ML19317G804
| ML19317G804 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 03/05/1976 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19317G794 | List: |
| References | |
| NUDOCS 8004010633 | |
| Download: ML19317G804 (16) | |
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REVISED ~
SUPPLDIENT NO. 2 TO Tl!E SAFETY EVALUATION I
BY THE I
OFFICE OF NUCLEAR REACTOR REGULATION U.S. NUCLEAR REGULATORY CO.41ISSION i
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IN THE MATTER OF i
SACRAME.\\TO MUNICIPAL UTILITY DISTRICT I
RANC110 SECO NUCLEAR GENERATING STATION UNIT NO. I DOCKET NC. 50-312 i
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s-Date: MAR.' 5 1978 seourog33
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i-i TABLE OF CONTENTS W
1.0 INTRODUCTION
1 2.0 OCONEE UNIT 1 OPERATING EXPERIENCE 2
3.0 RANC!lO SECO OPERATING EXPERIENCE 3
3.1 Testing and Operation 3
3.2 Boron Feed and Bleed System 4
a 3.3 Operational Occurrences S
3.3.1 Ejected Rod Worth 5
3.3.2 Reactor Vessel Noisc 6
4.0 REPORT OF THE ADVISORY CO:CIITTEE ON REACTOR SAFEGUARDS 7
i 5.0 TECHNICAL SPECIFICATION CHANGES 10
6.0 CONCLUSION
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APPENDIX A - BIBLIOGRAPHY A-1 APPENDIX B - J.NSPECTION SUSD!ARY B-1 t
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1.0 IfrTRODUCTION The operating license for Rancho Seco Unit I was issued on August 16, 1974, for operation at 2772 MWt (100*. full power). However, the j
Technical Specifications attached to the operating license temporarily i
limit core power to levels not exceeding 2568 FMt pending confirmation of anticipated operating performance of the boron feed and bleed sys-tem and review by the Advisory Committee on Reactor Safegrards (AC'!S).
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 Frdt.
In support of the proposal, SMUD submitted a Startup Report (2), a plant operating p operation *erformance report which evaluates the boron feed and biced I>
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, and reports of operating history (4).
The power esca-i 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 2S68 MWt as an interim limit for Rancho Seco operation took into consideration the 2568 Fidt design pouer icvel of the previously licensed Babcock 6 Wilcox (BGW) prototype plant, Oconee Unit 1.
The reactor core for Rancho Seco is substantially the same as that of Oconce Unit 1 (Docket No. 50-269).
Both cores contain 177 fuel as'semblies with 208 fuel rods per assenbly, however, the design heat output of the Rancho Seco core is 2772 Fr..~t which is 8% higher than the design output.of the Oconee core (256S Brdt). ~Miis S.00 power increase is accomplished by using a S.0% larger reactor coolant mass flow rate for Rancho Seco while permitting a 6oF 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 November 28, 1973, the staff issued Safety Evaluation Supplement No.1(6) which addressed areas of concern identified in the ACRS report. This safety evaluation (Supplement No. 2) updates the information contained in the November 28, 1973 Safety Evaluation and presents the staff safety ' valuation of the proposed Technical Specification changes e
permitting 100% full power operation.
Appendix A contains a Bibliography. Appendix B contains a report from the Office of Inspection and Enforcement (OISE) summarizing inspection results pertinent to the proposed power increase.
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I 2.0 OCONEE UNIT 1 OPERATING EXPERIENCE i
I We have reviewed the cperation of Oconee Unit I during its first fuel cyc1c(7,8,9,10,11612).
Oconce 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 l
down for refueling on October 18, 1974. Refueling was completed during December 1974.
Oconee was a first-of-a-kind system for BSN and as such was subjected to an extensive startup test program (12)
In addition to the usual 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, coeting all accep.tance 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 Oconce) has submitted reports (10F,11) on comparisons between measured and calculated assembly power at several l
times during the initial fuel cycle. Particular attention was given to
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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 Oconce operating history for the period July 1973 through December 1974 revealed no abnormal events that would preclude operating Rancho Seco at 2772 FR.'t.
