ML19317G632
| ML19317G632 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 06/10/1975 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19317G631 | List: |
| References | |
| NUDOCS 8003240904 | |
| Download: ML19317G632 (18) | |
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SUPPLDfENT NO. 2 10 THE SAFETY EVALUATION BY THE OFFICE OF NUCLEAR-REACTOR REGULATION U.S. NUCLEAR REGULATORY Com!ISSION IN THE MATTER OF SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO NUCLEAR GENERATING STATION UNIT NO. 1 DOCKET NO. 50-312 Date: June.10, 1975 Lacos 24a @I
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1 TABLE OF COYIENTS E*K*.
1.0 INTRODUCTION
1 2.0 OCONEE UNIT l'0PERATING EXPERIENCE 2
3.0
-RANCHO SECO OPERATING EXPERIENCE 3
3.1 Testin'g and Operation 3
4 3.2'
' Boron Feed and Bleed System 4
3.3 Operational Occurrences 5
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'3.3.1
-Ejected Rod Worth 5
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3.3.2 Reactor Vessel Noisc 6
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4.0 REPORT OF THE ADVISORY COSMITIEE ON REACTOR SAFEGUARDS 7
5.0 TECHNICAL SPECIFICATION ' CHANGES 10
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. CONCLUSIONS 11 APPENDIX A'- BIBLIOGRAPHY.
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. APPENDIX B - INSPECTION SUbMARY B-1 J
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1.0 INTRODUCTION
The operating license for ' Rancho Seco Unit I was issued on August 16 1974, for operation at 2772 MWt (100% full power). However, the Technical Specifications attached to the operating license temporarily limit core power to levels not exceeding 2568 brdt pending confirmation of anticipated operating performance of the boron feed and bleed sys-tem and review by the Advisory Committee on Reactor Safeguards (ACRS).
By letter dated March 12, 1975(1), the licensee, Sacramento Municipal Utility District (SMUD), proposed changes to the. Technical Specifica-tions to permit operation at the design power level of 2772 MNt.
In support of the proposal, SMUD submitted a Startup Report (2), a plant operatinggrformancereportwhichevaluatestheboronfeedandbleed operation
, and reports of operating history (4).
The power esca-lation program has been completed, and Rancho Seco as of May 31, 1975, had accumulated 1413 hours0.0164 days <br />0.393 hours <br />0.00234 weeks <br />5.376465e-4 months <br /> operation at 2568 MWt. The choice of' 2568 MWt as an interim limit for Rancho Seco operation took into considoration the 2568 MWt design power level of the previously licensed Babcock 6 Wilcox (B6W) prototype plant, Oconee Unit 1.
The reactor core for Rancho Seco is substantially the same as that of Oconee Unit i (Docket No. 50-269).
Both cores contain 177 fuel as'semblies with 208 fuel rods per assembly, however, the design heat output of the Rancho Seco core is 2772 Mit which is 8% higher than the design output of the Oconee core (2568 Mit). This 8.0% power increase is accomplished by using a 5.0% larger reactor coolant mass flow rate for Rancho Seco while permitting a 60F larger reactor coolant temperature rise across the Core.
-The ACRS issued a report on its review of the Rancho Seco operating license application.on September 11, 1973(5), and on Novembe 1973, the staff issued Safety Evaluation Supplement No.1(6)r 28, which addressed areas of concern identified in the ACRS report.
This safety evaluation (Supplement No. 2) updates the information contained
'in the November-28,197315afety Evaluation and presents the staff safety evaluation of the proposod Technient Specification changes
' permitting 100% full power operation.
-Appendix A contains a Bibliography. Appendix B contains a report from the Office of Inspection and Enforceme nt (OI6E) summarizing
' inspection results pertinent to the proposed power increase.
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2.0 OCONEE UNIT 1 OPERATING EXPERIENCE.
We have reviewed the operation of Oconee Unit I during its first fuel cycle (7,8,9,10,11612).
