ML20211A049
| ML20211A049 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 05/31/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-1057, NUREG-1057-S01, NUREG-1057-S1, TAC-62944, NUDOCS 8606110036 | |
| Download: ML20211A049 (94) | |
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NUREG-1057 Supplement No.1
' Safety Evaluation Report related to the operation of Beaver Valley Power Station, Unit 2 L
Docket No. 50-412 Duquesne Light Company, et al.
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U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation l-May 1986
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. ~ z c Availability of Reference Materic Cited in NRC Publications Most docunsnE cited in NRC publications will be available from one of the following sources:
- 1. The NRC Public Document Room,1717 H Street lN.W.
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Washington, DC 20555
- 2. The Superintindent of Documents,10.S. Government Printing Office, Po*;t Office Box 37082, Wa'shington, D.C 20013-7082
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- 3. : The Natiomi Technical Informotion Service, Springtield, VA 22101 a.-,
Although tt?e listing that follows represunts the majority of documents cited irtNRC ptblications, it is not inteyled to be exhaustive.
Referenced documents available for inspection and copying for a fee from the NRC Pdlic Docu-ment Room include NRC correspondence and mternal NRC memoranda: NRC Office of Inspection and Enforcement bulletins, circulars, iriformetion notices, inspection and investigatiory notices Licensee Event Rer:ns, vendor reports and correspondence; Commission papers; and applicant and licensee documents and correspondence.
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Documents available from the National Technical informatioISM;vice include NUREG series reports and technical reports prepared by other federal agencias and report's'hrepared by the Atomic Energy Commission, forerunner r:gency to the Nuclear Reguidory Comir.ission.
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Documents available from publS and special technical iltyaries include gl(opep literature items, such as books, journal and periodical articles, and transactions. Federae Register notir,es, federal and state legislation, and congressional reports can ust)lly be obtained from these libraries.*
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NUREG-1057 Supplement No.1 l
Safety Evaluation Report related to the operation of Beaver Valley Power Station, Unit 2 Docket No. 50-412 Duquesne Light Company, et al.
U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation May 1986 e= a'.w, p
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ABSTRACT This report, Supplement No. I 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 at the time the Safety Evaluation Report was published.
0 Beaver Valley 2 SSER 1 iii 1
l
TABLE OF CONTENTS Page ABSTRACT...........................................................
iii 1
INTRODUCTION AND GENERAL DISCUSSION...........................
1-1 l.1 Introduction.............................................
1-1 3
DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS......
3-1
- 3. 7 Seismic Design...........................................
3-1 3.7.3 Seismic Subsystem Analysis........................
3-1 3.9 Mechanical Systems and Components........................
3-1 3.9.6 Inservice Inspection of Pumps and Valves..........
3-1 4
REACTOR.......................................................
4-1 4.2 Fuel System Design.......................................
4-1 4.2.1 Design Bases......................................
4-1 4.2.2 Description and Design Drawings...................
4-1 4.2.3 Design Evaluation.................................
4-1 5
REACTOR COOLANT SYSTEM........................................
5-1 5.2 Integrity of Reactor Coolant Pressure Boundary...........
5-1 5.2.4 Reactor Coolant Pressure Boundary Inservice Inspection and Testing............................
5-1 5.4 Component and Subsystem Design...........................
5-1 5.4.7 Residual Heat. Removal System......................
5-1 7
INSTRUMENTATION AND CONTROLS..................................
7-1 7.2 Reactor Trip System......................................
7-1 7.2.2 Specific Findings.................................
7-1 7.3 Engineered Safety Feature Systems........................
7-1 7.3.3 Specific Findings.................................
7-1 7.5 Information Systems Important to Safety..................
7-2 7.5.2 Specific Findings.................................
7-2 Beaver Valley 2 SSER 1 v
TABLE OF CONTENTS (Continued)
PLg.e 8
ELECTRIC POWER SYSTEMS........................................
8-1 8.2 Offsite Electric Power System............................
8-1 8.2.2 Compliance With GDC 17............................
8-1 8.3 Onsite Power Systems..................................... 8-1 8.3.1 Onsite AC. Power System's Compliance With GDC 17...
8-1 8.3.3 Common Electrical Features and Requirements.......
8-2 13 CONDUCT OF OPERATIONS.........................................
13-1 13.2 Training...............................................
13-1 13.2.1 Licensed Operator Training Program...............
13-1 13.5 Station Administrative Procedures........................
13-3 13.5.2 Operating and Maintenance Procedures.............
13-3 15 ACCIDENT ANALYSIS.............................................
15-1 15.8 Anticipated Transients Without Scram....................
15-1 17 QUALITY ASSURANCE.............................................
17-1 17.4 Conclusion..............................................
17-1 18 HUMAN FACTORS ENGINEERING.....................................
18-1 c
18.l Detailed Control Room Design Review.....................
18-1 18.2 Safety Parameter Display System.........................
18-1
. 19 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS........
19-1 APPENDICES A
CONTINUATION OF CHRONOLOGY OF NRC STAFF RADIOLOGICAL REVIEW 0F BEAVER VALLEY POWER STATION, UNIT 2 B
BIBLIOGRADHY 0
ACRONYMS AND INITIALISMS 4
E:
NRC STAFF CONTRIBUTORS AND CONSULTANTS G
EG&G IDAHO, INC., TECHNICAL EVALUATION REPORT EGG-EA-6865, ON CONFORMANCE TO REGULATORY GUIDE 1.97 Beaver Valley 2 SSER l' vi
4 TABLE OF CONTENTS (Continued)
H ACRS LETTER AND STAFF RESPONSE I
BROOKHAVEN NATIONAL LABORATORY TECHNICAL EVALUATION REPORT ON S0IL STRUCTURE ANALYSIS J
STAFF SAFETY EVALUATION REPORT ON REACTOR TRIP SYSTEM RELIABILITY, 11 EMS 4.1, 4.2.1, AND 4.2.2 0F GENERIC LETTER 83-28 K
STAFF SAFETY EVALUATION REPORT FOR GENERIC LETTER 83-28, ITEM 1.1, POST-TRIP REVIEW LIST OF TABLES PaSe 1.2 Open issues...................................................
1-2
- 1. 3 Backfit issues................................................
1-3 1.4 Confirmatory issues...........................................
1-4 1.5 License condition issues......................................
1-7 l
Beaver Valley 2 SSER 1-vii
1 INTRODUCTION AND GENERAL DISCUSSION 1.1 Introduction The Nuclear Regulatory Commission (NRC) Safety Evaluation Report (NUREG-1057)
(SER) on the application of the Duquesne Light Company (DLC or the applicant) for a license to operate the Beaver Valley Power Station, Unit 2, was issued in October 1985.
This is the first supplement to that document.
The purpose of this first Supplemental Safety Evaluation Report (SSER 1) is to revise the SER by providing the results of the staff's review of new information subsequently submitted by the applicant.
The information provided in letters referenced in this SSER must be acceptably documented in amendments to the Beaver V, alley 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 is a list of references cited in this supplement.* In Appendix D, abbreviations used in this supplement are listed.
Appendix E is a list of the principal contributors to this SSER.
Appendices G, H, I, J, and K are being added to the SER by this supplement.
No changes were made to Appendices C or F.
Tables 1.2, 1.3, 1.4, and 1.5, all corresponding to tables of the same numbers in the SER, provide summaries of the status of open, backfit, confirmatory, and license condition issues, respectively.
If the status of an issue has changed since issuance of the SER, details of the change are documented in this supple-ment.
The next supplement (SSER 2) is expected to be issued in July 1986.
Copies of this SSER are available for public inspection in the NRC Public Docu-ment Room at 1717 H Street N.W, Washington, D.C., and at the B. F. Jones Memorial Library, 663 Franklin Ave., Aliquippa, Pa.
Copies of this 5SER 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. Mr. Tam may be contacted by calling (301) 492-9409 or by writing to the following address:
Peter S. Tam Division of PWR Licensing-A U.S. Nuclear Regulatory Commission Washington, D.C.
20555
- Availability of all material cited is described on the inside front cover of this supplement.
Beaver Valley 2 SSER 1 1-1
Table 1.2 Open issues Issue Status SER section (1) Preservice/ inservice testing Updated in SSER 1 3.9.6 but remains open (2) Pump and valve leak testing Updated in SSER 1 3.9.6 but remains open (3) Inadequate core cooling instrumenta-Under review 4.4.7 tion (Item II.F.2 of NUREG-0737)
(4) Preservice/ inservice inspection Updated in SSER 1 5.2.4.3, program but remains open 5.4.2.2, 6.6 (5) Safe and alternate shutdown Unchanged from SER 9.5.1 (6) Management and organization Unchanged from SER 13.1 (7) Cross-training program Closed in SSER 1; 13.2.1.2 confirmatory issue 45 opened (8) Emergency preparedness plan Unchanged from SER 13.3.3 (9) Initial test program Unchanged from SER 14 (10) Control room design review Updated in SSER 1 18.1 but remains open (11) Safety parameter display system Updated in SSER 1 18.2 but remains open l
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Beaver Valley 2 SSER 1 1-2
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 A
7.3.3.11 (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 - Issues were discussed in appeal meetings, and resolutions were ad-dressed in the SER (October 1985). As each issue is closed, it will be addressed in a future supplement.
C - Closed in SER (October 1985).
Beaver Valley 2 SSER 1 1-3
Table 1.4 Confirmatory issues Issue Status SER Section (1) Operating procedures for continuous Unchanged from SER 2.2.2 communication links (2) Differential settlements of buried pipes Unchanged from SER 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 Under review 3.9.3.1 components (11) Item II.D.1 of NUREG-0737 Under review 3.9.3.2 (12) Seismic and dynamic qualification of Unchanged from SER 3.10.1 mechanical and electrical equipment (13) Pump and valve operability assurance Unchanged from SER 3.10.2 (14) Environmental qualification of Unchanged from SER 3.11 mechanical and electrical equipment (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)
(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 Beaver Valley 2 SSER 1 1-4
Table 1.4 (Continued)
Issue Status SER section (21) Analysis of combined loss-of-coolant-Under review 4.2.3.3(4) accident and seismic loads (22) Natural circulation test Updated in SSER 1 5.4.7.5 but renains opta (23) Reactor coolant system high point vents Under review 5.4.12 (24) Blowdown mass and energy release Under review 6.2.1.3 analysis methodology (25) Containment sump 50% blockage assumption Unchanged from SER 6.2.2 (26) Design modification of automatic reactor Unchanged from SER 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 MOV113C and 1130 (2E) 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 Unchanged from SER 7.5.2.4.
(31) Revision of the FSAR--cold leg accumu-Unchanged from SER 7.6.2.4 lator motor-operated valve position indication (32) Control system failure caused by Unchanged from SER 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 1E system (b) Confirmation'of the protective Closed in SSER 1 8.3.1.2 bypass (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 Beaver Valley 2 SSER 1 1-5
Table 1.4 (Continued)
Issue Status SER section (33) Confirmatory site visit (Continued)
(f) Electrical interconnection between Closed in SSER 1 8.3.1.13 redundant Class IE 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 prngram 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 Under review 9.5.6 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) I11.0.1.1 Unchanged from SER 13.5.2 (b)
II.K.1.5 and II.K.1.10 Under review 15.9.2 15.9.3 (c)
II.K.3.5 Under rev.iew 15.9.9 (d) II.K.3.17 Under review 15.9.11 (e) II.K.3.31 Under review 15.9.14 (42) Plant-specific dropped rod analysis Under review 15.4.3 (43) Steam generator tube rupture Unchanged from SER 15.6.3 (44) Quality assurance program Closed in SSER 1 17.4 i
Beaver Valley 2 SSER 1 1-6
Table 1.4 (Continued)
Issue' Status SER section (45) Cross-training of Unit 1 & 2 operators Opened in SSER 1 13.2.1.1 (46) Control room isolation on high radiation Opened in SSER 1 7.3.3.9 signal (47) Review of procedures generation package Opened in SSER 1 13.5.2 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 I
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Beaver Valley 2 SSER 1 1-7
3 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS 3.7 Seismic Design 3.7.3 Seismic Subsystem Analysis In the SER, the staff reported that along with consultants from Brookhaven National Laboratory (BNL), it conducted a confirmatory audit of the applicant's scil-structure interaction (SSI) enalyses of the containment structure to gain further confidence in the seismic design adequacy.
As a result of this audit, the staff received additional information from the applicant on August 7, 1985.
The staff and BNL have now completed the review of this information.
The BNL technical findings are included as Appendix I to this supplement. The staff concurs with BNL's technical findings and the NRC safety evaluation is based on these findings.
The following is a brief summary of key technical findings (for details, see Appendix I).
As discussed in the appendix, the applicant has complied with the staff's posi-tion on SSI analysis, which requires SSI analysis by different methods [ Standard Review Plan (SRP) Section 3.7.2], by performing both finite element (PLAXLY and FLUSH analyses) and substructure analysis (three-step) for the containment building.
Furthermore, the comparisons of design floor spectra based on PLAXLY analysis with the spectra generated frca the three-step approach indicate that the design floor spectra, essentially, envelope the spectra from the three-step approach.
The design spectra have been smoothed considering wide variation in the soil properties and do not exhibit sharp valleys at any frequencies.
The comparisons between the site-specific spectrum at foundation elevation in the free field (note that the staff has completed the review of site-specific spectrum and, as discussed in Section 2.5.2.6.5 of the SER, considers it adequate) and translational spectra obtained from the FLUSH analysis and three-step approach at that level indicated that the site-specific spectrum is lower than both the FLUSH and three-step spectra except for a small region around 3 Hz.
- However, as discussed above, the design floor spectra do not exhibit dip at this frequency and they envelope the three-step spectrum in this region.
