WCAP 8687, Supp. 2, H01A. Operability' of the valve was verified by static deflection tests reported in Westinghouse-WEMD report Number 4995. As noted in the walk-down, the operator seemed quite large and was unsupported. For this reason a check of the piping analysis input parameters with respect.to the valve's C.G. location and weight was investigated in the audit. It was found that the rotational inertia of the' valve was approximately 20% below the actual characteristics of the valve as shown on the Westinghouse drawing. Duquesne Light's response was that an ongoing piping audit program, as described in report Number 2BVM-156, Rev. 3, would illuminate and reconcile this discrepancy. According to Duquesne Light, this line had not undergone that audit. This pipe stress reconciliation program uses the following criteria: (a)' pipe and valve parameter discrepancies which are more then 20% in error are evaluated for their contributing effects on the piping system of concern; (b) if the discrepancies are significant then a reanalysis is performed. In addition to the valve and operator, the motor control cabinet which supplies the power for the actuator and a cabinet containing Agastat and ASEA switching relays for the valve were inspected. The switching relay cabinet (tag No. PNL*REL-269) is a single bay cabinet qualified by test and documented in Acton Environmental Testing's report No. 20046-84N-2. The relays were monitored during the test and met the chatter criteria of 2 msec maximum discontinuities. The prime area of concern, however, was that the cabinet was mounted to the floor right next to a similar cabinet. Seismic excitation would cause the cabinets to impact causing high frequency excitations to be-t imparted to the relays which could malfunction since the qualification tests were concerned with frequencies below 33 Hz only. Duquesne Light said that a walkdown of electrical systems as outlined in project procedure BVM236 was in progress and thet this' system had not been inspected yet. The walkdown would identify ini.erfo<ences and then appropriate action would be taken to maintain equipment seismic qualification. As a result of all these findings, the valve and actuator are
, seisimically qualified but questions of a generic nature arise with respect to: (a) reconciliation of piping analysis parameters with actual valve characteristics and (b) determining interferences between electrical components which might affect their qualification. The programs outlined by Duquesne Light seem adequate to address these problems. Completion of these programs must be verified for seismic qualification of the effected plant components.
Beaver Valley 2 SSER 4 4 Appendix Q
Based on the observation of the field installation, review of the
- qualification documents and applicant's responses to our inquiries the-motor operated gate valve is qualified pending resolution and confirmation i of the completion of the programs.
2.4 Residual Heat Removal Heat Exchanger (NSSS-8) ! The residual heat removal (RHR) heat exchanger (tag No. 2RHS*E21A) was manufactured by Atlas Industrial Manufacturing Co. with Serial No. 3483. It stands vertically at the 709 ft level, is approximately 30 ft high, and
- is supported by lugs at the bottom and two trusses in the horizontal direction at the top. The heat exchanger was qualified by analysis.
1 i The qualification analysis was performed by Dynatech R/D Company and l is documented in report No. 1318, Rev. 2. Natural frequencies were calculated using beam theory and showed natural frequencies to be above 28.8 Hz. This is in the rigid range of the Beaver Valley spectrum so that a static analysis was performed for the seismic conditions. The computer codes, N0ZZLE, REACT, and REACT 1 were used to calculate nozzle loads on the heat exchanger. Stress conditions around the nozzles and the support j lugs were determined by the Bijlaard method described in the Welding Research Council Bulletin 107. Stress criteria for.the analysis was based on Section VIII, Div. 1 of the ASME Boiler and Pressure Vessel Code. Stresses calculated around the 24 in. nozzles were 41,060 psi while i the allowable was 42,000 psi. The Bijlaard method which is based upon 1 parametric curves derived from tests of nozzle-shell configurations is- ! predicated upon the nozzles being well away from shell discontinuities such { ' as the shell flanges and shell heads as found on the RHR heat exchanger. Since stresses are~so high and the stress calculation method questionable for these nozzles the vendor was asked to further justify the calculation. Inspection of the drawings indicated no weld sizes were dimensioned i for the lugs supporting the vessel vertically. Furthermore, no weld sizing l 4 calculations were found in the report. A Bijlaard analysis was also l performed for the lug-shell junction but the lug dimensions used seem l j- totally inappropriate for such an analysis. These two areas of the analysis must also be resolved. I Based on the observation of the field installation, review of the qualification documentation and applicant's responses to questions this item is qualified pending satisfactory resolution of the following three concerns: (a) the appropriateness of using Bijlaard analysis for the 24-in. nozzle-shell junction stresses; (b) sizing calculations of the support lug welds; and (c) the usage of specified lug dimensions in the support lug-shell junction stress analysis. 2.5 Reactor Coolant Pump and Seals (NSSS-11) l The reactor coolant pump (RCP) assembly and~ seals (MPL No. 2RCS*P21A; pump: model Number W-11001-A1(93A) and motor: serial No. IS-86P389, type and frame CS-62-1/2-SPL) was supplied by Westinghouse Electro-Mech. Div. according to purchase order No. 546-GPT-167187-BN. The equipment l Beaver Valley 2 SSER 4 5 Appendix Q
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l specification number for Beaver Valley-2 is 952204, Rev. 5, dated . January 14, 1977; which is a part of general specification Number G-677188, , Rev. 4, dated January 15, 1976. This unit is located at the 732 ft elevation of the containment building. It is a part of the reactor coolant i system. The pump is only needed for power production. . In an accident situation'the safety requirements for the pump is to be able to coast down and then natural circulation takes over. It does not need to run for.
, accident mitigation.
1 The Pump: The pump assembly was mounted in the vertical position and i supported by three 4-1/2 in. A579 grade 72 bolts. The qualification of i this assembly is based on static and dynamic analyses of various components 4 of the assembly. Seismic loads are considered in the analyses. The j qualification package of the assembly consists of an overall summary report and several other component-analysis reports. A listing of these reports is given below. ~
- 1. 93A Reactor Coolant Pump Pressure Boundary Component-Summary Report for the Duquesne Light Co. , Beaver Valley Power Station #2, No. EM5129, Rev. 2, dated January 15, 1982.
- 2. Stress Analysis of the Casing, Main Flange, Main Flange Bolts, and Thermal Barrier of the 93A Shaft Seal Pump, No. EM4487, Rev.1,-dated August 16, 1974.
l 3. Dynamic Seismic Anlaysis of the Model 93A Reactor Coolant Pump with j Seal Maintenance System for Duquesne Light Co. , Beaver Valley No.2 j Power Station, No. E4990, Rev.1, dated January 18, 1982. j 4. Static LOCA Analysis of the Model 93A Reactor Coolant Pump with Seal i Maintenance System for the Ducluesne Light Co. , Beaver Valley Unit No.2 Power Station, No. EM500L, Rev. 1, dated January 18, 1982.
- 5. Structural Analysis of the Seal Housing, Ring Clamps and Bolts for the i
93A Controlled Leakage Pump 6 pool-less Design), No. EM4556, Rev. 1, j dated June 1975. 5
- 6. Stress Analysis of the 93A Controlled Leakage Pump Thermal Barrier Heat Exchanger, No. EM492, Rev. O, dated June 25, 1975.
l 7. Structural Analysis of the Weir Plate for the 93A Discharge Nozzle, l No. EM4542, Rev. 1, dated June 1974. \
- 8. Structural Analysis of the Suction and Discharge Nozzles of the
- Model 93A1/93A Reactor Coolant Pump Under Umbrella Loads, No. EM4962, i Rev. 1, dated August 3, 1977.
! 9. Stress Analysis of 93A Pump Nonpressure Components for the Duquesne Light Co. Beaver Valley Unit No. 2 Station, No. EM5017, Rev. 2, dated l May 23, 1985. i 10. Analysis of 93A Casing Feet Using Umbrella Loads, No. EM4503, Rev. 1,
- j. dated February 8, 1974.
i Beaver Valley 2 SSER 4 6 Appendix Q
All the above reports were prepared by Westinghouse (WEMD). The following computer codes were used (individually or in groups) in these analyses: SEAL SHELL-2; BARTON, WECAN, FEASS, ANSYS, WISEC, NUMBRA, LUG 1. The Motor: The motor supplied by the Large ~ Rotating Apparatus Division was'also vertically mounted. The general specification and supplementary ordering information for this induction motor for shaft seal type pump are contained in E-569700, Rev. K and E-565698, Rev. E, respectively. The seismic qualification of the motor is based on static analysis as documented in the report: Seismic and Loss of Coolant Accident Stress Analysis of Non-rotating Parts on Nuclear Reactor Coolant Pump Motor, Memo No. 759, Rev. O, dated February 15, 1976. This is a 2D elastic stress analysis using the QUAKER computer code. The Seals: According to the applicant the RCP seal is a film riding element which does not carry any seismic load. It is contained in the RCP seal housing which is seismically qualified. However, it should be noted that the pump seals at Beaver Valley Unit 2 have undergone generic and specific plant performance tests to demonstrate seal integrity. Additionally, the Westinghouse Owners Group has been involved in an extensive seal testing program over the last several years. It should also be noted that the integrity is not a concern during a LOCA since the seal leak would be virtually negligible when compared to the LOCA. The RCPs are not required to function during a LOCA. The Assembly: The static and dynamic analyses performed on the pump and the motor are adequate. The resulting stresses for the pump are, however, compared to faulted condition criteria. These criteria do not ensure the operability of the pump. Technically, this is unsatisfactory. However, the calculated stresses are quite low and well within the allowables of nonfaulted condition. On an equipment specific basis, therefore, the results are satisfactory. The calculated stresses for the ; motor are compared to allowables which ensure operability. Based upon observation of the field installation, review of qualification documentation and responses by the applicant to our questions, the unit is adequately qualified for Beaver Valley-2 application. 2.6 Reactor Vessel Letdown Isolation Valve-8035A (NSSS-12) The reactor vessel letdown isolation valve (model No. 79AB-001, MPL No. 2RCS*S0V200A) was supplied by Target Rock according to purchase order No. 413910'and equipment specification No. G-955186.' It was located at the 746 ft 10 in elevation of the containment building. The valve was' socket welded to a pipeline. This post TMI modification vents hydrogen from the reactor vessel head. It allows 50 gpm letdown to the pressure relief tank (PRT). It is a part of the reactor coolant system and.must maintain a pressure boundary. Beaver Valley 2 SSER 4 7 Appendix Q
This valve is seismically qualified based on a combination of analysis and tests performed on a similar unit. The field unit is 79AB-001 serial No. 470, whereas the specimen was 79AB-001 serial No. 55. The test portion of the qualification is described in the report: EQTR Target Rock Isolation Solenoid Valves, No. WCAP-8687, Supp. 2-H10A, Rev. 2, dated January 1, 1985, prepared by Westinghouse. Seismic Analysis Report for the Target Rock Project 79Z-59 1-in. Solenoid Operated Globe and Modulating Valves, No. 2586, Rev. O, dated April 16, 1980 has the analysis details. The laboratory mounting for the tests consisted of socket welding the valve to one inch Sch. 160 pipe on each end. This pipe (6 in. long) was in i turn bolted to the test table using mounting blocks. A resonance search with 0.2 g magnitude sine sweep at 1/2 octave per minute did not indicate any frequency below 50 Hz. Subsequently, a series of qualification tests with sine dwell were performed. They had phase incoherent biaxial inputs. The magnitudes at 50 Hz were as follows: s/s f/b v OBE 3.9 g 3.9 g 3.9 g SSE 4.9 g 4.9 g 4.9 g. The required piping accelerations for the valve interface were: s/s f/b v OBE 0.17 g 0.13 g 0.21 g SSE 0.31 g 0.24 g 0.39 g. The seismic loads were corbined with other loads by SRSS (square root of the sum of the squares) technique. There were four OBE followed by two SSE level tests performed. Operability was monitored. Visual inspection indicated that the top cover gasket developed a radial tear which had to be , repaired during the tests. It was determined to be caused by excessive l handling. This was confirmed by supplemental testing. It was stipulated, however, that if handling occurs in the field the gasket should be examined each time and replaced as necessary. It has been noted in the instruction manual. It was further discovered that the switch potting had shrunk. This was attributed to excessive thermal aging temperatures as supplemental aging at correct temperature showed no such anomaly. The analysis consisted of hand calculations for stresses in the internal and external bonnet walls and plunger deflections. They were calculated for seismic plus dead weight plus operating loads. The results were compared to the respective allowables. The stresses and deflections are acceptable. Operability was verified. Anomalies were satisfactorily resolved. The applied loads were greater than required. Beaver Valley 2 SSER 4 8 Appendix Q
Based upon our observation of the field installation, review of qualification document and responses provided by the applicant to questions, the unit is adequately qualified for 9.5 years. However, final system qualification reconciling as-built loads is pending. This reconciliation is not complete on a generic basis. 2.7 Low Head Safety Injection Pump (NSSS-13) The low head safety injection pump was supplied by Goulds Pumps (Model 3405L, size 10 x 12-170V) and the pump motor was supplied by Westinghouse Large Motor Division (Model 77F14111 with a 5008-S frame). The pump and motor were mounted to a bedplate which was bolted to the floor. The pump and motor are located in the safeguards area at the 720 ft 4 in elevation. The pump provides emergency core cooling system injection. The pump was qualified by analysis in Mcdonald Engineering Analysis Co. Report ME-468, Seismic-Stress Analysis of ASME Section III Class 2 Pumps Model 3405L, Size 10 x 12-170V, Manufactured by Goulds Pumps, Inc., dated March 15, 1978. A three-dimensional dynamic analysis was performed using the ICES-STRUDL II program. The pump, notor, and bedplate were included in the model. The first two natural frequencies were 45.2 Hz and 66.8 Hz. Seismic stresses were found for OBE and SSE using these two modes. The accelerations used for OBE and SSE are shown below. OBE: 1.0 g vertical and 1.5 g horizontal (each direction) SSE: 2.0 g vertical and 3.0 g horizontal (each direction) The seismic loads were applied separately in each of the three directions. Stresses from the worst case horizontal loading were then combined with the vertical loading by square root of the sum of squares. Nozzle loads and operating loads were also included. All OBE stresses were below the normal condition allowable stresses. All SSE stresses were below the faulted condition allowable stresses. The method of load combination was thought to be possibly nonconservative since the highest stresses may not be caused from the direction chosen for application of the acceleration. However, the accelerations used in this analysis had large margins over the required accelerations. The generic Westinghouse specification requires 2.1 g simultaneously applied in all three directions. This analysis was done using acceleratuns that were roughly equivalent to the generic specification. The required accelerations for the Beaver Valley plant were 0.2 g horizontal and 0.175 g vertical. Therefore, the load combination method used in the analysis of the pump was considered satisfactory. The load combination method was considered in the review of other pumps during the audit. No example of the use of the load combination method without large margins between applied accelerations and required accelerations was ! observed. The functional capability of the pump was shown by finding bearing loads, coupling misalignment due to bending in the pump and motor shaft, stresses in the impeller key and impeller clearance. The impeller was Beaver Valley 2 SSER 4 9 Appendix Q