Oconee Unit I achieved initial criticality April 19, 1973; reached full power on November 9,1973; and after 310 equivalent full power days operation on the Cycle 1 core loading shut down for refueling on October 18,.1974.
Refueling was completed during December 1974.
.0conee was a first-of-a-kind system for B6W and as such was subjected to an extensive startup test program (12).
In addition to the usual c
startup tests, load-following transient tasks were performed at 75%
and 95*. full power (FP) and azimuthal and diagonal (combination of azimuthal and axial) xenon' transients were initiated and followed.
Oconee performed well during these tests, meeting all acceptance criteria. - The deficiencies encountered were of a nature to be expected during the.startup of a complex unit and it was concluded that Oconee could be safely operated at full power.
Duke Power Company (the licensee for Oconee) has submitte'd reports (10Sll) on comparisons between measured and calculated assembly power at several times during the initial fuel cycle.
Particular attention was given to control rod interchanges. These comparisons show that, for assemblies having powers within 5% of the peak power, the largest difference between calculation and measurement was 4.6% (at BOL) and the average difference for all the measurements was less than 4%.
Rod interchanges had only a small effect (<1%) on the comparison.
Review of Oconee operating history for the period July 1973 through December _1974 revealed no abnormal events that would preclude ~operatirg,
Rancho Seco at 2772 bMt. Several modifications to operating proce-dures and equipment related to reactor systems were instituted during this period. These modifications, which relate to the availability of ECCS equipment, have been reflected in the operating procedures for Rancho Seco.
On the basis of our review, we find the operation of Oconee Unit I during its initial fuel cycle to be satisfactory.
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- 3.0 ' RANCHO SECO OPERATING EXPERIENCE i
3.1 Testing and Operation-The opetsting license for the Rancho Seco Nuclear Station was granted to SMUD on August 16, 1974.
We reviewed the Rancho Seco Startup Report (2) which summarized significant. activities from the date of obtaining the operating license to the end of the power escalation-c testing at 92.6% FP.
The first fuel assembly was inserted into the core on August 19, 1974, and_ initial fuel loading-was completed on August 23, 1974. On
' September 16, 1974, Rancho Seco successfully achieved criticality.
Zero Power Physics testing'which commenced on September 16, 1974, was successfully. completed October 3,1974.
This program was conducted primarily at reactor coolant temperatures of 3000F and 532oF.
- Power escalation was begun on October 13, 1974, and further power level escalations occur.ed as required testing was satisfactorily completed.
Major' power plateaus as defined by the power escalation test program, were initially achieved as follows:
Power Level- ('.FP)
Date Completed 1
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15 October 13, 1974 40 December 2, 1974
-7S January 9, 1975 92.6-January 24,.1975
- As of May 31, 1975, Rancho _Seco had been operated for more than 2113 1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at power levels-greater than 75% FP and 1413 hours0.0164 days <br />0.393 hours <br />0.00234 weeks <br />5.376465e-4 months <br /> at 92% FP.
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Testing and demonstration of _the Engineered Safety Features for Rancho
. Seco have revealed normal operation:throughout the startup 4-phase of operation, ascension to power, and operation at 2568 MWt.
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-The Engineered Safety Features ' for Rancho Seco Unit '1' consist of the reactor building and its associated ventilatien and isolation systems, the~ emergency core cooling system, the containment cooling system, the containment' spray system, and:the emergen:y feedwater system.
1 The experience grined by operating personnel and the satisfactory
- equipment performance ~ during ~the period of almost a year since the operating license was issued contribute to the safety of the-proposed operation:at 2772 MWt._
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Acceptance criteria were established in advance of testing which allowed for anticipated calculational and measurement uncertainties.
The startup test program demonstrated that rod group worths, stuck rod worths, ejected rod worths (after change'in rod insertion limits),
and reactivity coefficients were well within acceptance criteria and met technical ~ specification limits.
Power distribution measurements were performed at: 15%, 40%, 75%, and 92.6% FP.