On the basis of the above findings, the staff concludes that the applicant's SSI analysis for the containment structure meets the intent of staff acceptance criteria delineated in SRP Section 3.7.2, and ccnfirmatory issue 9 is now considered resolved.
3.9 Mechanical Systems and Components 3.9.6 Inservice Inspection of Pumps and Valves The October 1985 SER stated that "the applicant has not yet submitted the pro-gram for the preservice and inservice testing of pumps and valves." By letter dated December 26, 1985, the applicant submitted the preservice testing program entitled "Duquesne Light Company Beaver Valley Power Station Unit 2 Preservice Inspection Program." The review of this program will be included as a part of the review of the inservice testing program.
Beaver Valley 2 SSER 1 3-1
By letter dated February 4,1986, the applicant committed to provide for staff review on April 1, 1986, the propcsed inservice testing program.
If such sub-mittal is not made or if the staff cannot satisfactorily complete its review, this item will be added as a license condition in the operating license.
For the present, there is no change in status of open issue 1.
In the SER, the staff indicated that the allowable leak rate limit for pressure isolation valves (PIVs) was to be no more than 1 gpm for each valve.
Since that time, NRC has adopted a revised and more realistic leak rate criterion for PIVs.
The new acceptable leak rate is 1/2 gpm for each nominal inch of valve size up to a maximum of 5 gpm.
In addition, the requirements of paragraph IWV-3427(b) of Section XI of the ASME Code are to be applied in order to deter-mine if the leak rates are acceptable.
The applicant may submit revised pro-posed technical specifications to comply with this new criterion.
The staff is currently seeking approval from the Committee to Review Generic Requirements (CRGR) on the list of PIVs to be tested and the frequency of testing.
Therefore, open issue 2 remains unresolved pending the staff's formal position on PIVs.
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Beaver Valley 2 SSER 1 3-2
4 REACTOR 4.2 Fuel System Design 4.2.1 Design Bases In its SER, the staff indicated that the applicant should confirm that the peak pellet design-basis burnup of 53,000 mwd /MTU is consistent with the region dis-charge burnup of 33,000 mwd /MTU.
In a letter dated September 13, 1985, the applicant stated that the peak pellet burnup of 53,000 mwd /MTU bounds the maxi-mum expected peak pellet burnup of fuel operating up to 33,000 mwd /MTU batch j
average discharge burnup.
Thus, confirmatory issue 15 is satisfactorily resolved.
4.2.2 Description and Design Drawings The applicant indicated (letter dated September 13, 1985) that the typographic errors in FSAR Tables 4.1-1 and 4.3-1, and Figures 4.2-1 and 4.2-2 would be i
corrected the next time the FSAR was revised.
By Amendment 11 to the FSAR, these errors were all s.orrected. Thus, the staff concludes that confirmatory issue 16 is resolved.
4.2.3 Design Evaluation 4.2.3.1 Fuel System Damage Evaluation (6) Rod Bowing In its SER, the staff stated that the applicant must confirm that the analysis for determining the magnitude of fuel rod bowing has been performed.
In a letter dated September 13, 1985, the applicant stated that the rod bowing analysis has been completed for Beaver Valley Unit 2 using the approved correlation given in Westinghouse report WCAP-8691, Revision 1.
The results of this analysis were used in the departure from nucleate boiling (DNBR) analysis as described in FSAR Section 4.4.2.2.5.
The staff therefore, concludes that confirmatory issue 17 is resolved.
(8) Fuel Rod and Nonfueled Rod Pressures In its SER, the staff indicated that the applicant should confirm that the rod internal pressure is consistent with the approved Westinghouse topical report, WCAP-8963.
By a letter dated September 13, 1985, the applicant stated that the rod internal pressure was analyzed using the approved method described in WCAP-8963, and meets the criteria described therein.
The staff, therefore, concludes that the rod pressure analysis is acceptable, and confirmatory issue 18 is resolved.
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Beaver Valley 2 SSER 1 4-1 l
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4.2.3.2 Fuel Rod Failure Evaluation (2) Cladding Collapse In its SER, the staff mentioned that the applicant must confirm that the calcu-lated cladding collapse time exceeds the expected residence time.
By a letter dated September 13, 1985, the applicant confirmed that the predicted cladding collapse time does exceed the expected residence time of the fuel using the approved WCAP-8377 methods.
The staff thus concludes that cladding collapse will not occur in Beaver Valley Unit 2, and confirmatory issue 19 is resolved.
4.2.3.3 Fuel Coolability Evaluation (4) Structural Damage From External Forces In its SER, the staff identified two confirmatory issues:
the combined seismic and loss-of-coolant-accident (LOCA) loads using the square-root-of-the-sum-of-squares (SRSS) method for the grid analysis, and the use of a new code, MULTI-FLEX 3.0.
In a letter dated December 18, 1984, the applicant showed that the combined seismic and LOCA loads using the SRSS method were applied to the grid analysis and that the results are acceptable.
Thus, the staff considers that the grid analysis is adequate under seismic and LOCA loading.
This resolves confirmatory issue 20.
As.for the use of MULTIFLEX 3.0, the status of confirmatory issua 21 remains the same; i.e., awaiting the outcome of the staff's MULTIFLEX 3.0 review.
Beaver Valley 2 SSER 1 4-2
f 5 REACTOR COOLANT SYSTEM 5.2 Integrity of Reactor Coolant Pressure Boundary 5.2.4 Reactor Coolant Pressure Boundary Inservice Inspection and Testing 5.2.4.3 Evaluation of Compliance With 10 CFR 50.55a(g)
The October 1985 SER stated that the applicant had not yet submitted the pro-i gram for the preservice and inservice inspection of welds.
By letter dated December-26, 1985, the applicant submitted the preservice inspection program entitled "Duquesne Light Company Beaver Valley Power Station Unit 2 Preservice Inspection Program." A preliminary review has been completed and a request for additional information will be transmitted to the applicant.
Open issue 4 re-mains unresolved.
By letter dated February 4, 1986, the applicant committed to provide the inservice inspection program for staff review on December 31, 1986.
The staff will evaluate this program before it commences during the first refueling outage.
Therefore, the inservice inspection portion of open issue 4 will be separated and imposed as a license condition.
5.4 Component And Subsystem Design 5.4.7 Residual Heat Removal System In a letter dated December 24, 1985, the applicant requested additional infor-mation from the staff to document why the North Anna 2 test results of the natural circulation test do not meet Branch Technical Position (BTP) RSB 5-1 guidelines.
In response to that request, the staff proposes three specific points pertain-ing to the inadequacy of the North Anna test for comparison.
According to BTP RSB 5-1, only safety-related systems may be used to take the reactor from normal operation to cold shutdown in the event of a power loss.
If the applicant wishes to take credit for non-safety-related equipment, supporting analysis must be provided to the NRC staff for consideration and review.
North Anna 2 does use non-safety-related equipment, yet the licensee has not submitted any docu-mentation that addresses the use of this equipment in the test.
Therefore, on this point alone, North Anna 2 has not demonstrated compliance with BTP RSB 5-1.
The second point involves the test procedure itself.
In accordance with BTP RSB 5-1, the coolant temperature was brought down from Mode 1 (power operation) to Mode 4 (hot shutdown) levels.
However, hot shutdown was maintained for less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in the North Anna test.
Standard Review Plan (SRP) Section 5.1 specifies that operation at hot shutdown must be maintained for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
Thirdly, Section G of SRP Section 5.1 also states that the period at Mode 4 must be followed by cooldown to the point that the residual heat removal system (RHRS) can be placed in operation.
The RHRS initiation temperature for North Anna 2 i
Beaver Valley 2 SSER 1 5-1
was 350 F, but the reactor was only cooled down to 449 F.
Again, these test results do not comply with BTP RSB 5-1.
However, the test that was conducted at Diablo Canyon also served as an effort to create the only prototype specifically designed for compliance with BTP RSB 5-1 requirements.
Costly equipment installed for this test enabled the licensee to make measurements--especially measurements of conditions in the upper-head region of the reactor vessel--that other plants are not equipped to provide.
The presence of voids in the upper head leads to a number of safety concerns, including resultant thermal stresses on the vessel.
Analysis of these significant conditions was feasible at Diablo Canyon because of the unique in-strumentation available in the upper head.
The Beaver Valley fuel loading date is almost 1 year away.
There is a sufficient amount of time available to conduct an accurate and valid analysis of BTP RSB 5-1 compliance.
Since the Diablo Canyon' test is the only valid prototype for a BTP RSB 5-1 natural circulation test, an analysis should be performed on the basis of the Diablo Canyon test results.
Confirmatory issue 22, therefore, remains unresolved.
Beaver Valley 2 SSER 1 5-2
7 INSTRUMENTATION AND CONTROLS 7.2 Reactor Trip System 7.2.2 Specific Findings 7.2.2.7 General Warning Alarm Reactor Trip In Section 7.2.2.2.3 of the FSAR, the applicant provided a description of the general warning alarm reactor trip.
This system monitors various conditions, such as power supply output and test switch positions, in the solid state pro-tection system.
If any monitored conditions in a train are abnormal, the alarm relay for that train is deenergized.
This actuates the train trouble annunciator in the control room.
If any abnormal condition occurs in one train while an abnormal condition exists in the other train, the reactor is automatically tripped.
This trip provides protection for conditions under which both trains of the reactor trip system may be inoperable.
Originally, the staff's review indicated that this trip had not been part of the Beaver Valley Unit 2 design.
In response to a request from the staff, the applicant has verified that this trip is part of the plant reactor trip system.
The staff finds that this additional reactor trip conforms to the applicable requirements and guidelines for the reactor trip system (RTS) as discussed in Section 7.2.3 of the SER (October 1985), and is, therefore, acceptable.
7.3 Engineered Safety Features Systems 7.3.3 Specific Findings 7.3.3.9 Control Roor. Isolation on High Radiation Signal Control room isolation occurs as a result of one of several events (see SER Section 9.4.1), among them, high radiation detected by area monitors.
In the SER, the staff stated that the applicant had revised FSAR Figures 7.2-1 (Sheet 8) and 7.3-13 in Amendment 8 to delete control room isolation on a high radiation signal.
Staff review of FSAR Amendment 11 indicates that control room isolation on this signal has now been correctly included in the plant design.
By letter dated April 1, 1986, the applicant submitted additional information.
The staff is reviewing this submittal and will report its~ findings in a future SER supplement.
This action is tracked by confirmatory issue 46.
7.3.3.10 Automatic Opening of SWS Valves M0V113C and 113D In the SER, the staff stated that FSAR Figure 9.2-4 erroneously showed that valves MOV113C and 1130 received automatic open signals.
In Amendment 10 to the FSAR, Figure 9.2-4 was revised to identify valves M0V113A and M0V113D as receiving automatic open signals.
The staff has reviewed this revised informa-tion and considers confirmatory issue 27 closed.
Beaver Valley 2 SSER 1 7-1
d 7.3.3.14 Independence Between Manual and Automatic Actions This entire subsection of the SER is replaced by the following revision:
The applicant's response to IE Bulletin 80-06 states that all circuitry for components actuated by an engineered safety feature (ESF) actuation signal have been designed so that the ESF signal cannot be overridden manually or automati-cally with an ESF actuation signal present.
A component may be reset by first resetting the ESF actuation signal and then manually resetting the components.
In addition to the applicant's response, the staff's review of the transfer from the control room to the emergency shutdown panel revealed that safety
]
injection (SI) pumps cannot be stopped manually if SI is initiated after the j
transfer.
During the June 29, 198^, meeting with the applicant, staff concerns were dis-cussed. As a result of that meeting and a subsequent audit of schematic drawings i
for control circuitry of safety-related components, the staff has reached the following conclusions:
(1) Staff concern centered on an interpretation of the applicant's statement:
"A component may be reset by first resetting the ESF actuation" that in-cluded not only manual termination but also manual initiation of protec-tion systems and components.
On the basis of further review and discussion, the staff now concludes that the statement (and the design) only applies to manual termination and that the operator is not prevented from manually initiating safety-related actions.
l (2) Also, on the basis of further review and discussion and the applicant's letter dated September 16, 1985, the staff concludes that even though an SI reset button is not provided on the emergency shutdown panel (ESP),
safety injection can be controlled or stopped from the ESP by following procedures included in operator training and based on the use of local controls.
The staff finds the plant's degree of independence between manual and automatic actions acceptable, and considers this issue closed.
7.5 Information Systems Important to Safety e
i 7.5.2 Specific Findings j
7.5.2.1 Emergency Response Capability, RG 1.97, Revision 2, Requirements Generic Letter 82-33 requested that the applicant provide a report to the NRC L
describing how the postaccident monitoring instrumentation meets the guidelines of Regulatory Guide (RG) 1.97 as applied to emergency response facilities.
The applicant responded to the generic letter on April 15, 1983.
Response specific i
to RG 1.97 was provided on September 12, 1983. Additional information was provided by letter dated June 28, 1985.
Under contract to the NRC, and supervised by the NRC staff, EG&G Idaho, Inc.,
reviewed the applicant's submittals in detail and provided a technical evaluation, reproduced here as Appendix G.
The staff has reviewed this report and concurs with the conclusion that the applicant either conforms to, or is justified in Beaver Valley 2 SSER 1 7-2
deviating from, the guidance of RG 1.97 for each postaccident monitoring variable, except for the variables accumulator tank level and pressure.
Af ter Generic Letter 82-33 was issued, the NRC held regional meetings in February and March 1983 to answer licensee and applicant questions and to respond to concerns regarding NRC policy on RG 1.97.
At these meetings, it was noted that the NRC review would only address exceptions taken to the guidance of RG 1.97.
Furthermore, where licensees or applicants explicitly state that in-strument systems conform to the provisions of the regulatory guide, it was noted that no further staff review would be necessary.
Therefore, the review performed and reported by EG&G only addresses exceptions to the guidance of RG 1.97.
