IR 05000295/1986012
| ML20199L992 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 06/26/1986 |
| From: | Mccormickbarge, Ring M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20199L961 | List: |
| References | |
| 50-295-86-12, 50-304-86-11, NUDOCS 8607100078 | |
| Download: ML20199L992 (13) | |
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U.S.-NUCLEAR REGULATORY COMMISSION
REGION III
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Report Nos. 50-295/86012(DRS);50-304/86011(DRS)
Docket Nos. 50-295; 50-304 License Nos. DPR-39; DPR-48
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Licensee: Commonwealth Edison Company P. O. Box 767 Chicago, Illinois 60690 Facility Name: Zion Nuclear Power Station, Units.1 & 2 Inspection At: Zion, Illinois Inspection Conducted: May 5 through June 5,1986, and June 11, 1986 Inspector:
C k-Barger
Date Approved By:
ef 26 (
Test Programs Section Date Inspection Summary Inspection _on May 5 through June __5_,
19_8_6_and _J_u_n_e ll 1986 (Report o
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y Nos.50-2957~860112DR_S)J 50 3M /86011(DRS))
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Treas Inspected': Routine, announced inspection of licensee action on previous inspection findings (92701), error in Zion Units 1 and 2 Technical
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Specification F Normalized Operations Envelope (92701), isothermal / moderator n
temperature coefficient measurements (61708), control rod worth measurements (61710),-and shutdown margin / estimated critical condition calculation (61707).
Results: Of the five areas inspected, no violations or deviations were identified in three areas; one violation was identified in the two remaining areas (inadequate test control - Paragraphs 4 and 5).
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8607100078 860703 PDR ADOCK 05000295
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DETAILS
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1.
Zion St_a_ tion Personn_el Contacted
- G. Plim1, Station Manager
- T. Rieck, Services Superintendent W. Kurth, Assistant Superintendent of Operations
- C. Schultz, Regulatory Assurance Administrator
- R. Cascarono, Technical Staff Supervisor A. Ockart, Assistant Technical Staff Supervisor
- R. Niederer, Station Nuclear Engineer (Nuclear Group Leader)
T. Petrak, Nuclear Group Engineer J. Kellerhals, Nuclear Group Engineer M. McWilliams, Nuclear Group Engineer P. Zwilling, Station Chemist
- W. Stone, Quality Assurance Supervisor
- J. Ballard, Quality Control Supervisor J. Rappeport, Quality Assurance Engineer CECO Headquarter _s Personnel Contacte_d-J. Silady, Supervisor Plant Support Engineer - Nuclear Fuel' Services Department
- E. Young, PWR Plant Support - Nuclear Fuel Services Department D. Wanner, PWR Nuclear Design - Nuclear Fuel Services Department
- K. Ramsden, Reactor Safety Analysis - Nuclear Services Department J. Bitell, Operations Quality Assurance Manager Additional sta. tion technical and administrative personnel were contacted by the inspector during the course of the inspection.
- Denotes those personnel present at the exit interview.
2.
Licensee Action on Previous Inspection Findings (Closed) Open Item (295/85027-01)':
Inconsistencies in boron sample a.
measurements during startup physics testing.
For the Unit 2 Cycle 9 startup, the licensee took the following actions regarding boron sampling: a) prior to the startup, discussions took place between
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the nuclear group and the chemists to emphasize the.importance of accurate boron sampling measurements, b) a boron standard which was closer to.the concentration of the boron samples being measured for physics testing was used {1250 ppm standard vs. standards of approximately 500 ppm which were used for previous.startup physics testing measurements), and c) three titrations were nade for each reactor coolant system boron sample measurement (previously only one titration was made). The inspector reviewed reactor coolant system and pressurizer boron sample data associated with the Unit 2 Cycle 9 startup. With the exception discussed below, the data appeared to be acceptable. Just prior to performing the' initial criticality procedures, for approximately 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br /> beginning at 8:00 a.m. on
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January 25, 1986, the pressurizer boron measurements were as much as 300 to 600 ppm lower than the reactor coolant system baron measurements. The licensee determined that the pressurizer boron samples were being diluted as a result of water leaking past a closed sample panel valve (Valve RC V1.1) into the line used to take the pressurizer samples. The licensee subsequer.t.ly replaced the defective valve.
b.
(Closed) Open Item (295/85027-02): Lack of timeliness of
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documentation of an evaluation of a failure to meet a test acceptance criterion. During the initial startup for Unit 1 Cycle 9, one of the acceptance criteria for the resistance temperature detector cross calibration startup test was not met. Although this te.st is required to be performed prior to. initial criticality, the evaluation of the failure to meet the acceptance criterion was not documented until after reactor power was at 88%.
During this inspection, the inspector reviewed numerous Unit 2 Cycle 9 startup tests and noted one case in which a test had failed to meet an acceptance criterion related to zero power flux map measurements. An evaluation was documented in Technical Staff Surveillance Procedure TSS 15.6.51, "Zero Power Physics Testing Following Refueling," prior.
to proceeding with the reactor startup power escalation.
c.
(Closed) Open Item (295/85027-04): Startup test procedures lacked signatures and dates to signify completion of the tests to the extent necessary to satisfy surveillance requirements prior to proceeding with a reactor startup. The licensee revised Technical Staff Surveillance Procedure TSS 15.6.51, "Zero Power Physics Measurements Following Refueling," such that Revision 16, dated January 17, 1986, contained places for signatures and dates related to review of zero power physics tests. Whereas the procedure revision was acceptable, the inspector had concerns with respect to the implementation of the procedure revision as discussed in Paragraph 5 of this inspection report.
d.
(Closed) Open Item (295/85027-05): During the initial Unit 1
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Cycle 9 startup, documentation did not exist to show that rod bottom lights cleared at 20 steps. As part of this inspection, the inspector reviewed Technical Staff Surveillance Procedure TSS 15.6.57, Revision 9, " Rod Drup and Timing Test," dated January 13, 1986, and performed for Unit 2 Cycle 9 on February 18, 1986, which had been revised to include steps to check that rod bottom lights go out at 20 steps.
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Normalized 3.
Error in Zion Units 1 and 2 Technical Specification F= ~ ~~ 4
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0_p_e_ra t%FsTnTelope ( K( ZTri g u re )~ ~~
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On May 1, 1986, an individual from Westinghouse notified Commonwealth Edison Company (CECO) that, in the course of reviewing the Zion 1 Cycle 10 Reload Safety Evaluation, he noticed that the K(Z) figure was incorrect. K(Z) values are used to reduce the maximum allowable heat flux hot channel factor (F ) in the upper portion of the core based on n
large Break and Small BreaR Loss of Coolant Accident (LOCA) considerations.
The K(Z) limiting value for the top of the core should have been.
1.42/F max (1.42/2.32 = 0.612) whereas, the value used was 0.647 (which correshondsto1.5/2.32). The K(Z) figure had been taken directly from Technical Specification Figure 3.2-9.
A brief history of activities pertinent to this issue is given below.
February 29, 1980 The NRC issued a confirmatory order to Zion which required that the heat flux hot channel factor be limited to 2.20.
(The Technical Specifications were not changed, plant ope' rations were simply administratively limited to this limit.)
November 3,1983 Westinghouse transmitted an Emergency Core Cooling System (ECCS) Large Break LOCA Reanalysis to CECO which, in the absence of the February 29, 1980, Confirmatory Order, permitted a heat flux hot channel factor of 2.32.
Included in the transmittal was a new F Normalized Operating Envelope Figure n
(hereYnafter referred to as the K(Z) figure) in which the third segment, corresponding to small break LOCA analysis limitations on F,.was nonconservative in thatthe12-footcoreelehationK(Z)valuewasgiven as 0.647 rather than 0.612.
September 24, 1984 Onsite Review of the Zion LOCA Reanalysis.
September 27, 1984 Offsite Review of the Zion LOCA Reanalysis.
January 24, 1985 Westinghouse transmitted an additional Zion LOCA reanalysis.
January 29, 1985 Onsite Review of the additional Zion LOCA. reanalysis.
(The erroneous K(Z) figure.was included in the review package.)
February 7, 1985 Offsite review of the additional Zion LOCA reanalysis.
(The erroneous K(Z) figure was included in the review package.)
April 10, 1985 Westinghouse transmitted a Zion LOCA reanalysis with a new input methodology.
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April 12, 1985 Onsite Review of the Zion LOCA reanalysis with the new input methodology.
(The erroneous K(Z) figure was included in the review package.)
April 18, 1985 Offsite Review of the Zion LOC.A reanalysis with the new input methodology.
May 24, 1985 The NRC approved Technical Specification Amendment-89/79 which included the nonconservative figure for K(Z).
June 10, 1985 Onsite Review of Technical Specification Amendment 89/79.
D.ccember~30, 1985 The NRC approved celetion of Items Al and A2 from the February 29, 1980, Confirmatory Order which eliminated the F limit of 2.20.
q January 15, 1986 CECO received NRC's December 30, 1985, letter.
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April 14, 1986 Values were changed in the computer code that Zion uses to analyze flux mapping data (the INCORE code)
from K(Z) values that corresponded to the F limit imposed by the February 29, 1980, Confirmathry Order to K(Z) values that corresponded to the nonconservative Technical Specification figure for K(Z).
Between April 14 One flux map was run to Unit 2 on April 30, 1986.
and May 1, 1986 Two flux maps were run for Unit 2 on May 1, 1986.
May 1, 1986 An individual from Westinghouse notified an individual from CECO Nuclear Services that, in the course of reviewing the Zion 1 Cycle 10 Reload Safety Evaluation', he noticed that the K(Z) figure
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was incorrect.
May 1, 1986 The analysis.of the April 30 and May 1,1986, flux maps had not yet been completed when an individual
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from CECO Nuclear Fuel Services contacted the Zion Station Nuclear Engineer to notify the Station Nuclear Engineer of the K(Z) error.
May 2, 1986 All three flux maps mentioned above were rerun.
Prior to rerunning them, the INCORE computer code inputs
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(Note: The original erroneous flux maps were never used.)
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Immediate Corrective Actions Within one day of being notified of the error in the Technical Specification K(Z) figure, the K(Z) inputs for.the computer code that Zion uses to analyze flux mapping data were corrected.
Long Term Corrective Actions A Technical Specification change will be submitted to the NRC to correct the K(Z) figure by August 15, 1986.
CECO's Quality Assurance Department will audit the Westinghouse LOCA Analysis Group in October 1986.
CECO's Nuclear Fuel Services (NFS) Department will make procedural changes to establish procedural guidance for the NFS review of special analyses perfonned by organizations other than CECO (LOCA
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analyses are considered special analyses).
NFS intends for the procedural guidance to be as specific as possible while still encompassing the wide variety of special analyses reviewed by NFS.
At the time of this inspection CECO was in the process of laying the groundwork for a Zion Interface Document that would combine requirements for safety-related parameters, that are currently contained in numerous documents, into a single document.
In developing this document, particular attention will be paid to the small break LOCA requirements.
The Nuclear Fuel Services Department will be involved earlier in
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the Zion Technical Specification change process for future Zion Technical Specification changes related to reactor core performance.
Safety Significance It appeared that there was no immediate safety significance in that the licensee did not use the Technical Specification K(Z) figure to analyze any flux mapping data because K(Z) values based on aAfter NRCConfirmatory(Orderwerebeingused.
February 29, 1980, the Confirmatory Order related to K Z) was lifted, Zion continued to use the confirmatory order values until April' 14, 1986, at which time-the K(Z) input values in the flux mapping analysis code were revised to correspond to the erroneous Technical Specification K(Z)
figure. However, the K(Z) error was recognized and corrected prior to completing the analysis of any Zion flux maps.
In addition, a sampling of flux maps at the beginning of Unit 1 Cycle 9 and the end of Unit 2 Cycle 8 indicated that a margin of about 30% existed between actual F values and the F limit based on the correct K(Z)
valuesatcoree9evationscorrespokdingtotheerroneousportionofthe K(Z) curve.
Finally, based on.a Westinghouse. calculation which used a K(Z) curve that was conservative with respect to the erroneous Zion
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Technical Specification K(Z) curve, the peak clad temperature following a small break LOCA was 1967 F which is less than the 10 CFR 50.46 limit of 2200 F (and also less than the large break LOCA peak clad temperature.of 2159 F).
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10 CFR 50, Appendix B, Criterion III requires that measures be established to assure that regulatory requi.rements and design bases are' correctly translated into specifications. Technical Specification 3.2.2 (Figure 3.2-9) contained erroneous
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(nonconservative) data with respect to a limitation imposed on the heat flux hot channel factor from May 24, 1985, to June 1986, which is a violation of 10 CFR 50, Appendix B, Criterion III. However, no Notice of Violation will be issued at this time based on the provisions of 10 CFR 2, Section V.A. which states that the NRC will not generally issue a notice of' violation for a violation provided, among other requirements, the violation is licensee identified and.
it was or will be corrected, including measures to prevent recurrence, within a reascnable time. With the exception of the requirement that the violation was or will be corrected, including measures to prevent recurrence, within a reasonable time, all of-the other requirements of the third paragraph of 10 CFR 2, Section V.A.
appeared to have been met. Since corrective actions were not-complete at the time of the inspection and, based upon the commitments alone, it was not possible for the inspector to assess the~ adequacy or timeliness of the corrective actions, this will be an unresolved item (295/86012-01; 304/86011-01) pending NRC review of the corrective actions against 10 CFR 2, Section V.A.,
Paragraph 3, Item (4) during a subsequent NRC inspection.
Because a vendor (Westinghouse) originally prov>ided the erroneous K(Z) data to CECO, this matter has been' referred to the NRC Vendor Branch for possible additional followup action.
This program area requires further review and evaluation and is considered to be an unresolved item as discussed above.
4.
Isothermal /Moderat_o_r_ Temperature Coefficient
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The inspector reviewed licensee procedures and results to verify that results obtained were within acceptance criteria and consistent with'
Technical Specifications, and that any discrepancies were properly evaluated. The inspector utilized the~ following procedures during the review:
Technical Staff Surveillance (TSS) Procedure TSS 15.6.51,
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Revision 16, "Zero Power Physics Measurements Following Refueling,"
dated January 17, 1986, and performed for Unit 2 Cycle 9 on January 27-29, 1986.
Procedure TSS 15.6.54, Revision 3, " Isothermal Moderator Temperature Coefficient Measurements," dated May 27, 1982, and performed for Unit 2 Cycle 9 on January 27-28, 1986.
Negative Moderator Temperature Coefficient Control Calculation, performed for Unit 2 Cycle 9, (no date given) and referred to in
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Step 7.23.c of Procedure TSS 15.6.51 performed for Unit 2 Cycle 9.
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Memorandum fron1 the Zion Technical Staff Superv'isor and Station'
Nuclear Engin~eer to the Unit 2 Operating Engineer concerning..
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operating guidelines and restrictions for the Unit 2 Cycle 9 initial power ascension, ZNG:86-007, dated January 29, 1986.
Report NFSR-0037, Revision 0, " Nuclear Design Report for Zion Unit 2 Cycle 9," dated December 1985.
With respect 'to negative moderator temperature con' trol.via establishing.
- rod withdrawal limits, the inspector noted the following:
(a) During this cycle the licensee used administratively established control rod withdrawal limits to-ensure that the moderator coefficient' would be negative, but Technical Specification. 3.2.1.C.1,
.which states, "Immediately prior to startup, the reactor coolant, temperature shall be shown to be greater than the temperature above which the. moderator temperature coefficient is always negative (except during low power physics tests)-and greater than 500 F,"
does not specifically address the use of rod withdrawal limits to
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maintain a negative moderator temperature coefficient. However, a1though the Technical Specifications are not-as clear as they could
be,' acceptability of using this method is implied in the Basis portion of the Technical Specifications which states, "When control rods are inserted, the temperature at which the moderator coefficient becomes negative is lower so that at the temperature determined during the physics tests and with the operational control rod program, the coefficient is expected to be negative."
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(b) The calculation of control rod withdrawal limits to assure a negative moderator temperature coefficient contained two errors; one which affected the calculation results-and~one which did not. The' error that did not have any impact on the results involved
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writing the equation for ITC=MTC+DTC (isothermal temperature
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coeffjcient = moderator temperature coefficient + doppler temperature coefficient) incorrectly so that the magnitude of the.
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value obtained was correct but it was positive when it should have been negative. However, when the value was-used in the calculation it was correctly" applied as a negative value. The second error was related to the fact that in one place a predicted boron concentration value.was 'used and in~anot.her a measured boron concentration value was used, This reduced the cbnserva'tism in the
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l calculation by 0.3 pcm/ F.
A conservatism had_been added to th'e i
calculation based on a recosendation in the Nuclear Design Report which equaled 0.5~.'pcm/ F and, therefore, the error reduced the
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conservatism from 0.5 pcm/ F to 0.2 pcm/ F.
The failure of the licensee's independent ~ review to identify and correct the errors in the calculation of control rod withdrawal-limits', to assure a negative moderator temperature coefficient, is considered an example of a violation of 10 CFR 50, Appendix B, Criterion XI
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(304/86011-02a(DRS)). A factor which may have contr.ibuted to this
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violation was that the Negative Moderator Temperature Coefficient Control Calculation had always been performed from-scratch for each
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violation is minimal in that, although the intended conservatism was reduced, it was not eliminated by.the~ errors and, in addition, although there was no procedure requiring a check to ensure the results of-the calculation were conservative as compared to actual measured data, the Station Nuclear Engineer stated that he had performed this check.
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An example of one violation was identified. No deviations were identified.
5.
Control Rod Worth Measurements
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The inspector reviewed licensee procedures and results to verify that control rod worths were within acceptance criteria and consistent with Technical Specifications, and that any discrepancies were propecly evaluated. The inspector utilized the following documents during the review:
Technica'l Staff Surveillance (TSS) