IR 05000395/1982005
| ML20042B327 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 02/26/1982 |
| From: | Burnett P, Jape F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20042B321 | List: |
| References | |
| 50-395-82-05, 50-395-82-5, NUDOCS 8203250219 | |
| Download: ML20042B327 (5) | |
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UNITED STATES
h NUCLEAR REGULATORY COMMISSION t
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101 MARIETTA ST N.W.,St>ITE 3100 o,
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ATLANTA, GEORGIA 30303
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Report No. 50-395/82,05
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Licensee:
South Carolina Electric and Gas' Cocoany"
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Col umbi a,'SC '29218 I
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Facility Name:
V. C. Summer
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- Docket No. 50-395
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Licenre No. CPPR-94
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Inspection at V. C. Summe site near Jenkinsv111e, South Carolina M/7'
Inspector-
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Date Si'gned F. Jape', Section
Engineering Inspection Bfanch
Division sf Engineering and.7echnical Progr--
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SUMMARY
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Inspection on January 25-29, 1982
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Areas Inspected
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This routine, announced inspection involved.36 inspector-hours on site in the
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area of procedure review.
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Results
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No items of noncompliance or deviations (ere identified.
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8203250219 820301 yDRADOCM-05000395
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REPORT DETAILS 1.
Persons Contacted Licensee Employees
- W. H. Williams, Jr., General Manager, Nuclear Operations
- 0. S. Bradham, Plant Manager
- J. Connelly, Deputy Plant Munager
- B. G. Croley, Assistant Plant Manager - Technical Support
- S. Fipps, Station Reactor Engineer G. Taylor, Reactor Engineer L. Storz, Assistant Plant Manager - Operations K. W. Woodward, 0parations Supervisor Other licensee employees contacted included technicians, two operators, one shift supervisor, and two office personnel.
Other Organizations B. Wooldridge - Westinghouse
NRC Resident Inspector
- J. L. Skolds, Senior Resident Inspector
- Attended exit interview 2.
Exit Interview The inspection scope and findings were summarized on January 29, 1982, with persons indicated in paragraph 1 above.
The licensee was reminded
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specifically of the requirement of Technical Specification 6.9.1.14 to submit values of F sub x y at least 60 days prior to criticality.
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Licensee Action on Previous Inspection Findings Not inspected.
4.
Unresolved Items Unresolved items were not identified during this inspection.
5.
Outstanding Items from Previous Inspections Nine open and followup items were identified in inspection 81-30. Each was discussed with the appropriate members of the plant staff and management. A meeting of the minds was reached on each issue, but since none of the-corrective actions was yet incorporated into an approved procedure,
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procedure revision or physical work, each item will remain open.
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conclusions of the discussions are given below for each item.
81-30-01 The precautions and limitations specific to the testing program will be presented to the operators as part of the requalification program prior to fuel load.
Further, the precautions and limitations will be removed from the
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administrative procedure AP-1700 and made a part of the test procedures such as ZPT-0,
" Technical Specification Surveillance and Periodic Data Acquisition During Low Power Physics Testing".
81-30-02 The organization charts of the fueling organization in FHP-601, " Refueling Organization" will be revised to clearly show an organization that conforms to technical specifications and IE Circular 80-21. The subordinate role
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of the fueling services contractor will be clearly shown on
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the charts.
Further, the plant manager made a specific
commitment to have licensed operators perform all in-vessel fuel manipulations.
81-30-03 Draft Procedure CST-3, " Initial Fuel Loading" will be revised to address Technical Specifications 3.9.6, 3.9.8, and 3.9.7.2.
The posting requirements of 10 CFR 20 will be addressed in a health physics procedure. Temporary storage
locations for fuel will be specifically identified in CST-3.
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81-30-04 A chi-square test will be based upon at least ten observations of at least 200 counts per observation.
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range of acceptable results will be 5 to 70 percent
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probability that a random sample gives no larger result. In
the event a counter system does not yield an acceptable result, an additional ten observations will be made, pooled j
with the first set and conclusions of systems acceptability based upon the results of the larger sample.
I For CST-3 the test will be conducted only for responding
detectors when first coming on scale, whenever moved, or at least once per twenty-four hours.