Total peaking factors were well within the acceptance criterion. The fuel assembly to average' fuel assembly power ratio was outside the criterion for one assembly at the three highest powerA However, further evaluation revealed this condition to be acce table, since the. total peaking factor in this assembly was well within the acceptance criterion and extrapolated DNBR and kW/ft values showed adequate safety margins.
It was concluded that continued operation was safe.
Minimum DNBR values and maximum linear heat rate values at each power-level were adjusted to account for uncertainties and conservatisms (yiciding " worst case" values) and these were extrapolated to the next highest power plateau in order to ensure safe margins at the next level.
" Worst case" values at the 92.6% FP level, extraporated
- to 105.5% FP, showed a 60.8% margin for DNBR and 12.6% margin for linear heat rate.
3.2 Boron Feed and BIced System BGW has provided Rancho Seco with a boron feed and bleed system of
- greater contro1 ' capability than for any B6W plant of earlier design.
Pursuant to the ACRS recommendation to review Rancho Seco operation, we requested:a special performance report for the feed and bleed system.
We havo reviewed the performance report (3) submitted by the licensee describing the boron' feed and bleed system test program and operating experience Data were gathered during the operation of the feed and bleed system in order to:
1.
permit verification of the ability of the nucicar steam supply system to perform power transients with feed and bleed operations, including the design transient;
- 2. _ determine the accuracy 'of the feed and bleed maneuvers; and 3.
permit evaluation of the proficiency of operating per-sonnel to perform feed and bleed operations.
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All of the operating shifts successfully demonstrated control of the plant during power ramps. The Integrated Control System was utilized in its various modes (Integrated,. Turbine.Following, and/or Reactor
'Following).
Both " Feed and Bleed" and " Batch" deboration techniques
_ were employed. A " Pseudo Design Transient" (80% FP to 30% FP and return to 80% FP after 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />) was successfully performed *.
This transient produced a.more severe xenon effect than predicted for the actual design transient (100% FP - 50% FP - 100% FP) since the power change was actually 80-30-80; i.e., greater than 50% of the initial power level. The results showed that extrapolated (to 100% 'FP) DNBR and linear heat generation rate margins were adequate so it will not be c
necessary to run the actual design transient.
The accuracy of the bleed and feed operations (as measured by the comparison between the targeted final boron concentration and the actual concentration) was within the accuracy of the boron cancen-tration measurement.
Based on our review, we have concluded that satisfactory operation of the feed and bleed system has been sufficiently demonstrated to parmit plant opeation at 2772 MWt.
3.3 Operational Occurrences We have reviewed abnormal occurrences, unusual events, and test results for the Rancho Seco startup and operation through May 1975(13). No events have occurred that would preclude operation at 2772 MWt. However, cer-tain findings of interest that were encountered during the startup test program are discussed below.
3.3.1 Ejected Rod Worth The ejected rod worth at scro power was measured.to be 1.24%
Ak/k compared to the predicted value of 0.9% Ak/k and the Technical Specification limit of 1.0% ak/k (for operation except for low power physics testing).
Following the measurement, additional calculations were performed (by B6W) and a new correlation was developed between ejected rod worth and inserted rod' worth using all available calculations and measurements. The new correlation was used to obtain a value for the maximum permissible rod insertion at zero
. power' and insertion limits were altered accordingly after
-our review and approval.
- The design transient could not be performed since 100% FP operation was prohibited by Technical Specifications..
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13.3.2! JReactor Vessel Noise
-An unexpected sound was recorded on the loose parts monitor during non-~ nuclear single pump operation. After analysis, lit :was. concluded that the sounds came from motion of the core support structure against.the core support lugs and that motion of the core support structure within its design envelope could produce the sounds. The sounds occurred only during operation of the "A" and "C" pumps singly in the operating-temperature range.
It should be notad that single pump operation is not a mode of operation permitted by Technical Specifications when the reactor is critical. The licensee and his consultant have initiated a program using excore detectors in addition to the loose parts monitoring system to gain additional information over an extended period of time.
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.4.0 REPORT OF THE ADVISORY CO5NITTEE ON REACTOR SAFEGUARDS 1Rus ACRS. has issued a report on its review of the Rancho Seco operating license application (5) and the staff has considered the comments and recommendations contained in that report.
The steps which the staff has taken or will take relative to these comments and recommendations are described in the following paragraphs.
The Committee recommended that three conditions be satisfied before this plant be. allowed to operate at full power.
a.
The operation of Unit 1 of 'the Oconee Nuclear Station should be reviewed and found satisfactory by the Regulatory staff.
b.
Following an appropriate. period of operation at power levels up to 2568 FMt, the operating. experience of Rancho Seco Unit I should be reviewed by the Regulatory staff and the ACRS.
Prior'to the review in b above, the Regulatory staff c.
should perform and report on an. independent confirmation of the licensee's linear heat gene' ration rates, operating limits and ECCS efficacy.
The staff has. completed its review of the Oconee power operation and the Rancho Seco operation at 2568 FMt. Oconee Unit 1 operation is discussed :in Section 2.0 of this report and Rancho Seco operation is discussed in Section 3.0 of this report.
With respect to e above and ECCS efficacy, the_ licensee has presented a new LOCA analysis in
. response-to 10 CFR 50.46.
The licensee's evaluation of the ECCS cooling performance submitted in August 1974 was-based on the model developed by BSW.
The staff
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concluded that the evaluation model required certain modifications
.to ' conform to 10-CFR Part 50, Appendix K.
As a result, the Commission issued an order on December :27,1974, limiting the linear heatfgenera-tion rate based. on the requi.rements 'of 10 CFR 50.46 (FAC). This order further states that the plant should be operated in accordance with
' limits based on -the IAC and FAC until a LOCA analysis with' an approved model is submitted.. The order requires the licensee to' submit a-reevaluation of ECCS calculated in-accordance with an acceptable evaluation model which conforms to the requirements of.10 CFR Part 50,:550.46 by July. 9,1975.
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With respect' to independent confirmation of linear heat generation rates and operating limits, the NRC staff through its consultant, BNL, has performed an independent calculation of BOL power distri-bution for the Rancho Seco reactor. Calculations were done with the PDQ-7, Version 2 code on the CDC-7600 computer. This version of PDQ-7-incorporates thermal-hydraulic feedback effects and has an automatic criticality.' search on boron concentrations. The standard BNL cross-section data set was processed by the HAhNER code (for cellhomogenization) and TWOTRAN code (for assembly homogenization) to produce four group cross-section sets for the PDQ-7 calculations.
i The resultsfl4) were compared to those obtained by B6W as reported in BAW-1393(15)- and to additional data obtained from BSN. Steady state power distributions were obtained for BOL full power conditions and the design power transient (100's FP - - 50*6 FP - 100's FP),at BOL was calculated. The BNL values of maximum total peaking factors agreed closely (<3.5's).with those calculated by B6W and we: e generally lower.
Radial peaking factors (F2g) agreed within 4.51, of values calculated by B6W.
On the basis of these audit calculations, we find the heat generation rates' calculated for Rancho Seco to be acceptable.
We have reviewed the methods used by B6W to determine operating limits. Topical Reports (15616) and informal discussions with B6W' were used in the review. The major calculational tool is the PDQ - code with thermal-hydraulic feedback. A large number of calculations are performed with.various regulating rod configurations and a complete range of axial power shaping rod positions. Xenon transients produced by the design power transient were calculated at BOL, near EOL, and at one or'more additional times during the cycle. The calculated power
. peaks are used, after correction for uncertainties and conservatisms,
.to_ establish operating limits. Operating limits for a particular
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parameter (e.g., axial imbalance) are estalilished under the assumption th'at all other parameters have their most adverse permissible values.
On.the basis of.our_ review,-we find the methods used to obtain operating
_ limits.to'be' acceptable.
Other concerns expressed by the ACRS(5)-have been addressed in our safety evaluation of November 8, 1973(6). The following information
. updates:the November 8, 1973 report.
Fuel Loading Procedures
' Detailed fuel loading procedures were developed by the licensee L.
which provided for1 obtaining a -permanent record of the installed -
locat. ion of. every fuel assembly.. These. procedures and records 8:
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' were reviewed by the staff. The performance of independent fuel-assembly identification with respect to core loading location
- was" required-by.the procedures and was confirmed by OISE personnel monitoring'the initialffuel loading.
. Common Mode Failure and Anticipated Transients Without Scram-
. The ~ staff technical report, " Anticipated Transients Without Scram
. for Water Cooled Power Reactors", and a request that' the licensee identify the course of action to resolve ATWS was sent to the licensee on October 19, 1973(17).
The licensee responded to the request by Ictters dated September 30, 1974, and December 30, 1974(18), r.eferencing B6W topical reports BAW-10016 and BAN-10099.
The staff. review of these reports ar.d application of them to individual plants is in progress.
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- 5.0 TECHNICAL SPECIFICATION CHANGES
,We have reviewed the cha'nges in Technical Specifications proposed by the ' licensee (1619) to permit operation of Rancho Seco at the. rated power of 2772 MWt. 'These include the rod withdrawal limits proposed-
-by.the licensee which represent extensions of the present limits to
. full power operation, and the core imbalance curves which have been modified to permit:100% FP operation and take into account the fact that core: exposure is now greater than 100 equivalent full-power days.
he have also added a Technical Specification change proposed by the
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licensee in his -letter of April 7, 1974(19), requiring regulating rod positioning prior to deboration. This change adds a restriction to the Technical Specifications which had previously been imposed administrative 1y by the licensee _ as a result of -the ejected rod worth measurements described in Section 3.3.1 of this report.
'Mie Technical Specification changes proposed (1619), as discussed above, address the ' changes needed for. full power operation of Rancho Seco at the current core' exposure, and also impose additional restrictions on rod positioning.
Br. sed on our review and recommended approval of Rancho Seco operation at full power, we have concluded that all of
. these Technical,Specificatio.. changes are acceptable.
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CONCLUSIONS.
We have reviewed the performance.of O' conee Unit 1. B6W prototype plant
.o'perating at 2568 bMt and have found its operation to be satisfactory.
Wefhave, reviewed the.oper.1 tion of Rancho Seco Unit 1, including the
- startup _ tests, the' boron _ feed and biced system performance, 'and
-initial operating ~ ~perfornance up_ to :and-including 1413 hours0.0164 days <br />0.393 hours <br />0.00234 weeks <br />5.376465e-4 months <br /> operation at 2568 bMt.
We have found-the operation ~of. Rancho Seco to be ' satis-
-factory.-~In ourJreview we have found no reason to preclude operation at;the propose'd power ~1evel-of 2772 MWt.
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. APPENDIX A' BIBLIOGRAPHY
-1. - LLetter SMUDfto' NRC. re Proposed Technical Specification Chance No. I dated March 12, 1975.
2,.
Sacramento Municipal Utility District, Rancho Seco Nuclear Generating Station, Unit 1-Startup Report, March 1975.
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' 3.= Sacramento Municipal Utility District, Rancho.Seco Nuclear Generating
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-Station,' Unit 1 Performance Report, March 1975.
- 4. _ Letter SMUD to NRC re January-March 1975 Operating Reports dated May 8, 1975.-
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Letter H. G. Pfangel'sdorf to D. L. Ray re " Report on Rancho Seco Nuclear Generating Station, Unit 1" dated September 11, 1973.
6.
' Supplement. No,' I to-the Safety Evaluation by the Directorate of Licensing, USAEC,ERancho'Seco Nuclear Generating Station Unit No.1, November 28,-
1973.
'7.
Duke Powe'r'Campany,-Oconee Nuclear Station, Semiannual Report, Period
.Ending December 30,:1973.
8.
DukcL Power Company, Oconee Nuclear Station, Semiannual Report, Period
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- Ending; June'. 30,.1974..
19./ Duke ~ Power. Company, Oconee Nuclear Station, Semiannual Report, Period Lending; December 31, 1974.