Several modificat. ions 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 1 during its initial fuel cycle to be satisfactory.
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3.0 RANCHO SECO OPERATING EXPERIENCE 1
3.1 Testing and Operation The operating license for :nc Rancho Seco Nuclear Station was granted to SMUD on August 16, 1974.
h'c. 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-testing at 92.6% FP.
The first fuel assembly was inserte'd 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 conmenced on September 16, 1974, was successfully completed October 3, 1974. This progran was condu.eted primarily at reactor coolan't temperatures of 3000F and 532oF.
l Power escalation was begun on October 13, 1974, and further power level
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escalations occurred as required testing was satisfactorily completed.
Major power plateaus as defined by the power escalation test program, were initially achieved as follows:
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Power Level (%FP)
Date Completed 15 October 13, 1974 40 December 2, 1974 75 January 9, 1975 92.6 January 24, 1975 As of May 31, 1975, Rancho Seco had been operated for more than 2113 hours0.0245 days <br />0.587 hours <br />0.00349 weeks <br />8.039965e-4 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.
Testing and demonstration of the Engineered Safety Features for Rancho Seco have ' revealed normal operation throughout the startup phase of operation, ascension to power, and operation at 2568 Mh*t.
The Engineered Safety Features for Rancho Seco Unit I consist of the reactor building and its associated ventilation and isolation systems, the emergency core cooling system, the containment cooling system, the containment spray system, and the emergency feedwater system.
The experience gained 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 i
allowed for anticipated calculational and measurement uncertaintics.
l 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%, 40S, 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 powers. However, f rther evaluation u
revealed this condition to be acceptable, since the total peaking factor in this assembly was well within the acceptance criterion g
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 uncertaintics 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 1 cycl.
"Korst case" values at the 92.6% FP Icvel, extrapolated to 105.5% FP, shoued a 60.8% margin for DNBR and 12.6% margin for linear heat rate.
3.2 Boron Feed and Bleed System l
BSW has provided Rancho Seco with a boron feed and bleed system of
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greater control capability than for any BGW 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 have reviewed the performance report (3) submitted by the licensee describing the boron fend and bleed system test program and operating experience Data were gathered during the operation of the feed and bleed system in ordor to:
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permit verification of the ability of the nuclear steam supply system to perform power transients with feed and bleed operations, including the design transient;
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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 Fo11cwing, 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 i
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 1cvel. The results showed that extrapolated (to 100% FP) DNBR and linear heat generation rate margins were adequate so it will not be necessary to run the actual design transient.
l The accuracy of the bleed and feed operations (as measured by the i
comparison between the targeted final boron concentration and the actual concentration) was within the accuracy of the boron concen-tration measurement.
Based on our review, we have concluded that satisfactory operation of the feed and biced system has been sufficiently demonstrated to parmit i
plant opeation at 2772 MWt.
l 3.3 Operational Occurrences i
We have reviewed abnormal occurrences, unusual events, and test results l
for the Rancho Seco startup and operation through May 1975(13).
No events I
have occurred that would preclude operation at 2772 MNt. However, cer-j tain findings of interest that were encountered during the startup test i
program are discussed below.
3.3.1 Ejected Rod North The ejected rod worth at zero 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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-3.3.2 Reactor Vessel Noise I
An unexpected sound was recorded on the loose parts monitor
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I it 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 noted 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 COMMITTEE ON REACTOR SAFEGUARDS The 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.
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The operation of Unit 1 of the Oconee Nuclear Station j
should be reviewed and found satisfactory by the i
Regulatory staff.
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Following an appropriato period of operation at power l
1evels up to 2568 Mit, the operating experience of 4
Rancho Seco Unit 1 should be reviewed by the Regulatory.
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Prior to the review in b above, the Regulatory staff c.
should perform and report on an independent confirmation i
of the licensee's linear heat generation rates, operatino limits and ECCS efficacy.
l The staff has completed its review of the Oconee power operation and the Rancho Seco operation at 2568 Mit.