This supplement addresses the applicant's submittals based on the review policy described in the NRC regional meetings and the conclusions of the review as reported by EG&G.
RG 1.97 recommends that instrumentation be provided to monitor the accumulator tank level and pressure.
The applicant has provided this instrumentation which conforms to the criteria for type D, category 2 variables with the exception of environmental qualification.
The applicant states that the accumulators are a passive system and action takes place within 1 minute of the accident or well after safety injection, depending on the leak size.
In either case, operator action is not required to mitigate the consequences of the postulated accident.
Also, the signals from accumulator tank level and pressure instrumentation do not contribute to any automatic safety function.
Although the staff agrees with the applicant that this is a passive system and the variables are not used for automatic initiation of a safety function, the staff believes that this instrument should be provided to permit the operator to determine if the plant safety functions (accumulator discharge) are being performed.
In this regard, the staff finds unacceptable the applicant's proposed exception to the guidelines of RG 1.97.
On the basis of its review of the EG&G report and the applicant's submittals, the staff finds that the Beaver Valley Power Station, Unit 2, design is accept-able except as noted below with respect to conformance to RG 1.97, Revision 2.
The staff recognizes that the operator can infer, either from level or pressure, that the accumulator has injected borated water into the reactor coolant system.
Therefore, it is the staff's position that the applicant should designate either level or pressure as the key variable to determine accumulator discharge and should provide for that variable instrumentation meeting the requirements of 10 CFR 50.49.
If accumulator level is selected as the key variable, then the range should be expanded to meet the regulatory guide recommendations.
It is also the staff's position that the applicant shall install and have operational qualified accumulator tank level or pressure instrumentation at the first sched-u' led outage of sufficient duration, but no later than startup following the first refueling outage.
This action will continue to be tracked by license condition 1.
7.5.2.2 NUREG-0737, Item II.F.1, Accident Monitoring Instrumentation, Positions 4, 5, and 6 In the SER the staff stated that information for positions 4 and 5 was provided in FSAR Table 7.5-1 and that information for position 6 would be provided later.
In FSAR Amendments 9 and 11, the applicant revised Table 7.5-1 to include infor-mation for position 6.
The staff has reviewed the information and finds that l
l Beaver Valley 2 SSER 1 7-3
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l the indication provided for containment pressure, containment water level, and containment hydrogen concentration satisfies the requirements of this NUREG-0737 item and is, therefore, acceptable.
The staff considers confirmatory issue 29 closed.
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Beaver Valley 2 SSER 1 7-4 i-i
l 8 ELECTRIC POWER SYSTEMS 8.2 Offsite Electric Power System 8.2.2 Compliance With GDC 17 8.2.2.3 Independence of Offsite Power Circuits Between the Switchyard and Class 1E System The confirmatory site visit described in the SER took place on February 11-13, 1986.
No new problems were uncovered.
Part (a) of confirmatory issue 33 is thus resolved.
8.3 Onsite Power Systems 8.3.1 Compliance With GDC 17 8.3.1.2 Bypass of Diesel Generator Protective Trips Regulatory Guide (RG) 1.9 requires that certain diesel generator protective trips be bypassed when the diesel generator is required for a design-basis event. The applicant, in accordance with RG 1.9, proposed to include backup phase fault overcurrent detection among those diesel generator protective fea-tures which are not bypassed under accident conditions.
The additional design for the backup phase fault overcurrent detection has two independent sets of relays with coincident logic for trip actuation.
The additional design meets Position 7 of RG 1.9 and, therefore, is acceptable.
Part (b) of confirmatory issue 33 is thus resolved.
8.3.1.8 Diesel Generator Start and Load Acceptance Qualification Tests See Section 8.3.1.13 below.
8.3.1.9 Diesel Generator Load Capacity Qualification Tests See Section 8.3.1.13 below.
8.3.1.10 Margin Qualification Test See Section 8.3.1.13 below.
8.3.1.13 Electrical Interconnections Between Redundant Class IE Buses The confirmatory site visit described in the SER under Sections 8.3.1.8, 8.3.1.9, 8.3.1.10, and 8.3.1.13 took place February 11-13, 1986.
No new problems were uncovered and Parts (c), (d), (e), and (f) of confirmatory issue 33 are con-sidered resolved.
Beaver Valley 2 SSER 1 8-1
8.3.1.14 Automatic Closure of 4160-V Circuit Breaker The following paragraph replaces SER Section 8.3.1.14 in its entirety.
The title of the section is also changed as above.
FSAR Section 8.3.1.1.3 indicates that when a Class 1E 4160-V circuit breaker (except for the diesel generator breakers) is tripped by a protective relay while a safety injection signal is present, the breaker will automatically re-close only if the electrical fault condition which tripped the breaker is cleared, and the lockout relay has been manually reset.
Trip signals to 4160-V emergency loads actuate an annunciator in the main control room. On the basis of this clarification and description, this issue is acceptably clarified.
8.3.3 Common Electrical Features and Requirements 8.3.3.5 Electrical Independence Between Power Supplies to Controls in Control Room and Remote Locations The site visit / verification described in the SER took place February 11-13, 1986.
Part (g) of confirmatory issue 33 is considered resolved.
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Beaver Valley 2 SSER 1 8-2
13 CONDUCT OF OPERATIONS 13.2 Training 13.2.1 Licensed Operator Training Program 13.2.1.2 Beaver Valley Operator Cross-Training Program The original cross-training program was submitted by the applicant to tne staff by a letter on May 28, 1985.
A copy of the program was rutnit%d to NPC Region I by a letter dated June 13, 1985.
The applicant provided a revised version of the cross-training program, which ' evised the format of the program, r
to the staff by letter on October 30, 1985.
The staff used the criteria contained in the applicable portions of 10 CFR 50 and 55, Regulatory Guide 1.8, and H. R. Denton's March 28, 1980, letter concern-ing qualification of reactor operators, to evaluate the acceptability of the applicant's cross-training program.
The following is the staff's evaluation of the elements of the cross-training program using the established criteria.
(1) The purpose of the cross-training program is to familiarize the licensed operators at Beaver Valley Power Station, Unit 1, with the differences between Unit 1 and Unit 2 systems and components; normal, abnormal, and emergency procedures; and plant controls and station para.cters.
The staff feels these objectives are appropriate.
(2) The personnel to be trained are identified as licensed operators on Unit I who would be required to hold a Unit 1/2 dual license.
The st&ff finds the references to dual-licensing of either reactor or senior reactor operators unacceptable.
Reactor operators should not be dually licensed because of the differences between units in component controllers, control board arrangement, and con-trol panel locations.
For example, Unit 2 has numerous controll us located on the back panels; Unit 1 has no controllers on the back panels.
There are different types of controllers used at the two units for most equip-ment, including: feedwater rcgulating valves, charging flow control valve, safety injection flow control valves, power-operated relief valves, and auxiliary feedwater flow control valves.
The location of instrumentation has been changed at Unit 2 by expanding the area over which some indica-tions are displayed, grouping related indications, and moving some indi-cations to the back panel.
For example, the safety-injection portion of the panel has been expanded and arranged into trains and some boron re-ecvery indications have been moved to the back panels.
The methods for displaying certain indications have been changed, such as rod position in-dication, auxiliary feeowater system alignment indications, volume control tank level and pressure, steam flow, and feedwater flow.
The locations of the nuclear instrumentation panels, radiation monitoring panels, the building service panel, and the incore panel differ between units.
The l
annunciators at Unit 2 are rearranged so that less iriiportant alarms are Beaver Valley 2 S50R I 13-1
3 1
grouped in small windows on the back panels and the more important alarms are displayed in large windows over the back panels.
This arrangement is significantly different from the annunciator arrangement at Unit 1.
The remote shutdown panels of Unit 1 and Unit 2 are different and have few similarities.
Because of these significant differences in the main con-trol rooms of Unit 1 and Unit 2, the staff concludes that dually licensing reactor operators would not provide adequate assurance that the facility would be operated safely.
A decision to dually license senior reactor operators should not be made until further reviews have been completed on similarities between operat-ing prccedures, Technical Specifications, and administrative procedures.
The Westinghouse Unit 1 and Unit 2 " Differences Analysis and Accident Analyses," dated January 1986, showed that there was a significant differ ;
ence in Unit 2 response to a feedline break accident owing to the differ '
ences in the auxiliary feedwater system.
The study also stated that plant responses during postaccident recovery and during surveillance /startup/
shutdown were different based on plant design differences.
However, the extent of these differences in plant response, and the effect of these differences on plant operations and procedures have not been reviewed.
Because of the number of unanswered questions concerning possible proce-dural differences resulting from design differences, the staff concludes that no decision can presently be made on dually licensing senior reactor operators.
(3) The training takes about 8 months; however, certain segments of the train-ing, such as simulator training 7.nd theory review, will be deleted if NRC Region I waives these areas of the licensing examination.
The staff finds this unacceptable.
The purpose of any training program for licensing operators is to ensure that the applicant has learned to operate the controls of the facility in a competent and safe manner.
An NRC examination is administered to ensure that each applicant has met the minimum requirements laid out in a training program and, therefore, cannot be used to define the limits of that train-ing program.
(4) The cross-training program consists of the following discrete segments:
(a) Unit 2 System Difference Phase This training provides 15 weeks of classroom lectures and in plant study based on the differences between Unit 1 and Unit 2 systems.
The staff finds this acceptable.
(b) Plant Layout Phase This training provides 3 to 4 weeks of in plant time during the syste.n difference phase of training.
Trainees will complete a checkout sheet and participate in system-oriented oral and/or written examinations.
The staff finds this acceptable.
Beaver Valley 2 SSER 1 13-2
(c) Procedures / Technical Specificatiori Training Phase This training provides instruction on the differences between Unit 1 and Unit 2 normal, abnormal, and emergency procedures and also differences in the Technical Specifications.
The staff finds this acceptable.
(d) License Review Series Phase This training provides a review of major safety-related system dif-ferences, instrumentation and control functions, administrative pro-cedures and various areas of related theory.
The program will be based on the results of a pre-examination and the type of examination to be administered by NRC Region I.
The staff finds the use of the expected NRC Region I examination as a basis for limiting training unacceptable.
(e) Simulator Training Phase-This training, which consists of 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> of simulator training, may be conducted, depending on the type of operational examination to be administered by the NRC Region I.
The staff finds the use of the expected NRC Region I examination as a basis for limiting training unacceptable.
(5) The training program will be administered by the Director-0perations Training.
Records of trainee attendance and examination grades will be maintained.
The staff finds this acceptable.
In summary, the staff finds the applicant's cross-training program acceptable with the following exceptions:
(1) reference to dually licensing operators (2) limiting training based on the format of licensing examinations.
On the basis of the above discussion, open issue 7 is closed and confirmatory issue 45 is opened to track resolution of the two concerns mentioned above.
13.5 Station Administrative Procedures 13.5.2 Operating and Maintenance Procedures The SER of October 1385 indicates that the stafi will describe the result of the review of the applicant's procedures generation package (PGP) but failed to identify such review as a confirmatory issue.
Confirmatory issue 47 is thus opened to track this item.
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15 ACCIDENTAhlYSIS
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15.8 ' Anticipated Transients Without Scram x.
Status of Salem ATW3-Event Issues
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On July 8, 1983, the NRC issued Generic Letter (GL) h3-28 as a result of the anticipated transient without scram (ATWS) events at Salem Nuclear Generating Station. This letter addressed acticns to be t.dk5n by: licensees and applicants to ensure that a comprehensive prograin of preventive maintenance and surveillance testing is implemented for the reactors trip breakers in pressurized water-reactors.
The staff has completed its review of parts of the anplicant's response tc GL 83-28, and will document its results in app 6ndices to the SER.
The staff's review of the following items his been added in Appendices J and K of this sup-plement:
r Item 1.1, Post-Trip Review (Appendix K)J-p
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Item 4.1, Trip Sys, tem Reliability (Appendix J)
. Items 4.2.1 and 4.2.2, Preventive Maintenance Program for Reactor Trip Breakers - Maintenance and Trending.(Appendix J)
Beaver Valley 2 SSER 1 15-1
17 QUALITY ASSURANCE 17.4 Conclusion The October 1985 SER stated that the staff had requested additional information regarding the items under the control of the quality assurance (QA) program and that, except for this information, the applicant's description of the QA program complies with applicable NRC regulations.
By letters dated June 28 and Decem-ber 20, 1985, the applicant has responded acceptably to the staff's request.
This closes confirmatory issue 44, and the QA program description is acceptable for the operations phase of Beaver Valley Power Station, Unit 2.
l Beaver Valley 2 SSER 1 17-1
18 HUMAN FACTORS ENG'NEERING 18.1 Detailed Control Room Design Review On the basis of the requirements of Supplement 1 to NUREG-0737, the applicant must submit a summary report at the end of the detailed control room design review (DCRDR).
By letter dated December 2, 1985, the applicant submitted the summary report.
The staff review of that report indicated that the DCRDR was incomplete.
As part of its review effort, the staff performed a site audit on February 11 and 12, 1986.
Preliminary results of this audit indicate that the applicant is conducting a DCRDR that should satisfy the requirements of Supplement 1 to NUREG-0737, but a supplemental summary report will be required from the applicant to close open issue 10.
The staff will report results of its review in a future supplement.
18.2 Safety Parameter Display System The October 1985 SER stated that the safety parameter display system (SPDS) was an open issue since additional information was requested, but response has not been received by the staff.
By letter dated December 20, 1985, the applicant provided the requested additional information.
Further information was provided by a phone call on January 16, 1986, and a response was submitted on April 9, 1986.
The staff is continuing its review and will report the result in a future supplement. Open issue 11 remains unresolved.
l Beaver Valley 2 SSER 1 18-1 1
V 19 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS The Advisory Committee on Reactor Safeguards (ACRS) met on November 7-9 to review, among other things, the application of Duquesne Light Company, et al.
for a license to operate the Beaver Valley Power Station, Unit 2.