Letter:ThiesitoGidmbussorePowerDistributionComparisonStatusReperi 10.
dated December. 21, 1973.
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211.;1 Letter ThiesEto Giambusso re Power' Distribution Comparison Status Report
. dated: July;19,:1974.
J 12. : DukeiPower. Company, - Oct Jee : Nuclear. Station Unit 1,'Startup Report,
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LNovember-16, 1973.,
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Sacramento Municipal Utility District, Rancho Seco Nuclear Generating Station, Unit:1 Annual' Report, March 1975.
~14.
D.-Diamond, Audit of B6W Power Peaking Calculations for Rancho Seco Unit 1,; BNL' memo -(to be. published).
15.
Rancho'Seco Unit-1 Fuel Densification Report, BAW-l'393, June 1973.
-16. ~ 0perational Parameters for B5W Rodded Plants, BAW-10078, September 1973.
-18.
Letters SMUD to NRC' re ATWS ' Analysis for Rancho Seco dated September 30, 1974, and December 30, 1974.
19.
Letter SMUD to NRC're Addendum 1 to Proposed Amendment No. 29 dated
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April 7, 1975.
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APPENDIX B' OFFICE 07 IMSPECTIC:: A :D E::FORCEME!:T
. INSPECTIO:.Sp0!ARY SACRAME!iTO 1ATICIPA1. UTILITf DISTRICT PANCHO SECO DOCKET No. 50-312 The operation.of the Rancho Seco Nuclear Generating Station by the Sacratento Municipal Utility District (SMUD) has been examined by the Office'of Inspection and Enforcement since the operating license vas issued on August 16, 1974 Since issuance of the operating licence ten inspections have been perforced, seven of chich were unannounced.
The total amount of tine spent at the plant site during these inspections n) was approximately seventy-five can, days, s
The results of the !!RC inspection program to date show that the Rancho Seco'Uuclear Concrating Station has'been operated sofcly since initial startup.in Septctber 3974.
The performance characteristics of the reactor and enginected safeguards systems have been determined during the startup test program threugh 92.6% of full paver (2560 lC;t).
The startup test 7
recults have been found to ccer the test acceptance criteria deter =ined from the plant do.nign bases dercribed in the Final Safety Analysis Report.
The significant results of the inspection program that are pertinent to the proposed power increase fro: 25 6 S M'o't to 2772 h"..'t are discussed below.
- 1..Startup Test Procran The startup test pregran ce==cnced with initial fuel loading on August 19,~1974 and'was.ccepicted through 92.6% of full power (256S' K.s't) on Parch 22, 1975.
-15%, 40%, 75% and 92% cf full power. Testing was perforacd at zero power, The test data obtained during
'the performance of the following startup tests have been reviewed and evaluated by our inspectors.
a.
2cro-Pover' Physics-Testl
- b.. NucleariInstrumentation Calibration at Power c.. Corc' Power-Distribution d.' Reactor / Turbine ~ Trip.
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- e. : Dropped Control Rod Asse=bly Lf. ' Power Icbalance Detector Calibration g.
Uucicar Steam Supply System Heat Salance.
h.~ Reactivity, Coefficient at Pcver
- i..Psuedo Control 4 Rod ' Asse=bly Ejection I
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The_ evaluation' the Psuedo Control Ecd. Assensly Ejcetion test,
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conducted during zero power testing, re'vealed that the ceasured value af
_of.the rod vorth differed significantly from the predicted value.
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The licensee inforced our inspector that the predicted value was in error andLa reevaluation of the physics calculation indicated that the measured worth should have been anticipated.
The resolutica of the
.psuedo'cjected rod worth value and corrective action taken by the licensoc are described in the licensce's letter to Licensing (RJR 74-401 dated Octcber-23,-1974) and in Proposed Arend=ent tio. 29 to the FSAR formally subsitted on Decc=ber 6,'1974.
The data for all other tests l
evaluated by the inspector vero found to be consistent with the predicted values of the para =cters measured.