found to contact the casing wear ring. The contact force was 122 lb or 54 lb/in. over the 2.25 in. wear ring width. This resulted in a stress of about 300 psi which is well below the allowable stress of 12,000 psi. The contact was predicted using an acceleration of 3 g. This acceleration is much greater than the required acceleration. Also, no hydrodynamic spring effect of the water in the pump was considered. This effect would tend to prevent contact. Since these two effects are conservative it was concluded by the reviewer that contact would probably not occur and if it did occur, the stresses would be very small. All other results of calculations mentioned above were satisfactory. Therefore, the pump was considered qualified for its application. The motor was qualified by a combination of test and analysis. Thermal and radiation aging of the stator were performed as discussed in Westinghouse report WCAP-8687, Supp. 2-A02A Environmental, Rev. 2 Equipment Qualification Test Report Westinghouse LMD Motor Insulation, March 1983. An electromagnetic shaker was used to excite the motor in the 5 to 33 Hz range. No natural frequencies were found. Bump tests were also performed to verify that no frequencies below 33 Hz existed. For qualification testing the motor was mounted rigidly to a test table at a 45 degree angle. Pseudo triaxial, multifrequency input was applied with all TRSs exceeding the RRSs for all tests except one SSE test. Five OBE tests and three SSE tests (one in each orientation) were performed. The damping was 2% for OBE and 5% for SSE. Operability of the motor was shown by analysis in Westinghouse Shop Order 77F14113, Seismic Analysis of Low Head Safety Injection Motors for Beaver Valley Power Station Unit No. 2 Duquesne Light Power Co., dated June 25, 1979. . Shaft and rotor deflections due to horizontal and vertical seismic loads, weights, and magnetic forces were found. Shear and bending deformations were included in the analysis. Bearing loads and shaft stresses were below the allowable values. The rotor deflection was 0.002 in. for SSE compared to a gap of 0.044 in. Therefore, the motor operability was demonstrated. Based on the observation of the field installation, review of the qualification documents, and the applicant's response to questions, the low head safety injection pump and motor are adequately qualified for the prescribed loads. 2.8 Plant Safety Monitoring System Cabinet (NSSS-14) The plant safety monitoring system (PSMS) Cabinet, Model 2D32338G01, was supplied by Westinghouse and is located in the control building at the 707 ft 6 in, elevation. The cabinet is part of the system that performs postaccident monitoring of reactor vessel level and core cooling thermocouples. The PSMS cabinet was qualified by test in Westinghouse Report WCAP-8687, Supp. 2E53A, Rev. O, Equipment Qualification Test Report Plant Safety Monitoring System (Seismic and Environmental Testing) dated March 1983. The cabinet was bolted to a rigid (4 in, plate) base with 8-3/4 in. bolts, two at each corner of the cabinet, during the tests. The Peaver Valley 2 SSER 4 10 Appendix Q
base was then welded to a biaxial test table. The 4 in. base was removed for the mounting in the Beaver Valley 2 plant to allow access beneath the cabinet. Westinghouse designed a mounting frame made from channel sections with a web thickness of 0.15 in. which Beaver Valley could use and not invalidate the testing. This design was discussed in Westinghouse Report EQ&T-EQA-837, Base Modification for Beaver Valley (DMW) PSMS Cabinets dated April 27, 1984. Beaver Valley, however, used a plate approximately 2 in. thick with the center part removed for mounting the cabinet. This plate appears to have approximately the same stiffness as the frame. Therefore, the field mounting was judged to be satisfactory. A 0.2 g resonance search was performed from 1 to 50 Hz at a rate of 1 octave / minute. Two resonances were observed: 14-15 Hz front /back and 12-13 Hz side / side. Multifrequency pseudo triaxial input was applied to the test table in the qualification testing. Five successful OBE tests and four successful SSE tests were performed. All TRSs exceeded the required RRSs. During one unsuccessful OBE test an electrical malfunction occurred due to frayed wires. The test was repeated successfully after repairing the wires. Some SSE tests to an acceleration level of about 15 g were unsuccessful but four SSE tests to a level of about 12 g were run. No structural damage occurred during the tests. The equipment functioned during both OBE and SSE testing. The PSMS cabinet is located in a mild environment area. The qualified life is presently being evaluated. An interim 5 year qualified life has been established until completion of the short term component aging program scheduled for completion in March 1987. The qualified life is discussed in Westinghouse letter ENG/1ES(86)-554, Position on 5-Year Q/L of Microprocessor Systems dated September 19, 1986. Based on the observation of the field installation, review of the qualification documents, and the applicant's response to questions, the PSMS cabinet is adequately qualified for the prescribed loads. 2.9 Loop Stop Valve Protection Cabinet (NSSS 15) The loop stop valve protection cabinet (Drawing No. 1060E29G01, MPL I No. RK*2VV-REL-A) was supplied by Westinghouse according to purchase order No. 206544 and equipment specification No. 953210. It was located on the 707 ft 6 in. elevation of the contial building. The field mounting consisted of welding it to the structural steel raised floor with 1/4-in. fillet welds 3 in. long at each corner. It is a part of the reactor protection system. It is an interlock system to prevent reopening of loop stop valves and reactor coolant pump protection. It nad only a temporary paper tag. The qualification of this equipment is based on a combination of tests and analyses. These tests and analyses are documented in the following reports:
- 1. Equipment Qualification Test Report Loop Stop Valve Cabinet, No. WCAP-8687, Supp. 2-E23A, Rev. O, dated October 1985 by Westinghouse.
Beaver Valley 2 SSER 4 11 Appendix Q
I
- 2. Equipment Qualification Data Package Loop Stop Valve Cabinet, I No. WCAP-8587, EQDP-ESE-23A, Supp. 2, Rev. O, dated October 1985 by Westinghouse.
- 3. Finite Element Seismic Analysis of the Aux. Safeguards Cabinet, Loop Stop Valve Cabinet, and Modified Safeguards Cabinet, No. WCAP-9858, Rev. O dated April 1981 by Westinghouse.
- 4. Seismic Qualification of the Rotary Relay for Use in the Troian and Diablo Canyon Auxiliary Safeguards Cabinets, No. WCAP-89L1, dated October 1977 by Westinghouse.