Before that meeting, members of the ACRS Subcommittee on Beaver Valley toured the facility on October 31, 1985, and met on November 1, 1985, to discuss the application.
. Results of the ACRS review are documented in its letter dated November 13, 1985.
A copy of that letter is included in this supplement as Appendix H.
In response to the ACR5 letter, the applicant submitted a letter dated January 9, 1986.
The staff responded to the ACRS letter on February 20, 1986 (also included in Appendix H).
Staff response stated that "the staff is preparing a supplement to the Beaver Valley Unit 2 Safety Evaluation Report (SER) which will speci-fically address the issues raised in the November 13, 1985 ACRS report." How-ever, the staff is continuing its review of the ACRS concerns and will provide information on resolution of the following items in future supplements to the SER:
(1) steam generator overfill and (2) application of leak-before-break criteria to balance-of plant piping.
i Beaver Valley 2 SSER 1 19-1
APPENDIX A CONTINUATION OF CHRON0 LOGY OF NRC STAFF RADIOLOGICAL REVIEW 0F BEAVER VALLEY POWER STATION, UNIT 2 September 30, 1985 Letter from applicant regarding implementation of 10 CFR 50.62, ATWS mitigation system.
October 8, 1985 Letter to applicant transmitting Final Environmental Statement.
October 8, 1985 Letter from applicant transmitting additional informa-tion on cable spreading room fire protection.
October 9, 1985 Letter from applicant transmitting Revision 1 to the Beaver Valley 2 Special Nuclear Material License application.
October 10, 1985 Letter from applicant transmitting WHIPJET program (an alternate pipe rupture protection program).
October 11, 1985 Letter to applicant transmitting advance copy of Safety Evaluation Report.
October 11, 1985 Letter to applicant transmitting schedular exemption to GDC 4 for primary coolant piping.
October 15, 1985 Letter from applicant transmitting analysis on control system failures.
October 22, 1985 Meeting on fire protection deviations, as a result of applicant's letter of March 27, 1985.
October 22, 1985 Letter to applicant transmitting 20 copies of Safety Evaluation Report.
October 30, 1985 Letter from applicant transmitting revised Cross-Training Program.
November 1, 1985 ACRS Subcommittee on Beaver Valley 2 meeting.
November 4, 1985 Letter from applicant transmitting additional information on Salem ATWS event items 4.2.1 and 4.2.2.
November 7-9, 1985 ACRS meeting on Beaver Valley 2 review.
November 13, 1985 Letter from Shaw, Pittman, Potts and Trowbridge regarding the proposed affiliation of Cleveland Electric Illuminat-ing Company and Toledo Edison Company.
Beaver Valley 2 SSER 1 1
Appendix A
November 13, 1985 ACRS letter addressing its review of Beaver Valley Unit 2.
November 15,.1985 Letter from applicant transmitting information on TMI Action Plan Item II.K.3.5, " Automatic Trip of Reactor Coolant Pumps."
November 22, 1985 Letter to applicant requesting revision of FSAR to remove steam generator high level trip system as part of ESF actuation.
November 22, 1985 Letter to applicant transmitting ACRS's November 13, 1985, letter, and requesting action.
November 22, 1985 Letter to applicant inquiring if it plans to perform independent design verification.
November 26, 1985 Letter to applicant requesting additional information on design documentation review.
December 2, 1985 Letter from applicant transmitting Summary Report on detailed control room design review.
December 20, 1985 Letter from applicant transmitting additional information on safety parameter display system.
December 20, 1985 Letter from applicant transmitting response to SER con-firmatory issue 44, quality assurance.
December 20, 1985 Letter from applicant transmitting Revision 3 to proposed technical specifications.
December 20, 1985 Letter from applicant committing to change FSAR as requested in the staff's November 22, 1985, letter.
December 24, 1985 Letter from applicant providing comments on SER of October 1985.
December 26, 1985 Letter from applicant transmitting Preservice Inspection Program.
December 26, 1985 Letter to applicant informing of assignment of Peter Tam as both Unit 1 and Unit 2 project manager.
January 3, 1986 Letter from applicant submitting WCAP-11004 and WCAP-11005, concerning MULTIFLEX 3.0.
January 9, 1986 Letter from applicant stating position regarding pressure isolation valve.
January 9, 1986 Letter from applicant responding to ACRS concerns as ex-pressed in ACRS letter of November 13, 1985.
January 10, 1986 New project manager's first meeting with Beaver Valley Unit 2 personnel (Summary dated January 24, 1986).
Beaver Valley 2 SSER 1 2
Appendix A
January 24, 1986 Letter to applicant requesting site visit to conduct control room design review.
January 30, 1986 Letter to applicant approving use of ASME Code Case N-32-3.
January 30, 1986 Letter to applicant transmitting FEMA evaluation of Beaver Valley Unit 2 alert and notification system.
1 February 4, 1986 Letter from applicant certifying distribution of FSAR Amendment 11.
February 5, 1986 Letter from applicant transmitting Unit 1-Unit 2 operational difference analysis.
February 11-12, 1986 Detailed control room design review audit (summary dated February 25,1986).
February 20, 1986 NRR staff responded to ACRS's November 13, 1985, letter.
February 28, 1986 Meeting with applicant to discuss independent design verification (summary dated March 10, 1986).
March 3, 1986 Letter to applicant advising of staff's position on use of leak-before-break assumption to balance-of plant piping.
March 3, 1986 Letter to applicant requesting additional information on inadequate core cooling instrumentation.
March 3, 1986 Letter to applicant requesting additional information on design documentation review.
March 3, 1986 Letter from applicant providing supplemental information on accumulator pressure and level indication.
March 4, 1986 First meeting with applicant at NRC office to discuss status of WHIPJET (summary dated March 18, 1986).
March 7, 1986 Letter from applicant providing evaluations to organiza-tional changes to the QA program.
March 10, 1986 Letter from applicant providing additional information on reactor coolant pump trip criteria.
March 10, 1986 Letter from applicant providing additional information on steam generator overfill.
March 10, 1986 Letter to applicant transmitting environmental assessment regarding extension of construction permit.
March 12, 1985 Second meeting with applicant to discuss status of' WHIPJET (summary dated March 31, 1986).
March 14, 1986 Letter to applicant approving withdrawal of commitment to install steam leakage collection system.
Beaver Valley 2 SSER 1 3
Appendix A
. March 14, 1986 Letter to applicant transmitting order to extend construction permit to December 31, 1986.
-March 17, 1986 Amendment 2 of construction permit issued to account for exemption granted on November 11, 1985.
March 19, 1986 Letter from applicant transmitting additional information as a result of NRC staff site visit of February 12, 1986.
March 20, 1986 Letter from applicant transmitting status of outstanding issues.
March 26, 1986 Letter from applicant transmitting revised log sheet to document periodic checks of diesel generator rocker arm lube oil reservoir.
March 27, 1986 Letter to applicant transmitting reports on probabilistic risk assessment insights.
April 1,1986 Letter from applicant transmitting information on control room isolation.
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Beaver Valley 2 SSER 1 4
Appendix A
APPENDIX B BIBLIOGRAPHY Duquesne Light Company et al., " Beater Valley Power Station Unit No. 2 Safety Analysis Report," Docket No. 50-412.
Code of Federal Regulations, Title 10, " Energy," U.S. Government Printing Office, Washington, D.C. (contains general design criteria).
Denton, H. R., NRC, letter to all power reactor applicants and licensees,
" Qualifications of Reactor Operators," March 28, 1980.
U.S. Nuclear Regulatory Commission, NUREG-0452, " Standard Technical Specifica-tions for Westinghouse Pressurized Water Reactors," various revisions.
--, NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980; Supplement 1, January 1983.
--,. NUREG-0800, " Standard Review Plan for Review of Safety Analysis Reports for Nuclear Power Plants - LWR Edition," July 1981 (contains branch technical positions).
Beaver Valley 2 SSER 1 1
Appendix B
APPENDIX 0 ACRONYMS AND INITIALISMS ACRS Advisory Committee on Reactor Safeguards ASME American Society of Mechanical Engineers ATWS anticipated transients without scram BNL Brookhaven National Laboratory BTP Branch Technical Position CRGR Committee for Review of Generic Requirements DG diesel generator DLC Duquesne Light Company ESF engineered safety feature ESP emergency shutdown panel FSAR Final Safety Analysis Report GDC General Design Criterion (a)
NRC Nuclear Regulatory Commission PGP procedures generation package PIV pressure isolation valve QA quality assurance RG Regulatory Guide RHRS residual heat removal system RTS reactor trip system SER Safety Evaluation Report SI safety injection SRP Standard Review Plan SSER Supplemental Safety Evaluation Report
.SRP Standard Review Plan SSI soil-structure interaction Beaver Valley 2 SSER 1 1
Appendix D
s APPENDIX E NRC STAFF CONTRIBUTORS AND CONSULTANTS This supplement is a product of the joint efforts of NRC staff reviewers and consultants.
Principal staff reviewers and consultants who contributed to this supplement are:
NRC STAFF Reviewer's Name Title Review Branch (Division)* **
H. Brammer Senior Mechanical Engineering (PWR Licensing A)
Engineer F. Burrows Electrical Engineer Electrical, Instrumentation &
Control Systems (PWR Licensing A)
N. Chokshi Reliability & Risk Reliability & Risk Assessment Analyst (Safety Review & Oversight)
N. Dudley Lead Reactor Engineer Region I R. Eckenrode Human Factors Electrical, Instrumentation &
Engineer Control Systems (PWR Licensing A)
J. Joyce Senior Task Manager Reactor Safety Issues (Safety Review & Oversight)
S. Lee Materials Engineer Engineering (PWR Licensing A)
S. Rhow Electrical Engineer Electrical, Instrumentation &
Control Systems (BWR Licensing)
N. Romney Mechanical Engineer Engineering (fWR t.icensing A)
D. Shum Reactor Systems Facility Operations (BWR Licensing)
Engineer J. Spraul Quality Assurance Quality Assurance (Quality Assurance, Engineer Vendor, Technical Training Center Programs)t
- List reflects changes made in Office of Nuclear Reactor Regulation (NRR) organization since SER was issued.
- NRR, except where otherwise noted.
10ffice of Inspection and Enforcement.
Beaver Valley 2 SSER 1 1
Appendiy. E
NRC STAFF (Continued)
Reviewer's Name Title Review Branch (Division)* **
S. L. Wu Nuclear Engineer Reactor Systems (PWR Licensing A)
CONSULTANTS Name Laboratory C. J. Costantino Brookhaven National Laboratory C. A. Miller Brookhaven National Laboratory A. J. Philippacopoulos Brookhaven National Laboratory J. W. Stoffel EG&G Idaho, Inc.
A. C. Udy EG&G Idaho, Inc.
- List reflects changes made in Office of Nuclear Reactor Regulation (NRR) organization since SER was issued.
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- NRR, except where otherwise noted.
f Beaver Valley 2 SSER 1 2
Appendix E
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APPENDIX G EG&G IDAH0, INC., TECHNICAL EVALUATION REPORT EGG-EA-6865 ON C#NFORMANCE TO REGULATORY GUIDE 1.97 r
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l Beaver Valley 2 SSER 1 Appendix G
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CONFORMANCE T0 REGULATORY GUIDE 1.97 BEAVER VALLEY POWER STATION, UNIT NO. 2 J. W. Stoffel A. C. Udy Published November 1985 EG&G Idaho, Inc.
Idaho Falls, Idaho 83415 Prepared for the U.S.-Nuclear Regulatory Commission Washington, D.C.
20555 Under 00E Contract No. DE-AC07-76ID01570 FIN No. A6493 Beaver Valley 2 SSER 1 Appendix G
ABSTRACT This EG&G Idaho, Inc., report reviews the submittals for Regulatory Guide 1.97 for Unit No. 2 of the Beaver Valley Power Station and ioentifies areas of nonconformance to the regulatory guide. Exceptions to Regulatory Guioe 1.97 are evaluated and those areas where sufficient basis for acceptability is not provided are ioentified.
FOREWORD This report is supplied as part of the " Program for Evaluating Licensee / Applicant Conformance to RG 1.97," being conducted for the U.S.
Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Division of Systems Integration, by EG&G Idaho, Inc., NRR and I&E Support Branch.
The U.S. Nuclear Regulatory Commission funded the work under authorization B&R 20-19-40-41-3.
Docket No. 50-412 Beaver Valley 2 SSER 1 ji Appendix G
CONTENTS A B S TR A C T..............................................................
11 FOREWORD..................,...........................................
ii 1.
INTRODUCTION.....................................................
1 2.
REVIEW REQUIREMENTS..............................................
2 3.
EVALUATION.......................................................
4 3.1 Adherence to Regulatory Guide 1.97.........................
4 3.2 Type A Variables...........................................
4 3.3 Exceptions to Regulatory Guide 1.97........................
5 4.
CONCLUSIONS......................................................
16 S.
REFERENCES.......................................................
17 Beaver Valley 2 SSER 1 iii Appendix G
CONFORMANCE TO REGULATORY GUIDE 1.97 BEAVER VALLEY POWER STATION, UNIT NO. 2 1.
INTRODUCTION On December 17, 1982, Generic Letter No. 82-33 (Reference 1) was issueo by D. G. Eisenhut, Director of the Division of Licensing, Nuclear Reactor Regulation, to all licensees of operating reactors, applicants for operating licenses and holders of construction permits. Tnis letter included aaditional clarification regarding Regulatory Guiae 1.97, Revision 2 (Reference 2), relating to the requirements for emergency response capability. These requirements have been published as Supplement No. I to NUREG-0737, "TMI Action Plan Requirements" (Reference 3).