For ZPT-1 a chi-square test will be performed each time baseline data are obtained.
81-30-05 The elevation indicator ("Z" tape) on the fuel handling bridge will be color-coded by painting to indicate the critical _ elevations for specific areas of operation such as
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the reactor core, upender and rod-change fixture.
For each area the expected slack cable position will be clearly indicated..The acceptable band about the indicated position
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for gripper release will be plus or minus one quarter inch.
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81-03-06 A test section will be aoded to ZPT-6 to use the results of boron endpoint measurements obtained in ZPT-2.1 through ZPT-2.5 to determine a best estimate dif ferential boron worth in pcm/ ppm.
81-30-07 Notes will be added to ZPT-3.1, ZPT-3.2 and ZPT-3.3 speci-fying the x y recorder range and reactivity scale.
81-30-08 STP-210.002 will be written to provide a formal', approved procedure for establishing rod-withdrawal limit curves in response to the anticipated positive moderate temperature coef ficient with all-rods-out.
81-30-09 A statement limiting the use of the reactivity computer to the dynamically calibrated range will be added to ZPT-0.
6.
Test Procedures The following test procedures were reviewed:
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CST-2,
" Core Loading Instrumentation Check", Revision 0, 5/22/81 b.
CST-5,
" Cold Shutdown Testing Rod Drive Mechanism Timing Test RCS Cold", Resision 0, 4/29/81 c.
CST-6,
" Rod Drop Time Test Measurement RCS Cold No-Flow", Draf t.
According to the licensee this test is procedurally similar to other draf t drop-time tests at different system tempera-tures and flow conditions.
By identifier those tests include: CST-7, HST-9, and HST-10.
lhe content of these test procedures was discussed with licensee personnel.
At the conclusion of the discussion the licensee committed to adding chi-souare tests of the source range nuclear instruments (SRMs) when those instruments are relied upon as the primary monitors of core reactivity. The af fected tests are CST-5, CST-6, CST-7, HST-3, HST-8, HST-9, HST-10 and HST-11. Addition of the statistical tests to the procedures will be tracked as open item 50--390/82-05-01.
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Fuel Handling Procedures The following fuel handling procedures were reviewed:
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FHP-602,
" Limitations and Precautions for Handling New and Partially Spent Fuel Assemblies", Revision 0, 2/21/78 b.
FHP-604,
" Functional Testing of Fuel Handling Systems", Revision 0, 8/9/78
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FHP-611.9 " Refueling Machine Operation", Revision 0, 9/28/78. Change 1 to this procedure dated 10/27/81 responds to IE Information Notice No. 81-23.
Following discussions of these procedures with operations personnel and discussion of the remedial action proposed by the licensee to address open items 81-30-05 (see paragraph 5) the inspector had no further questions.
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Other Documents Reviewed The inspector reviewed:
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" Nuclear Design Report for the SCE&G Company, V. C. Summer Nuclear Station".
This document is the source for the numerical acceptance criteria for the zero power and power-operations test, b.
WCAP-9773 "An Engineering Users Guide to Flux Maps and Flux Mapping",
(8/80) (Proprietory)
One section of this latter document addresses the identification of absorber rodlets broken from rod cluster control assemblies and left incore perturbing the power distribution.
(This issue is also addressed by 0IE TI2515/34.) The example used to argue that careful review of the INCORE code output could identify such a problem was not convincing.
In that example, taken from actual experience, the searcher had good reason from the earlier behavior of a rod drive to anticipate dropped rodlets as well as their exact location in the core. A comparison of the reaction rates of symmetric flux mapping thimbles was used to identify the dropped rodlets.
Three pairs of thimbles showed lack of symmetry indicative of a power perturbation, which was ascribed to the dropped rodlets. However, the fact that three other pairs of symmetric thimbles also showed significantly different reaction rates in locations not affected by the " identified" rodlets was ignored.
The possibility of identifying dropped rodlets using INCORE was discussed at length with members of the licensee's staff.
It was their opinion based upon experience, training and discussions with engineers at operating plants that INCORE analysis could not reliably make such identification.
It is the licensee's opinion that no power distribution analysis beyond that routinely performed using INCORE is feasible or necessary.
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