The loop stop valve cabinet was analyzed for structural adequacy. The analysis consisted of a three-dimensional, finite element model of plates, lumped masses, beams and springs. The model was restrained at the base nodes. The WECAN computer code was used in the analysis. The model was authenticated with a partial test which exhibited good correlation with the calculated frequencies. The following frequencies were found by analysis (below 50 Hz): s/s = 17.44 Hz, f/b = 21.60 Hz, v = none, torsional = 28.92 Hz. This model was also used to generate in-equipment response spectra. These spectra were generated with the help of synthetic time histories. The response spectra for these time-histories enveloped the generic required response spectra. Generic required response spectra enveloped the Beaver Valley-2 required spectra. All these analyses used a 5 % damping. An all-enveloping in-equipment spectrum was generated from the individuals. This spectrum was used for the components' qualification. The active components of the loop stop valve cabinet were either tested individually or in another (but similar) cabinet. MDR 131-1 and MDR 134-1 were tested on the Safeguard test cabinet. These showed adequate structural integrity and operability was verified. MIDTEX 156-14T200 relays were tested independently with pseudo triaxial sine beat input. One of the specimen showed contact chatter greater than 2 millisecond during SSE level testing. However, this contact bounce was within the acceptance criteria. The analysis model is correlated with test results. The analysis results are within allowable limits. The active components tested out to be acceptable. The components are qualified for a life of 5 years. The . procedure to ensure the operability of the equipment throughout its qualified life is being addressed by the utility surveillance and maintenance program. The field inspection, however, indicated that this cabinet was quite close to another one which raises the possibility of interaction during a seismic situation. This problem is generic in nature. Based upon the observation of the field installation, review of the qualification documentation and applicant's response to our questions, the loop stop valve protection cabinet and its components are adequately qualified pending satisfactory resolution of the generic problem of adequate clearance between cabinets. 2.10 8 in. Motor Operated Gate Valve (NSSS-16) The 8-in. motor operated gate valve (model No. 08000GM82FBB000, MPL No. 2CHS*LCV115B) was supplied by Westinghouse according to purchase order Beaver Valley 2 SSER 4 12 Appendix Q
No. 546-CCC-173659-BN and specification No. 952173 Rev. 6 and G678852 Rev 2. The assembly has a Limitorque SB-00, ID No. 714572WC, Serial No. 264170 operator. It was butt welded into the piping located at the 721-ft elevation of the auxiliary building. The stem was vertical for this mounting. This valve, a part of the chemical and volume control system, opens a path from the refueling water storage tank to the charging pumps' suction in the injection mode. The seismic qualification of the assembly is based on a combination of tests and analyses. These are documented in the following reports:
- 1. Operability Test Report of Westinghouse Nuclear Gate Valves, No. 4995, Rev. O, dated January 28, 1977, prepared by Westinghouse.
- 2. EQTR--Equipment Qualification Test Report Limitorque Electric Motor Operator (Environmental and Seismic Testing), No. WCAP-8687 Supp. 2 H04A, Rev. 2, dated March 1983 by Westinghouse /Limitorque.
- 3. Design Report for Westinghouse Class I Nuclear Valves, No. EM-4981, Rev.1, dated November 15, 1975, by Westinghouse.
Seismic loads are considered in the qualification. The required accelerations at the valve location are: s/s f/b v OBE (actual / generic) 0.195/1.05 g 0.283/1.05 g SSE (actual / generic) 0.201/1.05 g 0.515/2.1 g 0.558/2.1 g 0.396/2.1 g. LOCA loads are transmitted to the valve through piping. Allowable nozzle loads on the valve account for it. The qualification consists of a static, two-dimensional, finite element analysis for stresses and displacements. The stresses are compared to ASME allowables and the deflections are governed by available clearance. They are within the allowable limits. The valve calculated frequencies indicate that it is relatively rigid. The Limitorque operator was tested. For the test the actuator was bolted to a fixture in the same manner as it would be to a valve. In turn the fixture was bolted to the table. A resonance search test from 2 to 35 Hz with a 0.75 g magnitude, single axis, sinusoidal sweep input did not indicate any frequency below 35 Hz. Subsequently, single axis RIM (required input motion) tests with sine-beat inputs were performed. There were five OBE and one SSE level tests performed. Aging was performed before the tests. Structural integrity and operability were verified. Inputs levels were as follows: s/s f/b v OBE 5.0 g 5.0 g 5.0 g SSE 7.6 g 7.6 g 7.6 g. , Beaver Valley 2 SSER 4 13 Appendix Q
i There was another set of tests performed on assemblies with valves of sizes 4 and 12 in., mounted on pipe stubs 10.2 in. in length on each side and the stubs then bolted to the test fixture. This series was static bend tests. The input level was: s/s f/b v 4.5 g 4.5 g 4.5 g. Structural integrity and operability were verified. The present 8-in. valve is qualified by similarity and interpolation. The valve was leaking significantly when observed in the field. On inquiry, the applicant indicated that it was being tested and would be repacked to eliminate the leak. Based upon our observation of the field installation, review of qualification reports and applicant's responses to our questions, this unit is adequately qualified for Beaver Valley-2 application. 2.11 Nuclear Instrumentation System Cabinet (NSSS-19) This NIS cabinet (tag. No. RK*2NUC-INS) was built by Westinghouse with model No. 1061E43G01. It is a four bay cabinet located in the control room containing equipment for the reactor protection system which provides alarm functions, secondary control of indicating reactor status during startup, and functions in power operations and overpower trip protection. The cabinet was qualified by testing. Westinghouse documented its testing of the cabinet in WCAP-8687, Supplement 2-E47C. In this testing a resonance search was performed using multiaxis swept sine tests at 0.2 g from 1-50 Hz. Resonances were found in the 3 Hz region (side-to-side) and 5 Hz region (front-to-back). Subsequently, pseudo triaxial sine-beat tests were performed developing a TRS which when plotted with 5% damping enveloped with considerable margin the RRS plotted at 4% damping. The testing on the four bay cabinet did not include environmental qualification of the cabinets components. Therefore, Westinghouse Report WCAP-8021 which included the seismic and environmental qualification tests of a two bay cabinet was added to the qualification package. Based on the observation of the field installation and review of the qualification documentation, this cabinet is seismically qualified 2.12 Centrifugal Charging / Safety Injection Pump (NSSS-20) The centrifugal charging / safety injection pump was manufactured by Pacific Pumps (Model 2-1/2 in. RL/IJ, 11 stage). The motor and gear unit was supplied by Westinghouse. The pump, gear unit, and motor are located ' in the auxiliary building at the 737 ft 8 in, elevation. The pump is used for safety grade cold shutdown and high head safety injection. Beaver Valley 2 SSER 4 14 Appendix Q
The pump unit was qualified by a combination of test and analysis. A list of test reports used to qualify the pump unit is contained in Westinghouse Letter MED-PVE-4617, Beaver Valley Unit 2 Charging Pump Natural Frequencies, dated October 2, 1986. A similar motor, about 4 in. longer and 400 lb heavier, was tested as reported in Westinghouse Seismic Shop Order No. 75F32351, Seismic Analysis of Charginc Pump Motors for Blackhawk Nuclear Stations, No.1 and 2 and Braidwooc Nuclear Stations No. 1 and 2 Commonwealth Edison Co., dated 9/16/76. An electromagnetic shaker was used to excite the motor from 5 to 180 Hz. The lowest natural frequency was found to be 61 Hz. Stresses and rotor deflections were found for OBE and SSE by a static analysis as reported in Westinghouse Report No. M030601, Seismic Analysis Report for the DMW Charging / Safety Injection Pump Motors, dated 8/29/83. The shaft deflection was 0.00275 in. compared to a clearance of 0.051 in. All stresses in the motor were found to be below allowable values. A 2.1 g acceleration was applied in all three directions simultaneously in this analysis. This acceleration is much greater than the maximum required acceleration of 0.4 g. Therefore, operability of the motor is demonstrated. Thermal and radiation aging of a similar motor insulation was performed and discussed in Westinghouse Report No. WCAP-8687, Supp. 2-A05A, Rev. O, Equipment Qualification Test Report Westinghouse large Pump Motors for Use in a High Energy Line Break Environment Outside Containment (Environmental and Seismic Testing), dated November 1985. Five OBE and four SSE qualification tests were performed with all TRS exceeding generic RRS. The lowest natural frequency of a similar gear unit was 66 Hz as reported in Westinghouse Shop Order 79-R-52594. Stresses in the gear box were found in Westinghouse Shop Order 75-R-40761, dated 10/21/76. All stresses were below allowable values. The lube oil piping system was tested as reported in Westinghouse Order No. B-48901, Natural Frec uency Test, dated April 27, 1982. No frequencies below 35 Hz were icentified. Stresses in the piping were reported in Pacific Pumps Design Report K-156. This report was not reviewed. l Stresses in the pump were found using a static analysis in Pacific Pumps Design Report No. K-362-7, Rev. 2, Nuclear Service Pump Design Calculations, Class 2, Pump Size 2-1/2 RL/IJ, dated 11/30/81. The SSE stresses were found using 3 g horizontal and 2 g vertical acceleration acting simultaneously along with pressure loads. The highest stress (15,935 psi) occurred in the flattened section of the suction nozzle. The I normal allowable stress was 16,600 psi and the faulted allowable stress was 29,800 psi. The shaft deflection was 0.002 in. compared to a clearance of 0.006 in. Therefore, operability of the pump was demonstrated. The nozzle loads considered in the analysis were reported in Westinghouse Equipment Specification 678815, Rev. 2, Class 2 Pumps Based on ASME Boiler and Pressure Vessel Code Section III - Rules for Construction Beaver Valley 2 SSER 4 15 Appendix Q