Duquesne Light, the applicant for Unit No. 2 of the Beaver Valley Power Station, responded to the generic letter with a letter dated April 15, 1983 (Reference 4). A letter dated September 12, 1983 (Reference 5), provides a review of the instrumentation provided for Regulatory Guide 1.97. Additional information was provided on June 28, 1985 (Reference 6).
This report provides an evaluation of these submittals.
1 Beaver Valley 2 SSER 1 1
Appendix G
2.
REVIEW REQUIREMENTS Section 6.2 of NUREG-0737, Supplement No. 1, sets forth the documentation to be submitted in a report to the f;RC describing how the applicant complies with Regulatory Guide 1.97 as applied to emergency response facilities. The submittal should include documentation that provides the following information for each variable shown in the applicable table of Regulatory Guide 1.97.
1.
Instrument range 2.
Environmental qualification 3.
Seismic qualification 4.
Quality assurance 5.
Redundance ano sensor location 6.
Power supply 7.
Location of display 8.
Schedule of installation or upgrade Furthermore, the submittal should identify deviations from the regulatory guide and provide supporting justification or alternatives.
Subsequent to tne issuance of the generic letter, the NRC held regional meetings, in February and March 1983, to answer licensee and applicant questions and concerns regarding the NRC policy on this subject.
At these meetings, it was noted that the NRC review would only address exceptions taken to Regulatory Guide 1.97.
Furthermore, where licensees or applicants explicitly state that instrument systems conform to the regulatory guide, it was noted that no further staff review would be Beaver Valley 2 SSER 1 2
Appendix G
necessary. Therefore, this report only addresses exceptions to Regulatory Guide 1.97.
The following evaluation is an audit of the applicant's
~
submittals based on the review policy described in the NRC regional meetings.
Beaver Valley 2 SSER 1 3
Appendix G
~
3.
EVALUATION This evaluation is based on the following applicant submittals: the applicant's response to Generic Letter 82-33 dated April 15, 1983, the applicant's response to Section 6.2 of the generic letter dated September 12, 1983, the additional information submitted on June 28, 1985, and the Final Safety Analysis Report (FSAR-Reference 7).
3.1 Adherence to Regulatory Guide 1.97 The applicant states, in Table 1.8-1 of the FSAR, that Unit No. 2 of the Beaver Valley Power Station meets the intent of Regulatory Guide 1.97.
This statement was reaffirmed in the applicant's response dated September 12, 1983. Therefore, we conclude that the applicant has provided an explicit commitment on conformance to Regulatory Guide 1.97.
Exceptions to and deviations from the regulatory guide are noted in Section 3.3.
3.2 Type A Variables Regulatory Guioe 1.97 does not specifically identify Type A variables, i.e., those variables that provide information required to permit the control room operator to take specific manually controlled safety actions.
The applicant classifies the following instrumentation as Type A.
1.
Reactor coolant system (RCS) cold leg water temperature 2.
RCS hot leg water temperature 3.
RCS pressure t
4.
Core exit temperature 5.
Degrees of subcooling i
Beaver Valley 2 SSER 1 4
Appendix G
6.
Containment sump water level-wide range 7.
Containment sump water level-narrow range 8.
Containment pressure 9.
Containment area radiation level-high range
- 10. Pressurizer level
- 11. Steam generator level-wide range
- 12. Steam generator level-narrow range
- 13. Steamline pressure
- 14. Auxiliary feedwater flow
- 15. Primary plant demineralized water storage tank level
- 16. Secondary system radiation The above instrumentation, except degrees of subcooling, meets the Category I recommendations consistent with the requirements for Type A variables. The deviation for degrees of subcooling is addressed in Section 3.3.5.
3.3 Exceptions to Regulatory Guide 1.97 The applicant identified deviations and exceptions from Regulatory Guide 1.97.
These are discussed in the following paragraphs.
1 Beaver Valley 2 SSER 1 5
Appendix G
3.3.1 Neutron Flux In Reference 5, the applicant indicated the installation of nonenvironmentally qualified detectors for this variaDie.
In Reference 6, which superseces the information in Reference 5, the applicant has committed to the installation of detectors that fully meet the recomendations of Regulatory Guide 1.97.
We find the proposed instrumentation acceptable.
3.3.2 Reactor Coolant System Soluble Boron Concentration Regulatory Guide 1.97 recommends instrumentation for this variable with a range of from 0 to 6000 parts per million. The applicant has supplied instrumentation for this variable with a range of 50 to 6000 parts per million. The applicant has not provided justification for this deviation.
The applicant, deviates from Regulatory Guide 1.97 with respect to the range of this post-accident sampling capability. This deviation goes beyond the scope of this review and is being addressed by the NRC as part of their review of NUREG-0737, Item II.B.3.
3.3.3 Reactor Coolant System Cold Leg Water Temperatura Reactor Coolant System Hot Leg Water Temperature Revision 2 of Regulatory Guide 1.97 recommends instrumentation for these variables with ranges of 50 to 750'F. The applicant has supplied instrumentation for these variables with ranges from 0 to 700*F. The applicant presented no justification for the deviations.
Revision 3 of Regulatory Guide 1.97 (Reference 8) recommends a range of 50 to 700*F for these variables. The instrumentation supplied by the applicant meets this range. Therefore, the range supplied by the applicant for these variables is acceptable.
Beaver Valley 2 SSER 1 6
Appendix G
3.3.4 Coolant Level in Reactor Regulatory Guide 1.97 recommends instrumentation for this variable with a range from the bottom of the core to the top of the vessel. The applicant is installing instrumentation for this variable with a range of 0 to 100 percent of the plenum and core height. The applicant has identified this as a deviation. The instrumentation is being installed for the requirements of NRC Generic Letter No. 82-28 (Reference 9).
This deviation goes beyond the scope of this review and is being addressed by the NRC as part of their review of NUREG-0737, Item II.F.2.
3.3.5 Degrees of Subcooling The applicant has identified this as a Type A variable. As such, Table 2 of Regulatory Guide 1.97 recommends Category 1 instrumentation.
The applicant is providing Category 2 instrumentation. Tne NRC is reviewing the acceptability of this variable as part of their review of NUREG-0737, Item II.F.2.
3.3.6 Radiation Level in Circulating Primary Coolant The applicant uses the post-accident sample system, which is'being reviewed by the NRC as part of their review of NUREG-0737, Item II.B.3, to measure tnis parameter. Based on the alternate instrumentation provided by the applicant, we conclude that the instrumentation supplied for this variable is adequate and, therefore, acceptable.
3.3.7 Containment Area Radiation In Reference 5, the applicant indicated that the range of this instrumentation would be determined and supplied later.
In Reference 6, which supersedes the information in Reference 5, the applicant has committed to the installation of instrumentation that fully meets the recommendations of Regulatory Guide 1.97. We find the proposed instrumentation acceptable.
l Beaver Valley 2 SSER 1 7
Appendix G l
l
3.3.8 Effluent Radioactivity-Noble Gases and Vent Flow Rate In Reference 6, the applicant has identified the variable, plant vent radiation level, as a comon plant vent for all release points. This instrumentation complies with the range and Category 2 recommendations of Regulatory Guide 1.97, and is therefore acceptable.
3.3.9 Raciation Exposure Rate (Inside building or areas...)
In Reference 6, the applicant identified this variable as site environmental radiation level, identified the location of the area monitors and shown that the range of the portable and the area monitors satisfies the recommendeo range. Based on this additional information we find the instrumentation provided acceptable.
3.J.10 Residual Heat Removal (RHR) Heat Exchanger Outlet Temperature Revision 2 of Regulatory Guide 1.97 recomends a range of 32 to 350*F for this variable. Revision 3 changed the recommended range to 40 to 350*F. The applicant has supplied instrumentation with a range of 50 to 400*F. The lower limit of the range suppliea does not conform to either revision of the regulatory guide.
We find that this deviation is minor (2.5 percent of the upper limit of the range) with respect to the overall range ana system accuracy. The existing temperature range i adequate to monitor this variable during accident and post-accident conditions.
3.3.11 Accumulator Tank Level and Pressure Accumulator Isolation Valve Position Regulatory Guide 1.97 recommends Category 2 instrumentation for these variables. The applicant did not provide the information required in Section 6.2 of NUREG-0737, Supplement No. I for these variables in Reference 5, stating nonconformance for level and pressure and conformance for the isolation valve status. The justification provided for nonconformance is that these variables are not necessary to monitor the status of the plant while proceeding to a cola shutdown condition.
Beaver Valley 2 SSER 1 8
Appendix G
In Reference 6, the variable isolation valve position is shown to be incluoed under emergency core cooling valve status. No deviations were identified, therefore, the instrumentation for this variable is acceptable.
The existing pressure and level instrumentation is not acceptable. Ac env',ronmentally qualified instrument is necessary to monitor the status of these tanks. The licensee should designate either level or pressure as the key variable to determine accumulator discharge and provide instrumentation, for that variable, that meets the requirements of 10 CFR 50.49.
If accumulator level is selected as the key variable, then the range should be expanded to meet the regulatory guide recommendation.
3.3.12 Boric Acid Charging Flow Regulatory Guide 1.97 recommends instrumentation for this variaole.
The applicant does nct, provide this instrumentation, as it is not a part of the emergency core cooling system, and the boric acid tank is not used as a source of water for safety injection.
The applicant states that the units do not use boric acid charging flow as a safety injection system. Centrifugal charging pump flow, safety injection flow ano residual heat removal flow are the safety injection variables monitored. Therefore, we find that this variable is not applicab'le at the Beaver Valley Power Station, Unit No. 2.
3.3.13 Reactor Coolant Pumo Status In Reference 5, the applicant did not provide the information required by Section 6.2 of NUREG-0737, Supplement No. 1, for this variable.
In l
Reference 6, which supersedes Reference 5, the applicant provided the required information. No deviations were identified, therefore, the instrumentation for this variable is acceptable.
Beaver Valley 2 SSER 1 9
Appendix G
3.3.14 Pressurizer Heater Status In Reference 5, the applicant inoicated that this variable was monitored by circuit breaker position only.
In Reference 6, which supersedes Reference 5, tne applicant identifies instrumentation that reads the total heater bus current in addition to circuit breaker indication. We find that with the provided instrumentation the operator can verify heater bank energization. Therefore, the Regulatory Guide 1.97 recommendations are met.
3 3.15 Quench Tank Level, Temperature ano Pressure In Reference 5, the applicant did not provide the information required by Section 6.2 of NUREG-0737, Supp'enent No.1, for these variables.
In Reference 6, which supersedes Reference 5, the applicant provided the required information. No deviations were identified for the tank level and pressure. Therefore, the instrumentation for these variables is acceptable.
I Regulatory Guide 1.97 recommends instrumentation for quench tank temperature with a range of 50 to 750*F. The applicant has provided a range of 50 to 350*F. The applicant states that the existing range is adequate to determine if abnormal conditions exist.
In addition, the quench tank is provided with two rupture disks with an activation pressure of 100 psi. This prevents the temperature from exceeding the upper range provided. We find this deviation acceptable.
The range ccvers the anticipated requirements for normal operation, anticipatea operational occurrences and accident conditions. This range relates to the tank's rupture disk and the 100 psi tank design pressure that limits the temperature of the tank contents to saturated steam conditions under 350*F. Thus, we find that this deviation from the recommendation of the regulatory guide is acceptable.
Beaver Valley 2 SSER 1 10 Appendix G
f i
3.3.16 Steam Generator Level Regulatory Guide 1.97 recommenos instrumentation for this variable, with a range from the tube sheet to the separator. The applicant has provided instrumentation witn a range from 12 inches above the top of the tube sheet to above the top of the swirl vane cylinder's (Model 51 steam generators).
At 12 inches above the tube sheet, the steam generator is essentially empty. Therefore, this deviation is minor with respect to the overall range and system accuracy. The existing range is adequate to monitor this variable during all accident and post-accident conditions.
3.3.17 Heat Removal by the Containment Fan Heat Removal System Regulatory Guide 1.97 recommends plant ' specific instrumentation for this variable. The applicant, in Reference 5, does not address this variable. Our examination of Section 6.2.2 of the FSAR indicates that Unit 2 of the Beaver Valley Power Station does not have a containment fan heat removal system. This being the case, tnis variable is not needed at this unit.
l 3.3.18 Containment Atmosphere Temperature Regulatory Guide 1.97 recommends Category 2 instrumentation with a range of 40 to 400*F for this variable. The applic' ant has supplied Category 3 instrumentation with a range of 0 to 200*F. The applicant states that the containment atmosphere temperature is not a key variable for accident monitoring; that the key variables for monitoring containment cooling are containment spray flow (Category 2), containment water level (Category 1), containment spray system valve status (Category 2), and containment pressure (Category 1). The applicant further states that immediately after containment spray is initiated, the containment atmosphere is saturated steam and the temperature can be determined based l
on containment pressure.
I Beaver Valley 2 SSER 1 11 Appendix G
We find that the applicant's applicaticn of Category 3 backup instrumentation is in accordance with the regulatory guida. Since containment pressure is an alternate measure of monitoring containment teaperature the existing temperature range is acequate for this variable.
3.3.19 Containment Sumo Water Temperature Regulatory Guide 1.97 reccmmends instrumentation for this variable with a range of 50 to 250'F. The applicant has not provided this instrumentation. The applicant states that containment sump water temperature is not useo for emergency core cooling system operation or assurance of net positive suction head (NPSH). NPSH calculations conservattvely assume saturated water is present.
In addition, the applicant states that Category 1 recirculation spray pump suction line temperature instrumentation is provided and can be utilized to monitor containment sump water temperature in the control room.
We conclude that the alternate instrumentation provided for this variable is acceptable since the recirculation spray pump suction line temperature will be the same as the sump when in operation.