In n'ddition to the above tests, the in aial fuel loading, initial criticality, unit loss of cicctrical load test, and the loss of offsite
<pover test were directly. observed by our inspectors.
During the vitnessing of there tests, ec=pliance with approved procedures was verJfied and the-acceptability of test results was independently veriffed'by our inspector;vith no anomalics observed.
A special test procedure, " Bleed and Tecd De=onstration", was used by
.ihe licensce to satisfy the recuirc ents of technical specification 3.12Lvhich required that significant loed changes be performed by operating personnel _ to de::onstrate satisfa: tory systen operability.
I Du'r inspector rrvfewed the raw test data cbtained during the performance of_the special t'est and (cund it consistent with the inforr.atien 7
contained-in the licensec's performance report dated March 1975.
2.
Pinnt Operations Through observations of plant operation, examination of facility records, and discussion with licensee representatives, the operstien of the facility has been found consistent with the requirc..ents of the techn'ical specifications. -Tours of the facility have been t.ade on cach inspection.
During there cours, observation of techanical equiptent and piping systc=s:have'shown no execssive 1caks or vibrations.
' Instrumentation systc= incl ding nucicar instrua.catation, reactor u
protection-and safety l features actuation systens have been found to be
- operating: norcally with the recuired tests and calibrations teing.
perforced.as scheduled.
The unidentified reactor coolant leakage has
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averaged-loss than:0.4 gallen per minute throughout plant epcratica to date as = cocpared. to the technical specification li=1t.of 1.0 gallon
- per, uinu te. -
There-have been-no cajor outages causec by the failure of safety related.cquip=cnt or ec=ponents.
Two outages of significant duration have~ occurred'since' initial operation, a 26-day outage in October /
iNovember~1974 for the_ repair 1of condenser tube leakage and codification
- of :the ' turbinc _ stop. valves and a 22.tay ' outage in March / April 1975
.for the'-inspection ofiturbine bearings.
Maintenance of safety 6
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s e4 related equip =ent'has been recorded by the licensee in ::onthly
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Operating Reports.
Exacination of maintenance records verified that the caintenance had been perfor=cd consistent with the licensee's managecent control system.
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The plant availability. factor since coe=ercial operation at 92.6%
power has been 100%. The capacity factor.for the facility has been approxicately 95% since co =creial operation.
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Unusual-Occurrences The licensee submitted fourteen (14) abnormal occurrence reports in 1974 and to date has sub=itted nine (9) abnormal occurrence reports in 1975. _The circumstances and corrective action described byf the licensee in cach abnormal occurrence report have been verified during the inspection program.
4.
Radiological Protection Our inspectors have verified that the licensco's radiolog cal protection progran has been inplc=cated consistent uith regulatory requirements.
-Results of radiological surveys perfor. icd during plant startup tests indicate that the radiation. Nones, based en current data projected to 100% of f ull power, vill be as described in the FSAR, Section 11.2.1.1.
The results of *.he radiological environmental tonitoring pengrr.: fer the last two quarters of 1974 did not identify any significant
. adverse environmental effcets resulting fren the operation of the facility.
The records of routinc surveys perforacd by the licensee
~have chown-that r..diation and contcaination levels in uncontrolled
.arcan have been insignificant.
5.
_ Quality Assurance Procran for operations The deplecentation of a Quality Assurance Progratt for Operations was
~
verified prior to the receipt of the Operating License.
The c,uality assurance program has been subject to a continued exanination by
. our inspectors during the-initial operatica phase.
Itc=s of noncompliance. vere identified related to inspection planning, receipt inspection perfor=ance.. independence of inspection personnel and implenentatien' of corrective action for deficiencies identified by the licensec',s internal audit program.
The licensee has promptly
. respondedivith corrective actirn con =itments which ucre subsequently
= verified by cur inspectors to sitisfactorily resolve the enforec=ent items.. All other.cperational activitics'have been found to be j
consistent with the requirements 3f the quality assurance program for'opc*ations.-
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