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 cfficacy, 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 BSN. The staff 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 heat genera-tion rate based on the requirements 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.
On.Tolv 8,1975, the licensee submitted this analysis which is presently being rev:cwed.
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9-With respect to independent confirmatica 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 IIAMNIER code (for ce11 homogenization) and TWOTRAN code (for assembly homogenization) s to produce four group cross-section sets for the PDQ-7 calculations.
The resultsfl4) were compared to those obtained by B6W as reported in BAN-1393(15) and to additional data obtained from B6W. Steady state power distributions were obtained for BOL full power conditions and the design power trcnsient (100% FP
was calculated. The BNL values of maximum total peaking factors agreed closely (<3.5%) with those calculated by B6W and were generally lower.
Radial peaking factors (F$H) agreed within 4.5's of values calculated by B6W.
On the basis of these audit calculations, we find the heat generation rates calculated for Rancho Seco to be acceptabic.
We have reviewed the methods used by B6N to determine operating limits. Topical Reports (15616) and informal discussions with B6W were used in the review. The major calculational tool is the PDQ-7 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 cperating limits. Operating limits for a particular parameter (e.g., axial imbalance) are established under the assumption that 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(S) 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 which provided for 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 ~1oading location 6
was required by the procedures and was confirmed by OI6E personnel-monitoring the initial fuel 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 letters dated September 30, 1974, and December 30, 1974(18), referencing. B5W topical reports BAW-10016 and BAW-10099.
The staff review of these reports and application of them to individual plants is in progress.
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S.O TECliNICAL SPECIFICATION CHANGES We have reviewed the changes in Technical Specifications proposed by s
the licensec(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's FP operation and take into account the fact that core exposure is now greater than 100 equivalent full power days.
We have also added a Technical Specification change prorssed by the 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 administratively by the licensee as a result of the ejected rod worth measurements described in Section 3.3.1 of this report.
In addition we have added a Technical Specification change requested by f
the licensee by letter dated May 30, 1975, which replaces Figure 2.3-2 to provide more restrictive flux / imbalance / flow limiting safety system settings than exist in the present Technical Specifications. This change is required to correct an error which was discovered in the calculations on which the present settings are based.
The 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 and flux / imbalance / flow limiting safety system settings.
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6.0 CONCLUSION
S We have reviewed the performance of Oconee Unit 1 B6W prototype plant -
operating at 2568 MWt and have four i its operation to be satisfactory.
We have reviewed the operation of Iincho Seco Unit 1, including the startup tests, the boron feed and F4eed system performance, and initial operating performance 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 MNt. We have found the operation of Rancho Seco to be satis-factory.
In our review we have found no reason to preclude operation at the proposed power level of 2772 MNt. The Advisory Committee on Reactor Safeguards reviewed the proposed operation of Rancho Seco at 2772 MNt on July 10, 1975, and by letter dated July 16, 1975, concluded that there is reasonable assurance that Rancho Seco can be operated at 2772 SMt without undue risk to the health and safety of the public.
We have determined that the amendment does not authorize a change in-l effluent types or total amounts nor an increase in power level not i
previously accepted and will not result in any significant environ-mental impact. Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact a'd, pursuant to 10 CFR n
851.5(d)(4), that an environmental statement, negative declaration, or environmental impact appraisal need not be. prepared in connection with the issuance of this amandment.
We have concluded, based on the considerations discussed above, that:
(1) because the change does not involve a significant increase in I
the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the change does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safey of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or t,o the health and safety of the public.
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i-APPENDIX A, 1.
BIBLIOGRAPHY 1.
Letter SMUD to NRC re Proposed Technical Specification Change No I dated March 12, 1975.
2.
Sacramento Municipal Utility District, Rancho Seco Nucicar Generating
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Station, Unit 1 Startup Report, March 1975.
3.
Sacranento Municipal Utility District, Rancho Seco Nuclear Generating Station, Unit 1 Performance Report, March 1975.
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Letter SMUD to NRC re January-March 1975 Operating Reports dated May 8, j
1975.
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Letter H. G._Mangelsdorf to D. L. Ray re " Report on Rancho Seco Nuclear i
Generating Station, Unit 1" dated September 11, 1973.
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Supplement No. I to the Safety Evaluation by the Directorate of Licensing, USAEC, Rancho Seco Nuclear Generating Station Unit No. 1, November 28, 1973.
7.
Duke Power Company, Oconee Nuclear Station, Semiannual Report, Period Ending December 30, 1973.
8.
Duke Power Company, Oconee Nuclear Station, Semiannual Report, Period, Ending June 30, 1974.
9.
Duke Power Company, Oconee Nuclear Station, Semiannual Report, Period Ending December 31, 1974.
10.
Letter Thies to Giambusso re Power Distribution Comparison Status Report dated December 21, 1973.
11.
Letter Thies to Giambusso re Power Distribution Comparison Status Report dated July 19, 1974.
12.
Duke Power Company, Oconee Nuclear Station Unit 1, Startup Report, November 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-1393, June 1973.
16.
Operational Parameters for B5W Rodded Plants, BAW-10078, September 1973.
17.
Letter AEC to SMUD re Request for ATWS Information dated October 19, 1973.
.18.
Letters SMUD to NRC re ATWS Analysis for Rancho Seco dated September 30, 1974, and December 30, 1974.
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Letter SMUD to NRC re Addendum 1 to Proposed Amendment No. 29 dated April 7, 197S.
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A-2
APPENDIX B' 07FICE OF INSPECTICN A'!D E'; FORCE:ENT INSPECTION SUM" DRY SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO DOCRET NO. 50-312 The operation of the Rancho Seco Nuclear Generati,ng Station by the Sacranento Municipal Utility District (SMUD) has been exanined by the Office of Inspection and Enforce =ent since the operating license vas issued on August 16, 1974.
Since issuance of the operating license ten inspections invc been perforced, seven of which were unannounced.
The total amount of tine spent at the plant site during these inspections 4
vas approximately seventy-five can days.
s The results of the NRC inspection pregram to date show that the Rancho Seco'Uucicar Cencrating Station han'been operated safely since initial r.tartup in September 1974.
The perforrance characteristics of the reactor and engineered safeguards systems have been determined dur?ng the startup test progran through 92.6% of full pcuer (256S
- 7. t ).
The startup test results have been found to ccet the test acceptance e !teria determined from the plant design bases described in the Final Sciety Analysis Report.
The significant results of the inspection program that are pertinent the proposed power increase from 2568 MWt to to 2772 17,4t are discussed below.
1.
,Startuo Test Precram The startup test progr.'a cornenced with initial fuel loading on Aup,ust 19, 1974 and was cc pleted through 92.6% of full pcuer
-(2568 MWt) on Y2rch 22, 1975.
Testing was perfor:cd at zero power, 15%, 40%, 75% and 92% cf full power.
The test data obtained during the perforn nce or the following startup tests have been reviewed and evaluated by our inspectors.
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Zero Power Physics Test b.
Nuclear Instrumentation Calibration at Power c.
Core Power Distribution d.* Reactor / Turbine Trip Dropped Control Rod Asse=bly c.
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Power I= balance Detector Calibration
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Muc1 car Stean Supply System Heat Balance h.
Reactivity Coefficient at Pcecr 1.
Psuedo Control Rod Assemb1 E
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111IS DOCUMENT CONTAINS POOR QUALITY PAGES o
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The evaluation of the Psuedo Control Rod Assembly Ejection test,--
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conducted during zero power testing, re'ealed that the ceasured value v
e'J of the rod worth differed significane.ly from the predicted value.
The
,,r licensee informed our inspector that the predicted' value was in error and a reevaluation of the physics calculation indicated that the measured worth should have been~ anticipated.
The resolution of the psuedo ejected rod-worth value and corrective action taken by the licensee are described in the licensce's letter to Licensing (RJR 74-401 dated October' 23, 1974) and in Proposed A=end:ent No. 29 to the FSAR formally submitted on Deccaber 6,' u4.
The data for all other tests evaluated by the inspector were found to be consistent with the predicted values of the parameters censured.
In addition to the above tests, the initial fuel loading, initial criticality, unit loss of electrical load test, and the loss of offsite power test were directly observed by cur inspectors.
During the uitnessing of these tests, co:pliance uith approved procedures was verified and the acceptability of test results uns independently verified by our inspector with no anomalics observed.
A special test procedure, "31ced and Tced De=onstratica", uas used by the licencce to satisfy the requirements of technical specification 3.12 which required that significant load changes be performed by operating personnel to denonstrate satisfactory system operability.
Our inspector revicwed the raw test data cbtained during the performance of the special test and fcund it consistcnt with the infortation contained in the licensec's performance report dated March 1975.
2.
Nnnt Operatione.
Through observations of plant operation, exacination of facility records, cad d'scussion with licensec representatives, the operation of the facil;.y has been fcund censistent eith the requiretents of the technical specifications.
Tours of the facility have been cade on-each inspection.
During these tours, observation of techanical equip =cnt and piping systc=s have shoun no excessive ]caks er'vibraticns.
Instrutentatica system including nucicar instruncatation, reactor protcetien and safety features actuation systens have been feund to be operating normally with the required tests and calibrations being perforced as scheduled.
The unidentified reactor coolant leakage has o
averaged less than 0.4 gallen per minute throughout plant cperatica to date as compared to the technical specification liait of 1.0 gallon per ninute.'
Th5re have been no cajor outages caused by the failure of safety related equip ent or cceponents.
Two outages of significant duration have occurred since initial operation, a 26-day outage in cetober/
November'1974 for'the repair of condenser tube leakage and codification of the turbine stop valves and a 22-day cutage in March / April 1975 for the inspection of turbine bearings.
Maintenance of safety e
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related equip =ent has been recorded by the licensee in Monthly
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Operating Reports.
Examination of maintenance records verified that the maintenance had been perforced consistent. with the licensce's manage =ent control systen.
The plant availability factor since co==ercial operation at 92.6%
power has been 100%. The capacity factor for the facility has been approxicately 95% since co==creial operation.
.3. Unusual Occurrences The licensee submitted fourteen (14) abnor=al occurrence reports in 1974 and to datc *has subaitted nine (9) abnorcal occurrence reports in 1975.
The circumstances and corrective action described by the licensee in each abnor=al occurrence report have been verified during the inspection program.
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Radiologica) Protection Our inspectors have verified that the licensec's radiological protection prograa has been impicnented consistent with regulatory requirements.
Results of radioleg.ical surveys perforced during plant startup tc-sts indicate that the radiation zones, based on current data projected to 100% of full power, will be as described in the FSAR, Section 11.2.1.1.
The results of the radiological environmental conitoring progrer. fer the last two quarters of 1974 did not identify any significant t
adverse environ: ental effects resulting fre= the operation of the i
facility.
The records of routine surveys perfor cd by the licensee have shown that radiation and centcaination levels in-uncontrolled arcas have been insignificent.
5.
Quality Assurance Procran for Operations The implccentation of a Quality Assurance Prograa for Operations was verified prior to the receipt of the Operating License.
The quality cssurance progran has been subject to a continued exanination by our inspectors during the initial operatica phasc.
Ite=s of noncompliance were identified related to inspection planning, receipt inspection performance, independence of inspection personnel and implenentatien*of corrective actica for deficiencies identified by the licensec's internal audit program.
The licensee has promptly responded with corrective action ect=itncets which were subsequently verified by cur inspectors to satisfactorily resolve the enforectent items.
All other cperational activitics have been found to be consistent with the requirc=ents of the quality assurance program.
for operations.
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