3.3.20 High Level Radioactive Liould Tank Level In Reference 5, the applicant did not provide the information required by Section 6.2 of NUREG-0737, Supplement No. 1, for this variable.
In Peferer.ce 6, wnich supersedes Reference 5, the applicant provided the required information. No deviations were identified, therefore, the instrumentation for this variable is acceptable.
3.3.21 Radioactive Gas Holdup Tank Pressure Regulatory Guide 1.97 recommends instrumentation for this variable with a range of 0 to 150 percent of design pressure. The instrumentation provided has a range of 0 to 100 psig (design pressure). The licensee Beaver Valley 2 SSER 1 12 Appendix G.
states that rupture disks set at 100 psig protect the tank to 100 percent of design pressure and no operator action is predicatea on pressures in excess of 100 percent of aesign pressure.
Tne applicant indicates that tnese tanks have a' design pressure of 100 psig and rupture disks are provided to keep the pressure from exceeding tne design pressure. Based on this, we find the deviation from the recommended range acceptable.
3.3.22 Condenser Air Removal System Exhaust-Noble G93es and Vent Flow Rate In Reference 5, the applicant identified this variable as Category 3 L
with the range to be provided later.
In Reference 6, which supersedes Reference 5, the applicant states that this release path is routed through the common plant vent. This complies with the recommendations of Regulatory Guice 1.97 and is, therefore, acceptable.
L 3.3.23 Vent From Steam Generator Atmospheric Dump Valves 7
In Reference 5, the applicant did not provide the information required by Section 6.2 of NUREG-0737, Supplement No. I for this variable.
In Reference 6, which supersedes Reference 5, the applicant provided the required information. No deviations were identifiec, therefore, the instrumentation for this variable is acceptable.
3.3.24 Information Not Supplied, Type E Variables In Reference 5, the applicant did not provide the information required by Section 6.2 of NUREG-0737, Supplement No. I for the four following b
variables.
o Particulates and Halogens--all identified plant release points o
Airborne radiohalogens and particulates L
1 Beaver Valley 2 SSER 1 13 Appendix G
o Plant and environs radioactivity o
Wind direction In Reference 6, which superseces Reference 9, the applicant provided the required information. No deviations were identified, therefore, the instrumentation for these variables is acceptable.
3.3.25 Plant and Environs Radiation In Reference 5, the applicant indicated that the range for this variable would be provided later.
In Reference 6, which supersedes Reference 5, the applicant provided the instrument ranges. No deviations were identified, therefore, the instrumentation for this variable is acceptable.
3.3.26 Wind Speed Revision 2 of Regulatory Guide 1.97 recommends a range of 0 to 67 miles per hour (mph) for this variable. Revision 3 changes the recommendation to O to 50 mph. The applicant has identified a deviation from the Revision 2 recommendation in that the range provided is 0 to 50 mph. They state that this is acceptable because_it meets Regulatory Guide 1.23 requirements.
We find that the range provided is satisfactory. Furthermore, it meets the recommendation of Revision 3 of Regulatory Guide 1.97.
3.3.27 Estimation of Atmospheric Stability Regulatory Guide 1.97 recornends instrumentation for this variable with a range -9 to +18'F for 164 feet intervals or an analogous range for alternative stability analysis. The applicant has provided instrumentation with a range of -4 to +8'F for a 115 feet interval and -6 to +12*F for a 465 feet interval. The licensee justifies this deviation by saying that Beaver Valley 2 SSER 1 14 Appendix G
the vertical temperature ranges cover the range of lapse rate guidance of Regulatory Guide 1.23 (Reference 10) required to estimate the atmospheric stability class.
Table 1 of Regulatory Guide 1.23 provides 7 vertical atmospheric stability classifications based on the difference in temperature per 100 meters elevation change. These classifications cover from extremely unstable to extremely stable. Any temperature difference greater than +4*C or less than -2*C does nothing to the stability classification. The applicant's instrument accuracy is as specified in Regulatory Guide 1.97, the temperature range ano the vertical separation are both greater than that recommended in Regulatory Guide 1.23.
Therefore, we find that this instrumentation is acceptable to determine the atmospheric stability.
Beaver Valley 2 SSER 1 15 Appendix G
4 CONCLUSIONS Based on our review, we find that the applicant either conforms to or is justified in deviating.from Regulatory Guide 1.97, with the following exception:
.l.
Accumulator tank level and pressure--the applicant should provide a level or pressure instrument for this variable that is environmentally _ qualified in accordance with 10 CFR 50.49.
If the level instrument is the variable environmentally qualified then the range should be expanded to that recommended by the regulatory guide (Section 3.3.11).
i a
Beaver Valley 2 SSER 1 16 Appendix G
...w-
n 5.
REFERENCE S 1.
NRC letter D. G. Eisenhut to All Licensees of Operating Reactors, Applicants for Operating Licenses, and Holders of Construction Permits, " Supplement No. I to NUREG-0737--Requirements for Emer Response Capability (Generic Letter No. 82-33)," December 17,10ency
.82.
2.
_ Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess ylant ana environs Loncitions During ana Following an Accident, Regulatory Guide 1.97, Revision 2, NRC, Office of Standaros Development, Decenber 1980.
3.
Clarification of TMI Action Plan Requirements, Requirements for Emergency Response Capability, NUREG-0737, Supplement No. 1, NRC, Office of Nuclear Reactor Regulation, January 1983.
4.
Duquesne Light Company letter, E. J. Woolever to U.S. NRC, " Generic Letter 82-33, Supplement i to NUREG-0737, Requirements for Emergency Response Capability," April 15, 1983, 2NRC-3-017.
5.
Duquesne Light Company letter, E. J. Woolever to H. R. Denton, NRC,
" Regulatory Guiae 1.97 Implementation Reoort," September 12, 1983, 2NRC-3-072.
6.
Duquesne Light Company letter, J. J. Carey to H. R. Denton, NRC,
" Regulatory Guide 1.97 Implementation Report," June 28, 1985, 2NRC-5-095.
7.
Beaver Valley Power Station Unit 2, Final Safety Analysis Report, Duquesne Light Company, Amendment 4, December 1983.
8.
Instrumentation for Light-Water-Cooled Nuclear Power Plants to 1ssess etant ana tnvirons conditions Uuring anc rollowing an Accicent, Regulatory Guide 1.97, Revision 3, NRC, Office of Nuclear Regulatory Research, May 1983.
9.
NRC Letter, D. G. Eisenhut to all Licensees of Operating Westingnouse and CE PWRs, " Inadequate Core Cooling Instrumentation System (Generic Letter No. 82-28)," December 10, 1982.
- 10. Onsite Meteorological Programs, Regulatory Guide 1.23 (Safety Guioe 23), NRC, February 17, 1972 or Meteorological Programs in Support of Nuclear Power Plants, Proposed Revision 1 to Regulatory Guide 1.23, NRC, Office of Standards Development, September 1980.
37384 Beaver Valley 2 SSER 1 17 Appendix G
4 4
APPENDIX H ACRS REPORT ON BEAVER VALLEY POWER STATION, UNIT 2 (NOVEMBER 13, 1985),
AND STAFF RESPONSE (FEBRUARY 20, 1986) 1 4
4
.i I
Beaver Valley 2 SSER 1
/,ppendix H
/ wsw UNITED STATES 7
1, NUCLEAR REGULATORY COMMISSION f
E ADVISORY COMM6TTEE ON REACTOR SAFEGUARDS
[
W ASHINGTON, D. C. 20555 e
l November 13, 1985 Honorable Nunzio J. Palladino Chairman U. S. Nuclear Regulatory Comission Washington, D. C. 20555
Dear Dr. Palladino:
SUBJECT:
ACRS REPORT ON BEAVER VALLEY POWER STATION, UNIT 2 During its 307th meeting, November 7-9, 1985, the Advisory Comittee on i
Reactor Safeguards reviewed the application of Duquesne Light Company (Applicant), acting on behalf of itself and as agent for Ohio Edison Company, The Cleveland Electric Illuminating Company, and The Toledo Edison Company, for a license to operate the Beaver Valley Power Sta-tion, Unit 2.
The ACRS comented on the construction permit application for the Beaver Valley Power Station, Unit 2 in a report dated December 11, 1973. Members of the ACRS Subcomittee on Beaver Valley toured the' facility on October 31, 1985 and met in Coraopolis, Pennsylvania on November 1,1985 to discuss the application.
During our review, we had the benefit of discussions with representatives and consultants of the Applicant, Stone and Webster Engineering Corporation (SWECO), Westing-house Electric Corporation, and the NRC Staff. We also had the benefit of the documents referenced.
Beaver Valley, Unit 2 is adjacent to Beaver Valley, Unit I and the Shippingport Atomic Power Station (the latter terminated operations in 1982 and is scheduled for decomissioning by the Department of Energy).
Deaver Valley, Unit I uses a three-loop pressurized water reactor (PWR) supplied by Westinghouse with a net output of 810 MWe; it was licensed to operate in January 1976. Beaver Valley, Unit 2 is similar in design to Unit I with a number of improvements and will also use a three-loop PWR with a net output of 836 MWe.
SWECO is the architect-engineer-constructor for both units.
Construction of Unit 2 is about 90 percent complete and the Applicant currently estimates the fuel load date to be April 1987.
The reinforced concrete containment is maintained during normal opera-tion at a subatmospheric pressure.
The containment depressurization systems are designed to reduce the containment temperature and return the containment pressure to a subatmospheric level following a break in either the prin'ary or secondary system piping within the containment.
Redundancy and diversity in the feedwater systems for Unit 2 are pro-vided by two electric motor driven main feedwater pumps, each capable of providing 60 percent of the feedwater flow required for full power Beaver Valley 2 SSER 1 1
Appendix H
Honorable Nunzio J. Palladino November 13, 1985 operation, two electric motor driven auxiliary feedwater pumps of 50 percent of required capacity each, and one steam turbine driven auxil-
)
fary feedwater pump with 100 percent of required capacity.
In addition, Unit 2 has one electric motor driven start-up pump thai: will provide 30 percent of the feedwater flow required for full power operation.
The Applicant has extensive nuclear power plant operating experience and plans to utilize experienced personnel from Unit I to fill key positions in staffing Unit 2.
The Applicant appears to have an effective training program which utilizes a control room simulator specific to Unit I and a control board " mock-up" specific to Unit 2 for training Unit 2 per-sonnel.
Considering the similarity between Unit I and Unit 2 and the associated control boards, we believe that the plant staff can be successfully trained by the Applicant.
During our meeting, the NRC Staff identified a number of open issues that must be resolved prior to the granting of an operating license. We believe that these issues can be resolved in a manner satisfactory to the NRC Staff. We wish to be kept informed.
The Applicant stated that a review of the plant will be conducted in order to assure that there are no unacceptable seismically induced interactions between nonsafety equipment and safety-related systems.
Seismic Category 1 structures (including buried piping) at this plant have been designed for an SSE corresponding to a peak horizontal ground acceleration of 0.1259 at the ground surface. The Committee recomends that the Applicant evaluate the seismic capability of the emergency AC power supplies, DC power supplies, and small equipment such as actuators and instrument lines that are part of the decay heat removal system to assure that adequate safety margins exist.
l The potential for overfilling the steam generatcrs and the satisfactory I
accomodation of the consequential effects remains unanswered.
The NRC Staff has been working with the Applicant to detemine the ability of the instrumentation and control systems to prevent steam generator overfill. However, the capability of the main and auxiliary steam lines to accomodate the consequential effects from overfilling the steam gen-erators from any cause, such as instrumentation failures, steam gen-erator tube ruptures, etc., has not been fully addressed. We recomend that these somewhat interrelated matters receive further study by the licensee and the Staff.
We recognize that USI A-47, " Safety Implica-tions of Control Systems," addresses some aspects of this matter; I
however, there are additional issues which require resolution.
We wish to be kept informed.
The licensee is embarking on an evaluation of alternate pipe rupture protection for the balance of plant beyond the primary system (that is, the application of the leak-before-break criterion to evaluate the need for pipe whip restraints).
This is a departure from previous practice l
Beaver Valley 2 SSER 1 2
Appendix H I
Honorable Nunzio J. Palladino November 13, 1985 and could have benefits, but it must be considered carefully.
We wish to be briefed on this program before implementation.
We believe that, subject to the resolution of open items identified by the NRC Staff and the items mentioned above, and subject to the satis-factory completi(n of construction, staffing, and preoperstional test-ing, there is reasonable assurance that the Beaver Valley Power Station, Unit 2 can be operated at power levels up to 2652 MWt without undue risk to the health and safety of the public.
Sincerely, David A. Ward Chaiman
References:
1.
Duquesne Light Company, " Final Safety Analysis Report Beaver Valley Power Station, Unit 2," with Amendments 1-10 2.
U. S. Nuclear Regulatory Comission, " Safety Evaluation Report Re-lated to the Operation of Beaver Valley Power Station, Unit 2."
USNRC Report NUREG-1057, dated October 1985 3.
Undated letter from a member of the public regarding Beaver Valley Nuclear Power Station, Unit 2 Beaver Valley 2 SSER 1 3
Appendix H l
parta q
'o,,
UNITED STATES 8
NUCLEAR REGULATORY COMMISSION o
G E
WASHINGTON, D. C. 20555 c
e
%*****/
February 20, 1986 MEMORANDUM FOR:
Raymond F. Fraley, Executive Director Advisory Committee on Reactor Safeguards
)
FROM:
Victor Stello, Jr.
Acting Executive Director for Operations
SUBJECT:
307th ACRS MEETING (N0VEMBER 7-9, 1985) F0LLOW-UP ITEMS
REFERENCE:
Memo from R. Fraley to W. Dircks, subject as above, dated November 25, 1985 The following information is provided in response to those specific items in the referenced memorandum that pertain to the Office of Nuclear Reactor Regulation.