! of Nuclear Power Plant Components, dated 9/6/73. These loads were found by an empirical method based on experience. Beaver Valley 2 is in a process 1. of reconciliation of nozzle loads on equipment and piping. We request confirmation of the completion of this program. I Based on the observation of the field installation, review of the qualification documents, and the applicant's response to questions, the centrifugal charging / safety injection pump is adequately qualified for the i prescribed loads. 2.13 Position Transmitter-7300 Printed Circuit Card (NSSS-21) The printed circuit (PC) card (model No. NCT1, MPL No. 2RCS*ZT2508) 1 was in slot 50 of cage 41, channel 10-2HS/4438. It was supplied by Westinghouse according to purchase order No. 180506 and specification - No. 952255. It was located in the control room at the 707 ft 6 in. elevation. The PC card was mounted within a card frame secured with two set screws. The card frame was in the 7300 cabinet. The cabinet was mounted onto the structural steel raised floor. The mounting of cards in the cabinet (all of them) was not complete. l This item was picked as a surprise item to evaluate the efficiency of the retrieval system. The retrieval system is adequate. It was only briefly reviewed for completion. The review established that a complete qualification documentation package was in place but did not establish the merit of the qualification due to lack of time. l 1 i i Beaver Valley 2 SSER 4 16 Appendix Q
- . .__ - -- _ -_. - - . - . . . _ . -. - - - , . . -- ~
- 3. BALANCE OF PLANT (80P) EQUIPMENT 3.1 Motor Operated Butterfly Valve (80P-1)
The valve with tag No. 2CCP*M0V150 . is model No. 1400 from the Henry Pratt Co. Its actuator is a Limitorque model No. SMB-00-5/H3BC. The valve is bolted to a 14-in. component cooling water line at elevation 724 f t, inside containment, and functions for containment isolation. Seismic qualification was performed by a combination of testing of the actuator and structural analysis of the valve. Acton Environmental Testing Company documented their tests on this actuator model in report No. 19913-84N. This actuator was required by specification to meet 3 g acceleration in all three directions. A swept sine resonance search was performed indicating no natural frequencies below 33 Hz. Single axis swept sine tests at 3 g were performed for the OBE level of tests. Sine dwell tests were performed for the SSE level of tests, dwelling at twenty discrete frequencies below 33 Hz with input levels at 3 g at 2 Hz and 4.5 g at the remaining dwell frequencies. Piping accelerations were then required to be kept below these acceleration levels. Six OBE and three SSE tests were performed. Henry Pratt Co. performed the analysis on the valve itself and documented their calculations in report Number A-0027-2. A static analysis was performed on the valve with the 3 g imposed since this is rigid with respect to the 33 Hz cutoff. Stress criteria was based on ASME Section III allowables. The highest stresses were located in the trunnion bolts with a calculated stress of 29,604 psi and an allowable of 30,000 psi. As part of this system the motor control cabinet which houses the power supply for the actuator was also inspected during the walkdown. This cabinet with tag No. MCC*2-E05 was a Gould cabinet ITE motor control center. Upon inspecting the cabinet it was discovered that a large cable tray was routed very closely along the top, backside of the cabinet and actually touching the cabinet at one location. Concern for the inducement of impact loads into the cabinet during a seismic event was expressed. Duquesne Light indicated that this was another example of interference which would be addressed when the electrical systems are walked down according to Beaver Valley 2 Project procedure BVM-236. Based on the observation of the field installation and review of the qualification documentation the valve and actuator are considered qualified. However, the addressing of the specific as well as the generic questions of component interference must be verified before qualification is complete. 3.2 Feedwater Isolation Valve (B0P-5) The feedwater isolation valve was supplied by Borg-Warner /NVD, model No. 435XABS-001. The valve is located in the main steam valve area at the 775 ft 8 in. elevation. The valve provides rapid stoppage of feedwater flow in the event of a main steam line break. Beaver Valley 2 SSER 4 17 Appendix Q
The feedwater isolation valve was qualified by a combination of test I and analysis. The operator was qualified using a test of an identical operator but with the addition of two small manifolds consisting of a shutoff valve, a pressure gauge, and several 0-rings. These manifolds were installed on the hydraulic and pneumatic sides of the system. The testing is reported in Borg-Warner Nuclear Valve Division Report No.1736, Rev. C, Qualification Test Plan and Test Results Report for a Pneumatic-Hydraulic Operator P/N 38991, dated 2/22/83. Thermal and radiation aging of the operator were performed before vibration testing. The operator was installed on a production yoke which was mounted on a section of carbon steel pipe to closely simulate an actual valve. The yoke was then attached to the pipe using the production clamp ring. A hydraulic cylinder coupled to an operator rod end was used to simulate operator loading. This mounting adequately simulated the plant mounting. A 0.2 g resonance search was performed from 1 to 100 Hz. One natural frequency (28 Hz) was identified below 33 Hz. Qualification testing is reported in Wyle Laboratories Report 57530, Nuclear Environmental Qualification Testing of One Hydraulic Valve Operator P/N 38991 for Nuclear Valve Division, dated April 13, 1978. Random motion biaxial input from 1.2 to 35 Hz was applied to the table. It was necessary to superimpose sine beats on the random input because of the high acceleration levels required for qualification. The TRSs exceeded the RRSs for all tests above a frequency of about 4 Hz. Five OBE tests and five SSE tests were performed in two horizontal orientations. Operability of the operator was monitored during testing. Stresses in the valve body were found from a static analysis as reported in Borg-Warner Nuclear Valve Division Report No. NSR435XAB5, Rev. A, Static Deflection Test Procedure, dated 7/27/78. The lowest natural frequency was calculated as 60 Hz. A static analysis was performed using 5.83 g horizontal and vertical accelerations applied simultaneously. The highest stress for the faulted condition was 19 ksi compared to an allowable stress of 21.6 ksi for the condition. A static deflection test was performed. The opening and closing times of the valve were checked and found to be within allowable limits. Leakage rates under pressure were also checked and found to be within allowable values. Based on the observation of the field installation, review of the , qualification documents, and the applicant's response to questions, the ( feedwater isolation valve is adequately qualified for the prescribed loads. 3.3 Alternate Shutdown Panel (B0P-8) The alternate shutdown panel (no model number, MPL No. PNL*2ALTSHUTDN) was manufactured and supplied by Systems Control Corporation according to purchase order No. 2BV-731 and specification No. 2BVS-731. It was located at the 755 ft elevation of the auxiliary building. It was vertically mounted on the floor. The attachment to the floor consisted of welds 1/4 in. x 2 in. on 6 in. center (inside and outside) on front and back of the cabinet. This panel is a part of the control system and intended for use in case of fire or other disabling situations of the main shutdown panel. The panel had only a temporary paper tag on it for identification. Beaver Valley 2 SSER 4 18 Appendix Q
The qualification of this panel is based on a combination of tests and analysis. Details of the analysis and tests are in the reports: Seismic Structural Qualification of Systems Control Emergency Shutdown Panels for Beaver Valley Power Station - Unit #2 Per Stone Webster Engineering Specification No. 2BVS-731, No. 16239-81N-3, Rev. 4, dated December 10, 1985 and Re.) ort of Test for Class IE Equi) ment Seismic (ualification, No. 2004 ,34N, Rev. O, dated July 23, 1936, prepared by FTS/Acton and reviewed by System Controls. l The cabinet is analyzed based on a two-dimensional (a 2-D vintage plant) finite element model. The model is comprised of beam and plate elements. The calculated lowest frequencies are in the 13 to 20 Hz range. Thirteen Hz frequency is judged to be local. Hence, the 20 Hz. frequency is assumed to be the first fundamental mode. It is therefore concluded that the cabinet is relatively rigid. Consequently, no magnification takes place through the cabinet. As a result, all the instruments inside are to be qualified based on floor response spectra (RRS). Stone & Webster Engineering Corporation (SWEC) had not completed the review of the qualification documents for the internal class 1E items. Therefore, a list of active IE items in the cabinet was not available and the qualification of these items could not be established. The review indicates the following deficiencies with the qualification of the alternate shutdown panel, and must be completed in order for this item to be qualified:
- 1. The finite element model for the cabinet structure was not authenticated and hence the conclusion of no magnification through the cabinet is not supported.