1.
ACRS Report on Beaver Valley Power Station, Unit 2 The staff is preparing a Supplement to the Beaver Valley Unit 2 Safety Evaluation Report (SER) which will specifically address the issues raised in the November 13, 1985 ACRS report. This supplement is scheduled for issuance in March 1986. The current estimate of construction completion for this plant is April 1987.
In response to the steam generator overfill issue, a number of actions have been completed or are in progress. The staff identified the need for more information on steam generator tube rupture (SGTR) as a confirmatory item in the Beaver Valley SER. The Ceaver Valley Unit 2 FSAR contains an analysis for system response and radiological con-sequences of SGTR which assumes leak termination in 30 minutes. The staff requested the applicant to justify this time period.
In response, the applicant stated that the Westinghouse Owners Group is investigating several SGTR licensing concerns and will address the staff's concerns through generic evaluations. When this additional information is received, the staff will complete its review of this confirmatory ! tem and report the results of the review in a supplement to the SER.
The Beaver Valley SER (Section 7.3.3.12) discusses steam generator level control and protection as a backfit issue.
In accordance with the staff's plant-specific backfit procedure, a backfit appeal meeting was requested by the applicant. This meeting was held on May 9, 1985 with the Director, Division of Licensing. As stated in the SER, a conclusion was reached that, if the applicant did not make hardware modifications to the steam generator level control system, there was no undue risk to the public health and safety for the interim period it will take to Beaver Valley 2 SSER 1 4
Appendix H
I resolve USI A-47 (" Safety Implications of Control Systems"). The applicant will be required to meet any requirements that evolve from the resolution of USI A-47 to allay steam generator overfill concerns.
On a generic basis the staff is initiating a separate generic safety issue on causes of steam generator overfill other than instrument and control system failures.
In conjunction with the staff's effort, the Westinghouse Owners Group has submitted Topical Report WCAP-10698 "SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill". This document predicts the margin to overfill for various models of steam generators assuming occurrence of the design basis SGTR.
The necessary operator actions were considered and the report states that operator action times were based partially on simulator runs. The analysis assumed a concurrent loss of offsite power and sensitivity analyses were performed to detennine the effects of a spectrum of single active failures. The staff has performed a preliminary review of this analysis, which indicates that the methodology is reasonable but that the assumed operator action times appear optimistic. The WOG plans to submit a supplement to this report which will deal with the consequences of overfill, including steam line dynamic load analyses, the effects on the operability of secondary safety valves and the con-sequences of safety valve failures. Upon receipt of this document (tentatively scheduled for February 1986) the staff will initiate a detailed review of WCAP-10698 and the overfill supplement. Satisfactory resolution of the generic steam generator overfill problem for Westinghouse plants and acceptable results for the offsite radiological consequence analyses would satisfy the staff's concerns regarding the effects of overfill as a result of SGTR. As far as other causes of overfill, the staff has included an assessment of instrument failures as a potential cause of steam generator overfill as part of USI A-47.
The ACRS report states that the Comittee wishes to be briefed on the application of leak-before-break criteria to certain Beaver Valley Unit 2 balance-of-plant piping systems "before implementation".
Implementation of the engineering and design aspects of this program had already begun at the time of the 307th ACRS meeting and will be a continuing effort up until the time of plant completion. The staff plans to meet with the licensee at least quarterly to discuss the status of this program as it develops.
Implementation as applied to actual construction would be in the form of not installing certain pipe whip restraints and jet impingement barriers presently required by NRC regulations. Therefore, before an operating license can be issued, the staff must either change the regulations via rulemaking or grant an exemption to portions of this regulation. The staff is in the process of proposing a change to 10 CFR 50 Part A, General Design Criteria No. 4 to allow application of leak-before-break technology to Beaver Valley 2 SSER 1 5
Appendix H
. piping systems under certain conditions. This proposed rule was discussed with the CRGR on January 22, 1986 and will be discussed with the ACRS Metal Components Subcomittee on February 27-28, 1986.
Briefings of the ACRS on the Beaver Valley Unit 2 program will be scheduled, as appropriate, prior to licensing the plant.
f
- & }-
or tello, Jr.
Acting Execut" ve Director for Operations Beaver Valley 2 SSER 1 6
Appendix H
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' APPENDIX I c
BROOKHAVEN NATIONAL LABORATORY TECHNICAL EVALUATION REPORT ON S0IL STRUCTURE ANALYSIS l'
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Beaver Valley Containment, Unit 2 SSI Evaluation by A.J. Philippacopoulos, C.A. Miller and C.J. Costantino Structural Analysis Division Department of Nuclear Energy Brookhaven National Laboratory Upton, NY 11973 September 12, 1985 Beaver Valley 2 SSER 1 i
Appendix I
1.0 INTRODUCTION
The BVPS-2 seismic analysis was originally performed using two finite element codes, PLAXLY and FLUSH. During a structural design audit ' conducted by the NRC staff.on January 31 through February 3,1984 several action items were identified. One of the action items directed the applicant to " perform soil-structure interaction analysis.... to show that the intent of SRP 3.7.2.4 is met".
The concern was raised because a lumped parameter SSI analysis with the criteria motion input at the base of the structure was not performed. The PLAXLY and FLUSH results were obtained based on the criteria motion applied at the surface with convolution methods used to obtain the motion at depths of interest (e.g., the facility foundation).
The applicant responded to this question by performing a "three-step" interaction analysis (see DLC 2NRC-5-016 dated February 1,1985).
This analysis is a half-space type analysis where the free field is modeled as a continuum rather than with finite elements such as with ' LAXLY and FLUSH. The seismic input to this model is identical to that used for PLAXLY and FLUSH.
This new analysis therefore does not represent a case where the criteria motion is applied at the foundation level of the facility.
The question of applying the criteria motion at the foundation level was addressed by the applicant in two respects. First, they developed site specific spectra reflecting the actual site properties (Site Dependent Response Spectra, by SWC, February 1985). These spectra were then shown to be less than the motion used as input to their SSI models (DLC letter 2NRC-5-114 dated May 7, 1985). Second, the applicant assembled a panel of six experts to review their SSI methodology. The panel agreed with the applicant's approach (DLC letter 2NRC-5-049, dated March 25,1985).
t Beaver Valley 2 SSER 1 1
Appendix I
On June 19-20, 1985 an audit of this work was performed by the NRC staff (with the assistance of BNL) at the Stone and Webster offices in Boston, Massachusetts. The following is a summary of the items reviewed during this audit:
2.0 PLAXLY ANALYSIS The PLAXLY analysis is contained in Calc No. 12241-NS(B)-045-JD entitled
" Soil-Structure Interaction of Containment". A stick model of the containment was coupled to a plane strain finite element model of the free field. The criteria motion was input at the surface and convoluted to the depth of bedrock at about 115 feet below the surface.
Several analyses were performed with the soil properties modified to reflect local "hard spots" that occurred because of grouting of the material during construction. The final response spectrum were taken as the envelop of the spectrum for all of the cases.
Results obtained with PLAXLY were compared with those obtained using FLUSH and the "Three-Step" method and found to be in general agreement.
The PLAXLY models and analysis methods were found to be consistent with standard practice and one would expect the results to be reasonable.
3.0 FLUSH ANALYSIS The FLUSH model run was similar to the PLAXLY model described above, namely, a containment stick model coupled to a finite element model of the free field. This soil model extended to a depth of about 115', down to bedrock. The horizontal criteria motion was input at the ground surface, convoluted down to the bedrock level and then input to the base of the FLUSH finite element mesh. Although no specific written report.of this FLUSH model was reviewed, the microfiche containing copies of the computer run was available ffnd reviewed.
Beaver Valley 2 SSER 1 2
Appendix I
Spectra output from the FLUSH run was generated at several depths during the convolution phase of the program. The audit team requested that the spectra generated at the level of the base of the foundation during the convolution be plot *ed and compared with the results of the KINACT c'omputer run. This was done and presented by the applicant in Document No. 2NRC-5-114.
The two spectra are in general agreement except in the range of about 3 to 8 cps.
In this range, the FLUSH spectra lies below the KINACT output. Both of these spectra, however, are above the site specific spectra.
4.0 THREE-STEP ANALYSIS The applicant performed a soil structure interaction study which is based on the so-called "Three-Step" method. The latter type of analysis is basically a form of the general substructure method used in the SSI area.
During the audit (June 19-20,1985) several documents were reviewed related to the three-step analysis:
- 1. " Stiffness Functions of Rectangular Foundations (REFUND)", ST-e32, February 1977.
- 2. Calc No.12241-NS(B)-140, Verification of the Computerized Calculation " EMBED".
- 3. Calc No.12241-NS(B)-141, Verification of Computerized Calculation KINACT2.
- 4. Frequency Response Interaction Dynamic Analysis (FRIDAY), ST-243, September 1972.
- 5. Calc. No.12241-NS(B)-161, Seismic Analysis of Containment (Three-Step Method).
Beaver Valley 2 SSER 1 3
Appendix I
The three steps followed by the applicant in the soil structure interaction studies for the containment structure are:
a) Calculation of foundation impedances b) Calculation of foundation input c) Structural response calculations.
The calculation of the impedance functions was performed in two steps.
In the first, impedances were calculated for the corresponding surface foundation.
Since the containment is embedded into the soil, these impedances were subsequently modified to account for embedment effects. The impedances associated with the surface foundation case, were computed through the REFUND code. The latter handles the harmonic response of rigid rectangular footings resting on a viscoelastic layered medium. This is done by a discretization of the layered medium with provisions made for the transmitting boundaries of the mesh. A verification study was performed, which demonstrates that results from the REFUND code compare well with analytical results by Luco (J.E. Luco,
" Impedance Functions for a Rigid Foundation on a Layered Medium", Journal of Nuclear Engineering and Design, Vol. 2, 1974).
It may be noted that in this work by Luco, relaxed boundary conditions were assumed at the footing-soil interface whereas the REFUND code assumed welded type. Furthermore, in the particular publication sited above, the shape of the foundation is circular, whereas the results from REFUND were obtained from an equivalent rectangular footing.
In spite of the differences mentioned above the answers from the analytical method do not differ much from those obtained by REFUND which is a numerical method.
This demonstrates that the results obtained by REFUND for this case are statisfactory.
Modification of the results from the REFUND code in order to account for embedment, was done by the program EMBED. The latter, is based on a simplified procedure, according to which correction factors are applied to the impedances for surface foundations in order to obtain corresponding impedanceffortheembedmentcase. The basis for the EMBED code is described i
Beaver Valley 2 SSER 1 4
Appendix I
r in the paper:
Kausel, et al., "The Spring Method for Embedded Foundations",
Vol. 48, Journal of Nuclear Engineering and Design,1978. The applicant performed a verification study which is described in Calc. No.
12241-NS(B)-140.
In this study, results from EMBED computer code were compared with hand computations. A review of this calculations reveals that both results are in agreement.
The second step i.e., determination of the foundation input motion, was done by the KINACT code. The procedure followed is based on the derviation of the input motion of the foundation level of the containment building including the effects due to kinematic interaction. Based on this, from the surface motion at the free field a set of two input time histories were computed at the foundation level of the containment i.e., a translational and a rotational. During the audit Stone & Webster was requested to provide the phasing between the translational and the rotational components of the input.
It was concluded that the rotational component is basically additive to the translational one.
The third step i.e., containment response calculation was performed using the FRIDAY code.
During the audit the document ST-243, listed above, was reviewed. Furthermore, comparisons between FRIDAY-STARDYNE were reviewed and found to be good.
Essentially the results from these two codes are the same for the particular model used in the verification study.
The three-step analysis performed by the applicant for the Containment Building, Unit 2 follows the general formulation of the well-known substructure methodolcgy. The generation of the impedances a re done using a discretization of the foundation medium.
Comparisons with analytical solutions were made. Analytical data compared well with numerical predictions by REFUND. The procedure followed in order to modify the impedances for embedment effects is based on engineering simplifications of a rathe.r complex proflem and gives good results for a wide range of applications.
Beaver Valley 2 SSER 1 5
Appendix I
With respect to the foundation input motion, the procedure followed by the applicant is based on sound theoretical principles. Moreover, as indicated by the applicant, the effect from the rotational componegt is generally additive to the translational one.
Finally, the procedure followed to compute the building response is based on standard techniques.
It is therefore concluded that the three-step approach, which was applied for the SSI response evaluation of the containment structure, is acceptable.
5.0 SITE SPECIFIC The methodology used to generate site specific spectra was presented by SWC but was not reviewed in detail. This is being done by Geosciences Branch of NRC as a separate effort.
It is important to note that any results of this audit are dependent of the acceptance of the SWC site spekific study.
Comparisons were made between the site specific spectra that would be input at foundation level to that which was actually input for the PLAXLY, FLUSH, and Three-Step analyses. The input that was used is higher than the site specific spectra at the foundation level except for a small region around s
3 cps where the site specific spectra is slightly higher than that used for input to the SSI analysis. The design floor response spectra all have valleys at this frequency range. The smoothing techniques used by SWC in generating design spectra from calculated spectra raised the spectra at this frequency range.
It is therefore concluded that the design spectra used for BVPS-2 are conservative relative to spectra that would be generated based on site specific input spectra.
Beaver Valley 2 iSER J 6
Appendix I
6.0 HORIZONTAL SURFACE ASSUMPTION During the meeting, the audit team raised a question as to the adequacy of the assumption implicit in all the analyses performed (PLAXLY, FLUSH, Three-Step Hethod); namely, that the ground surface at the site is norizontal. The site is located along the banks of the Ohio River, with nearby hills located to the south of the plant. The audit team asked that a cross-section of the site bt drawn to scale, this cross-section being taken perpendicular to the slope. The results of this plot indicate that the ground surface rises approximately 150 fe..in a distance of about 1500 feet. Within a distance of about 4 building di.ieters from the centerline of the containment building, the soil / rock irofile is essentially horizontal.
Therefore, the assumption used in the,alyses is satisfied.
7.0 CONCLUSION
The soil structure inte action studies preformed by Stone & Webster for the Beaver Valley Containment, Unit 2 are found to be acceptable. This conclusion is based on the review of procedures and calculations presented during the audit of June 19-20, 1985 and the subsequent information provided by Stone & Webster.
It is pointed out, however, that this conclusion is a.
subjected to the condition tl}t the site specific study is acceptable. The latter study is being reviewed by Geosciences Branch of NRC as a separate effort and will be addressed in the NRC staff Safety Evaluation Report.
Beaver Valley 2 SSER 1 7
Appendix I l
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APPENDIX J STAFF SAFETY EVALUATION REPORT ON REACTOR TRIP SYSTEM RELIABILITY, ITEMS 4.1, 4.2.1, AND 4.2.2 0F GENERIC LETTER 83-28 I
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j Beaver Valley 2 SSER 1 Appendix J r
r SAFETY EVALUATION REPORT BEAVER VALLEY POWER STATION, UNIT 2 REACTOR TRIP SYSTEM RELIABILITY ITEMS 4.1, 4.2.1 AND 4.2.2 0F GENERIC LETTER 83-28 1.
INTRODUCTION On July 8,1983, the Nuclear Regulatory Commission (NRC) issued Generic Letter (GL) 83-28. This letter addressed intermediate-term actions to be taken by licensees and applicants aimed at assuring that a comprehensive program of preventive maintenance and surveillance testing is implemented for the reactor trip breakers (RTBs) in pressurized water reactors.
In particular, Item 4.1 of the letter required licensees and applicants to verify that all vendor-recommended reactor trip breaker modifications have been implemented.
Item 4.2 required them to submit a description of their preventive maintenance and surveillance program to ensure reliable reactor trip breaker operation. The description of the submitted program was to include the following:
GL, Item 4.1 All vendor-recommended reactor trip breaker modifications shall be reviewed to verify that either:
~
each modification has, in fact, been implemented, or a written evaluation of the technical reasons for not implementing a modifications exists.
GL, Item 4.2.1 A planned program of periodic maintenance, including lubrication, housekeeping, and other items recommended by the equipment supplier.
GL, Item 4.2.2 Trending of parameters affecting operation and measured during testing to forecast degradation of operation.
l Beaver Valley 2 SSER 1 1
Appendix J
Duquesne Light, the applicant for Beaver Valley 2, submitted responses to the Generic Letter on March 30, 1984, and November 4,1985. This report presents an evaluation of the adequacy of the applicant's responses and of his preventive maintenance and surveillance programs for RTBs.
2.
EVALUATION CRITERIA 2.1 Periodic Maintenance Program The primary source for periodic maintenance program criteria is Westinghouse Maintenance Program Manual for DS-416 Reactor Trip Circuit Breakers, Rev. O.
This document was prepared for the Westinghouse Owners Group and is the breaker manufacte er's recommended maintenancc program for the DS-416 breaker.
It provides specific direction with regard to schedule, inspection and testing, cleaning, lubrication, corrective maintenance and record keeping. The document was reviewed to identify those items that contribute to breaker trip reliability consistent with the generic letter. Those items identified for maintenance at six month intervals (or when 500 breaker operations have becn counted, whichever comes first) that should be included in the applicant's RTB maintenance program are:
1.
General inspection to include checking of breaker's cleanliness, all bolts and nuts, pole bases, arc chutes, insulating link, wiring and auxiliary switches; 2.
Retaining rings inspection, including those on the undervoltage trip attachment (UVTA);
3.
Arcing and main contacts inspection as specified by the Westinghouse Maintenance Manual; 4.
UVTA check as specified by the Westinghouse Maintenance Manual, including replacement of UVTA if dropout voltage is greater than 60% or less than 30% of rated UVTA coil voltage; Beaver Valley 2 SSER 1 2
Appendix J
5.
Shunt Trip Attachment (STA) check as specified by the Westinghouse Maintenance Manual; 6.
Lubrication as specified by the Westinghouse Maintenance Manual; 7.
Functional check of the breaker's operation prior to returning it to service.
The applicant's RTB periodic maintenance should also include, on a refueling interval basis:
1.
Pre-cleaning insulation res'istance measurement and recording; 2.
RTB dusting and cleaning; 3.
Post-cleaning insulation resistance measurement and recording, as specified by the Westinghouse Maintenance Manual; 4.
Inspection of main and secondary disconnecting contacts, bolt tightness, secondary wiring, mechanical parts, cell switches, instruments, relays and other panel mounted devices; 5.
UVTA trip force and breaker load check as specifiea by the Westinghouse Maintenance Manud; 6.
Measurement and recording of RTB response time for the undervoltage trip; 7.
Functional test of the bre~aker prior to returning to service as specified by the Westinghouse Maintenance Manual.
2.2 Trending of Parameters Generic Letter Item 4.2.2 specifies that the applicant's preventative maintenance and surveillance program is to include trending of parameters affecting operation and measured during testing to forecast degradation of l
Beaver Valley 2 SSER 1 3
Appendix J
operation. The parameters measured during the maintenance program described above which are applicable for trending are undervoltage trip attachment dropout voltage, trip force, response time for undervoltage trip and breaker insulation resistance. The staff position is that the above parameters are acceptable and recommended trending parameters to forecast breaker operation degradation or failure.
If subsequent experience indicates that any of these parameters is not useful as a tool to anticipate f ailures or degradation, the licensee or applicant may, with justification and NRC approval, elect to remove that parameter frora those to be tracked.
3.
EVALUATION 3.1 Evaluation of the Applicant's Position on Item 4.1 The applicant committed to incorporation of vendor-recommended modifications on the DS-416 RTBs. The staff finds the applicant position on Item 4.1 to be acceptable.
3.2 Evaluation of the Applicant's Position on Item 4.2.1 The applicant states that his preventative maintenance program will include the Group A and Group B activities described in the Westinghouse Maintenance Manual for the DS-416 RTB, and that the maintenance will be performed according to the schedule discussed therein. The staff finds that the applicant's maintenance interval is acceptable. This ecceptance is based on the Westinghouse recommendation that maintenance on RTBs located in mild environments should be performed annually. The vendor recommendation that RTBs located in harsh environments or experiencing severe load conditions be maintained more frequently is not applicable to these RTBs because of their location in a mild environment and reduced service duty at Beaver Valley 2 (less than 200 RTB cycles per refueling interval). The staff finds the licensee position on Item 4.2.1 to be acceptable.
Beaver Valley 2 SSER 1 4
Appendix J
3.3 Evaluation of the Applicant's Position on Item 4.2.2 The applicant has commited to an RTB parametric trend monitoring program which will include all of the parameters listed in Section 2.2 of this SER. The staff finds the licensee position on Item 4.2.2 to be acceptable.
4.
CONCLUSIONS The staff finds the applicant's position on Items 4.1, 4.2.1 and 4.2.2 of the Generic Letter to be acceptable.
Beaver Valley 2 SSER 1 5
Appendix J
r APPENDIX K STAFF SAFETY EVALUATION REPORT ON POST-TRIP REVIEW, ITEM 1.1 0F GENERIC LETTER 83-28 Beaver Valley 2 SSER 1 Appendix K
SAFETY EVALUATION REPORT FOR GENERIC LETTER 83-28, ITEM 1.1 - POST-TRIP REVIEW (PROGRAM DESCRIPTION AND PROCEDURE)
BEAVER VALLEY POWER STATION, UNIT 2 DOCKET NO.:
50-412 I.
INTRODUCTION On February 25, 1983, both of the scram circuit breakers at Unit 1 of the Salem Nuclear Power Plant failed to open upon an automatic reactor trip signal from the reactor protection system. This incident occurred during the plant start-up and the reactor was tripped manually by the operator about 30 seconds after the initiation of the automatic trip signal. The failure of the circuit breakers has been determined to be related to the sticking of the under voltage trip attachment.
Prior to this incident, on February 22, 1983, at Unit 1 of the Salem Nuclear Power Plant, an automatic trip signcl was generated based on steam generator low-low level during plant start-up.
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 generic implications of these occurrences at Unit 1 of the Salem Nuclear Power Plant. The results of the staff's inquiry into the generic implications of the Salem unit incidents are reported in NUREG-1000, " Generic Implications of 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 an operating license, and holders of construction permits to respond to certain generic concerns. These concerns are categorized into four areas:
(1)
Post-TripReview,(2) Equipment Classification and Vendor Interface, (3) Post-Maintenance Testing, and (4) Reactor Trip System Reliability Improvements.
The first action item, Post-Trip Review, consists of Action Item 1.1, "Progran Description and Procedure" and Action Item 1.2 " Data and Information Capability." This safety evaluation report (SER) addresses Action Item 1.1 only.
Beaver Valley 2 SSER 1 1
Appendix K
n
- II.
REVIEW GUIDELINES The following review guidelines were developed after initial evaluation of the various utility responses to Item 1.1 of Generic Letter 83-28 and incorporate the best features of these submittals. As such, these review guidelines in effect represent a " good practices" approach to post-trip review. We have reviewed the applicant's response to Item 1.1 against these guidelines:
A.
The licensee or applicant should have systematic safety assessment procedures established that will ensure that the following restart criteria are met before restart is authorized.
The post-trip review team has determined the root cause and sequence of events resulting in the plant trip.
Near term corrective actions have been taken to remedy the cause of the trip.
The post-trip review team has performed an analysis and determined that the major safety systems responded to the event within specified limits of the primary system parameters.
The post-trip review has not resulted in the discovery of a potentici safety concern (e.g., the root cause of the event occurs with a frequency significantly larger than expected).
If any of the above restart criteria are not met, then an independent assessment of the event is performed by the Plant Operations Review Committee (PORC), or another designated group with similar authority and experience.
Beaver Valley 2 SSER 1 2
Appendix K B.
The responsibilities and authorities of the personnel who will perform the review and analysis should be well defined.
The post-trip review team leader should be a member of plant management at the shift supervisor level or above and should hold or should have held an SR0 license on the plant. The team leader should be charged with overall responsibility for directing the post-trip review, including data gathering and data assessment a~nd he/she should have the necessary authority to obtain all personnel and data needed for the post-trip review.
A second person on the review team should be an STA or should hold a relevant engineering degree with special transient analysis training.
The team leader and.the STA (Engineer) should be responsible to concur on. a decision / recommendation to restart the plant. A nonconcurrence from either of these persons should be sufficient to prevent restart until the trip has been reviewed by the PORC or equivalent organization.
C.
The licensee or applicant should indicate that the plant response to the trip ever.t will be evaluated and a determination made as to whether the plant response was within acceptable limits. The evaluatio:,should include:
A verification of the proper operation of plant systems and equipment by comparison of the pertinent data obtained during the post-trip review to the applicatle data provided in the FSAR.
An analysis of the sequence of events to verify the proper functioning of safety related and other important equipment. Where possible, comparisons with previous similar events should be made.
Beaver Valley 2 SSER 1 3
Appendix K
- D.
The licensee or applicant should have procedures to ensure that all physical evidence necessary for an independent assessment is preserved.
E.
Each licensee or applicant should provide in its submittal, copies of the plant procedures which contain the information required in Items A through D.
As a minimum, these should include the following:
The criteria for determining the acceptability of restart The qualifications, responsibilities and authorities of key personnel involved in the post-trip review process The methods and criteria for determining whether the plant variables and system responses were within the limits as described in the FSAR The criteria for determining the need for an independent review.
III.
EVALUATION AND CONCLUSION By let_ter dated March 3, 1984, the applicant of Beaver Valley Power Station, Unit 2, provided information regarding its Post-Trip Review Program and Procedures. We have evaluated the applicant's program and procedures against the review guidelines developed as described in Section II. A brief description of the applicant's response and the staff's evaluation of the response against each of the review guidelines is provided below:
A.
With regard to the criteria for determining the acceptability of restart, the applicant indicated that the Shift Supervisor will be responsible for filing a Draft lncident Report for any unscheduled reactor trip prior to the request for authorization to restart. The Draft Incident Report will contain:
a presentation of an effective reconstruction and analysis of the event; a verification that the Beaver Valley 2 SSER 1 4
Appendix K reactor protection system and the engineered cafety features and systems which are important to reactor safety have performed as required; the cause of the trip and the subsequent corrective action taken; and, if a discrepancy in safety-related equipment or system operation was identified, verification that corrective actions required are complete.
We find that the applicant's criteria for determining the acceptability of restart are acceptable.
B.
The qualifications,. responsibilities and authorities of the personnel who wM1 perform the review and analysis have been clearly described.
We have reviewed the applicant's chain of command for responsibility for post-trip review and evaluation, and find it acceptable.
C.
The applicant has described the methods and criteria for comparing the event information with known or expected plant behavior.
Based on our review, we find them to be acceptable.
D.
With regard to the criteria for determining the need fer independent assessment of an event, the applicant has indicated that if the cause of the trip cannot be determined, or an effective reconstruction and analysis of the event cannot be performed, or performance of specified systems is in question, an independent assessment of the event will be performed.
In addition, the applicant has established procedures to ensure that all physical evidence necessary for an independent assessment is preserved. We find that these actions to be taken by the applicant conform to the guidelines as described in the above Sections II.A and D.
l Beaver Valley 2 SSER 1 5
Appendix K
-- E.
The applicant has provided for our review a systematic safety assessment program to assess unscheduled reactor trips.
Based on our review, we find that this program is acceptable.
Based on our review, we conclude that the applicant's Post-Trip Review Program and Procedures for Beaver Valley Power Station, Unit 2, are acceptable.
i Beaver Valley 2 SSER 1 6
Appendix K
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