The authenticity of the model should be established and nonmagnification of the floor spectra may then be demonstrated by response analysis.
- 2. A list of all 1E internals of this cabinet should be prepared. '
Their qualification should be established, reviewed and forwarded ' to NRC for staff review.
- 3. An auditable trail and identification between the field equipment and qualification documents must be established and maintained on a permanent basis.
Based upon the observation of the field installation, review of the qualification documents and the responses piovided by the applicant, the alternate shutdown panel is qualified for Beaver Valley-2 site pending the satisfactory resolution of the above three items. 3.4 Central Station A/C Unit (B0P-9) This unit (tag No. 2HVR*ACU2078) is model No. 39ED21 manufactured by Carrier Air Conditioning Company and is located in the Safeguards Building at the 743 ft elevation. The unit sheet metal housing is roughly 10 ft Beaver Valley 2 SSER 4 19 Appendix Q h
square and 3 ft deep and anchored on its skids by 16-1/2 in. bolts. Its function is to remove equipment heat load during a design basis accident. John Henry Associates, Inc. documented the unit's seismic qualification by analysis in their report No. JHA-85-263, Rev. 1. A NASTRAN finite element beam and plate model of the unit indicated its lowest natural frequency to be 25.9 Hz (f/b) while the ZPA for the support location starts at 10 Hz. Therefore, a static analysis of the unit was performed with a 0.204 g ZPA horizontal and 0.122 g vertical for OBE and 0.44 g ZPA horizontal and 0.231 g vertical for SSE. The unit boundary conditions were assumed pinned at the anchor bolting locations. Calculated stresses were 30% of the allowables while maximum deflections were 20% of those required for operability. Based upon the observation of the field installation and review of the qualification documents this unit is seismically qualified. 3.5 Vital Bus Distribution Panel Board (B0P-14) The vital bus distribution panel board was supplied by Systems Control. The model number was not attached to the panel but the Beaver Valley mark number (PNL*VITBS2-1C) was attached. The qualification package included a drawing which showed this mark number. Therefore, traceability was established. This panel supplies power to the chlorine gas monitor and to the plant safety monitoring system. The panel is located in the control building at the 736 ft 2 in. elevation. The vital bus panel was qualified by analysis in Acton Environmental Testing Corporation Report 15902, Rev. 4, Seismic Structural Qualification , of Distribution Panels for Beaver Valley Power Station Unit No. 2, dated , 4/25/86. A finite element model of a similar panel--the 120/240 Vac emergency distribution panel--was developed using the STARDYNE program. Three natural frequencies were identified below 33 Hz. They were 22.11 Hz, 23.22 Hz, and 30.54 Hz. All three of these frequencies were attributed to vibration of the small door on the panel. No components are mounted on the small door so the local modes of the door were considered to not have any effect on the rest of the panel structure. Stresses in the panel were found using a static analysis. The accelerations used in the analysis were 0.53 g horizontal and 0.32 g vertical. The highest stress in the panel was 1,068 psi compared to an allowable value of 24,000 psi. The circuit breakers inside the panel were qualified by testing in Acton Environmental Testing Corporation Test Procedure 15891, Rev. 4, Qualification Testing of Class 1E Electrical Equiament Used for Nuclear Power Generatinc Stations per the Guidelines of IEEE Std. 323-1974 and IEEE Std. 344-1975, c ated 6/30/84. Thermal, radiation, and mechanical aging were performed. The vital bus panel board contains 8 Heineman Model CD1-G3-U15-120/240-AC-1 breakers and one Heineman Model CD3-A0-DU-240 VAC breaker. The report stated that a circuit breaker similar to the former, Model CD1-G3-U-40-120/240 AC-1, was used in qualification testing. The 3 report did not specifically mention the CD3-A0-DU-240 VAC breaker. During the audit, Beaver Valley personnel found that a similar breaker, a 3 pole , Beaver Valley 2 SSER 4 20 Appendix Q
breaker with one pole used for an alarm function, was tested. The breakers were tested to SSE peak level of about 8 g. This acceleration level exceeds the required valve. Based on the observation of the field installation, review of the qualification documents, and the applicant's response to questions, the vital bus distribution panel board is adequately qualified for the prescribed loads. However, the issue of permanent link between the field equipment and qualification documentation (involving the use of mark number) remains open. It would be handled on a generic basis. 3.6 Air Operated Control Valve (BOP-15) The air operated control valve, supplied by Masonellan (model No. 38-20571), is located in the reactor containment building at the 692 ft 2 in. elevation. The valve is a passive component which is located on a drain line. It is normally closed to preserve the pressure boundary for the accumulator system. It is opened periodically for the accumulator leak test. The qualification documentation referred to the Beaver Valley mark I number rather than the manufacturer's model number. The mark number for this valve had been changed from 2-SSR-A0V-106C to 2-SIS-A0V-8508. The qualification number. package referred to the former number instead of the latter Documentation of the change was not in the " mini" package which ) was prepared specifically for the SQRT audit. However, Beaver Valley personnel located the documentation of the change in the complete package. The reviewer examined the change documentation and considered it adequate. Therefore, traceability between the valve and its documentation was identified. The actuator testing is contained in Masonellan Nuclear Division Seismic Report 1007 Rev. C, Seismic Qualification of Masoneilan Control Valves for Duquesne, Light Company Beaver Valley Unit 2 Test Valve Number 805, dated December 15, 1977. A resonance search identified two natural frequencies at 31 and 33 Hz. Amplification factors were found at these two frequencies by dividing the response of the valve by the input accelerations. accelerations at the center of gravity be between 3 and 4. These two factors were found to Stresses were calculated in the valve for accelerations which were multiplied by these amplification factors. All stresses were found to be acceptable. The maximum amplified acceleration was 14.2 g. The test mounting is discussed in Acton Environmental Testing Corporation Report 13374, Seismic Vibration Testina of One 1" 600 # Valve N-00168-12 for Masoneilan International. Inc.,. dated August 15, 1977. The valve was bolted to a test fixture which was securely attached to a 45 degree biaxial test table. The valve is actually line mounted with no supports in close proximity of the valve. The mounting was considered to be satisfactory because of the application of the amplification factors to give a maximum acceleration of 14.2 g and also because the valve is passive. The requirements of a seismic operability test for this valve are given in Masonellan International, Inc., Report NE-126, Rev. D, General Operability Test, dated 2/11/76. available. Beaver Valley personnel The results of these tests were not stated that the results of the Beaver Valley 2 SSER 4 21 Appendix Q
- operability test are not required since the valve is passive. The reviewer j agreed with this statement.
The NAMC0 limit switch was qualified to an acceleration level of 14.2 g. The documentation was not reviewed. Based on the observation of the field installation, review of the qualification documents, and the applicant's response to questions, the air operated control valve is adequately qualified for the prescribed loads. 3.7 Vital Bus Uninterruptable Power Supply System (B0P-17) I The vital bus uninterruptable power supply (UPS) system was manufactured by Elgar Corporation (model UPS-253-1-110). It is located in the service building at the 730 ft 6 in. elevation. The UPS supplies 120 Vac power for engineered safeguard protection channels and other Class IE 120 Vac electrical loads. The qualification procedure for the UPS is discussed in Elgar Document i No. 1006589, Qualification Procedure to Demonstrate a Forty Year Qualified
' Life Plus One Year for Elgar Model UPS-253-1-110 Vital Bus Uninterruptable
- Power Supply System Under Stone and Webster Purchase Order 2BV-361A, dated
- July 1984. A similar bus was tested (model No. UPS-253-1-101). The cabinets of the two buses were identical but some of the electrical components were different. A comparison of the two buses is included in i
Appendix A of the above report. Two types of integrated circuit operational amplifiers, models LF356H and LM308AH, present in the model 110 but not in the model 101 were not tested. Beaver Valley personnel found that these amplifiers did not perform a Class 1E function. The Model 101 UPS cabinets were qualified by testing as described in Wyle Laboratories Report 58733, Seismic Testing of One Static j Uninterruptable Power Supply Model UPS-253-1-101, Serial No.101 for Elgar Corporation, dated March 11, 1982. The bus system consists of two cabinets l bolted together and welded to a channel embedded in the floor. The a cabinets were welded to a test fixture which was welded to a biaxial test j machine in a manner adequately simulating the field mounting. Tests were j performed with the cabinets bolted together and also separated. The
- cabinets were tested in two different horizontal orientations (90 degrees j apart) with vertical excitation included. A 0.25 g sinusoidal resonance j search was performed in each direction from 1 to 35 Hz. Several natural
- frequencies below 33 Hz were observed, the lowest being about 8 Hz. The i qualification tests were performed using biaxial random motion input from i 1.2 to 35 Hz in one-third octave intervals. Five OBE and one SSE tests i were completed. All TRS exceeded the RRS. The damping was 0.5% for OBE 4
and 1.0% for SSE. Thermal aging was performed as described in Wyle Laboratories Test Report 56872, February 1982. } Based on the observation of the field installation, review of the j qualification documents, and the applicant's response to questions, the vital bus uninterruptable power supply system is adequately qualified for the prescribed loads. Beaver Valley 2 SSER 4 22 Appendix Q
l l 3.8 Quench Spray Chemical Injection Pump (BOP-19) This pump with tag No. 2QSS*P24A is manufactured by Crane Deming with its model referenced by catalog Figure 1549 size 3WF. It is bolted to its rigid support by means of 4-7/8 in. anchor bolts. The horizontally mounted pump injects sodium into the quench spray system at the 728 ft elevation of the safeguards building. Mcdonald Engineering Analysis Co. documented in its report No. ME-981 the analysis that was performed to qualify the pump seismically. An ICES-STRUDL beam finite element model indicated that all natural frequencies were above 33 Hz; therefore a static analysis of the pump was performed. The RRS ZPA for the OBE was 0.137 g horizontal and 0.075 g vertical while the SSE ZPA was 0.282 g horizontal and 0.156 g vertical. a Stress criteria was based upon ASME Code Appendix XVII and showed the maximum calculated stress to be 20,291 psi in the base plate with an allowable stress of 21,600 psi. Critical deflections were calculated to be 0.001 in, with an allowable of 0.0023 in. Based on the observation of the field installation and the review of the qualification documentation, the pump is seismically qualified.
- 3. 9 Service Water Pump (B0P-20)
This deep-draft Byron Jackson pump (tag No. 2SWS*P21A) is model No. 36RXM/Two Stage VCT and is located in the Primary Intake Structure at
. the 705 ft level. The pump is flange mounted using twelve 1-1/2 in, bolts into a steel plate embedded in the concrete floor. The casing has horizontal seismic restraints located at two elevations. The service water pump supplies water for various cooling purposes throughout the plant.
The lower pump section was qualified by the analysis of Stress Analysis Associates, Inc., and reported in their report No. 731-N-0027, Rev. 2/1. The analysis was performed using the computer codes BMOAT, CAMBM, MDLDF and stresses were compared to the allowables of the ASME Code Section III. Since no frequencies were calculated below the ZPA cutoff frequency, which was 10 Hz in this case, a static analysis was performed applying 0.213 g horizontal and 0.198 g vertical for OBE and 0.37 g horizontal and 0.296 g vertical for SSE. The largest stresses calculated were 13,167 psi and allowable stress was 15,000 psi for normal plus SSE level loads. Deflections were 0.0003 in. versus the allowable deflection of 0.004 in. Mcdonald Engineering Analysis Co. analyzed the upper pump portion with the finite element code ICES-STRUDL. This analysis was documented in report No. EL-8-5117-90371, Rev. 1. The lowest natural frequency calculated was 27 Hz which is well above the 10 Hz ZPA cutoff. However, quite conservative acceleration loads of 1 g horizontal and vertical were applied. The largest normal operating plus SSE loads were located in the lower stator yoke bolt (29,783 psi calculated versus 32,000 psi allowable) and the lower bearing housing (8365 psi calculated versus 9375 psi Beaver Valley 2 SSER 4 23 Appendix Q
allowable). Operability was indicated by a calculated displacement of 0.003 in, compared to an allowable of 0.060 in. Based on the observation of the field installation and review of the qualification documents, the pump is seismically qualified. 3.10 Motor Operated Damper (B0P-23) The motor operated damper (vendor model No. DAA-P-7402, manufacturer model No. NH95 reverse type, MPL No. 2HVC*M002068) was supplied by ITT Gen. Contr. according to purchase order No. 2BV-185, specification No. 2BVS-185. The vendor model number could not be verified in the field. However, the qualification report was for DAA-P-7402 which covered drawing Nos. 14130 and 14130-2. According to the applicant, 14130 and 14130-2 are the drawing numbers for damper and operator in the field. This identification trail does not provide a direct auditable link between the field unit and the documentation. This deficiency has been included in the generic findings. The mounting consisted of thirty two 5/8 in, diameter grade 5 bolts attached to a rigid duct and a support. It is a part of the control room air conditioning system. It was located at the 735.5 ft elevation of the control building. The qualification of the unit is based on a combination of test and analysis. The damper is analyzed for stresses and the actuator is tested. The analysis is documented in the report: Revised Seismic (ualification Report of DAA-P-7402 Dampers, No. 90247-1-A, Rev. A, dated March 7, 1979, prepared by American Warming and Ventilation and reviewed by Stone & Webster Engineering Corporation. The tests are discussed in the following two reports: Qualification Test Program on Hydromotor Actuators NH90 Series (Model B) for ITT General Controls, No. 58784, Rev. B, dated April 12, 1984, prepared by Wyle Laboratory and Stone & Webster Engineering Corporation and Test Report for Recualification of ITT GC NH90 Series Hydromotor Actuators, No. 730-1-140, Rev. 1, dated April 24, 1984 prepared byITTGeneralControlsandreviewedbyAmericanWarmingandVentIlation. For the damper a static analysis (analyzed with only two directional loads at a time due to older vintage plant) with one horizontal (E) and one vertical (V) load of 1 g and 1 g, respectively, was performed. Another analysis with another horizontal (W) and vertical (V) having the same magnitude followed. The stresses and deflection are within the allowable limits. The actuator was tested for qualification. For the tests, the specimen was bolted to a flat plate and book-end which was welded to test table. The first series of tests was resonance search with an input of 0.25 g at two octaves per minute from 1 to 200 Hz. Resonances found were: s/s = 24 Hz, f/b = 26 Hz, v = 24 Hz. Subsequently, the qualification tests were done with phase incoherent, blaxial, random inputs. Test response spectrum (TRS) was generated for each case using 5% damping. The TRS enveloped the RRS for each case Beaver Valley 2 SSER 4 24 Appendix Q
adequately. The RRS were generated using 0.5% and 1.0% damping for OBE and SSE, respectively. Vibration aging was performed. There were five OBE and one SSE level tests performed. The stresses from the analyses are within allowables. There were an adequate number of tests performed. The TRS and RRS comparisons from the tests are satisfactory. However, there were 18 deficiencies recorded (too many to list here) during the test. Some of them are not relevant and easily disposed of. But the resolution of deficiencies as reported in report No. 730.1.140 by ITT Nos. 7, 8, 11, 12, 13, 17 and 18 are either not complete or unsatisfactory. Based upon the observation of the field installation, review of the qualification reports and applicant's response to our inquiries the motor operated damper is adequately qualified for Beaver Valley-2 pending satisfactory resolution of the deficiencies above. 3.11 Fuel Oil Pressure Switch (80P-24) The fuel oil pressure switch (MPL No. 2EGF*PS202-2) was supplied by l Colt Industries according to purchase order No. 2BV-230 and specification : No. 2BVS-230. It was located at the 732 ft elevation of the diesel generator building. The mounting consisted of two 3/16-in. diameter bolts. This item was picked as a surprise to monitor the efficiency of the i retrieval system and the completeness of the documentation retrieved. The l documents were retrieved in due time and the documentation was found complete. However, the adequacy of the qualification could not be verified due to limited time available for the audit. Beaver Valley 2 SSER 4 25 AppendixQ
!, 4. FINDINGS AND CONCLUSIONS 4, The review of the Beaver Valley Power Station Unit 2 will be completed l when the following open items are closed. ' 4.1 Generic Issues j 4.1.1 During the field observation, it was found that several of the ! equipment items did not have a model number or a serial number for i identification. They, instead had a mark number put on by the utility. l This made it very difficult to establish a permanent auditable link between
! the field equipment and the qualification documentation. On inquiry, the
! applicant indicated that, in the documentation, the utility has the linkage ! established between the mark numbers and the model or serial numbers. I This, however, puts an extra layer of paper work between the field item and ! the qualification documentation and has the potential of loosing the direct
; traceability. Therefore, the utility should install the model or serial 6 numbers (as the case may be) provided by the supplier on the equipment in F
- ! the field for positive identification and traceability.
S j 4.1.2 During the field observation of the loop stop valve protection i cabinet (NSSS-15), it appeared that the clearance between this unit and the i adjacent cabinets was not adequate. On inquiry, it was learned that this j problem may be associated with other cabinets. Therefore, this problem of i adequate clearance between adjacent cabinets should be addressed on a j generic basis. The response should include the examples of resolution for j typical cases. I 4.1.3 The completion of the program of verification of as built loads for l i pumps and valves is to be confirmed. 4.1.4 The applicant is to inform the staff of the completion of the
- seismic and dynamic qualification program. The completion and confirmation
! must occur prior to fuel load. ), 4.2 Equipment Specific Issues i 4.2.1 The review of the qualification documentation for the RHR heat exchanger raised the following concerns: 1 (i) The appropriateness of using Bijlaard analysis for the 24 in, nozzle-shell junction stresses, 1 (ii) Sizing calculations for the support lug welds, and (iii) The use of specified lug dimensions in the support lug-shell junctionstressanalysis. 1 I The above concerns must be resolved for the RHR heat exchanger to be qualified. l I t 4 1 Beaver Valley 2 SSER 4 26 Appendix Q { 1 I _ ._.__ _ , _ _ _ - - - _ - - ._ - _ . _ _ _ . _ .,_- _ _ _ _ - _ , - - _ _
? \ i
- b ,
p
't 4.2.2 The field observaticn and the review of quglification documents indicated the following issues with respect to the alternate shutdown panel (BOP-8).
{ r (i) Finite element model used for the analysis was not u authenticated, i (ii) There was. neither a list nor qualification of internal 1E instruments, and '
~
(iii) There was no permanent auditable link between the field item A and the documentation. , The qualification is pending and requires resolution of the above s
\
issues of concern. Y s g s. \, 4.2.3 The review of the qualification documentation for the motor operated
' damper indicated that there were a substantial number (too many to
. enumerate here) of anomalies detected. In some cases the acceptance
' . . criteria were changed. The applicant made an initial response to the 1 i
anomalies at the time of the audit, isowever, most of these responses do not explain the reason for acceptabijity' by'the applicant. s i The list of anomalies in report no. '5W84 by Wyle Laboratories should g , l be resolved and the reason of the acceptability discussed in each case. i Also, the change of acceptance criteria should'be justified wherever it was *, l
} done. ,e, "t t
si 4 ' 4.3 Conclusion ! Based on our review, we conclude that, although the open issues need i to be resolved, an appropriate qualification. program has been defined and implemented for the seismic Category I mechanical and electrical equipment g which will provide reasonable assurance that such equipment will function i properly during and after the excitation due to the vibratory forces A
' imposed by the safe shutdown earthquake in combination with normal operating loads.
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Beaver Valley 2 SSER 4 27 Appendix Q
feRC P08041338 U S. feUCLEAR ARGULATOR Y CORInstSSeOgg i REPORT NuM9tR fAss,parf ey TfDC, ede Ver Afe, ea emys
! #,'. . . . - NUREG-1057 m,. 3m 818UOGRAPHIC DATA SHEET Supplement 4 StE IN5tmuCisO885 08e THE RtytR$$
- 3. Tif Amo sue TiTLt 3 Lt AVE SLANK f" Safety Evaluation Report Related to the Operation of /
Beaver Valley Powe Station, Unit 2 f , o ,8 ,,E,o,,, co ,a,E,
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. Au r,.O. . , 1987 e e OAtt m& PORT 155utD
/ MON T ** vtAR s / March 1987 7 Ptm80MM V10mGAN12 ATION Naut AND MaitsN ADOm E 55 t#acter le Cears O PROJECT,T Ass wome umiT Nuustm Division of PWR Licensing Office of Nuclear Reactor gulation / e P N Oa oaANT Nu eta UeS. Nuclear Regulatory Co ssion Washington, D.C. 20555 [-
10 EPONSORIN 4 0RGANilAfiON NAWE AND MAILING ADORESS *easele Ceaes lle T YPE OF REPOmi j Same as 7. above
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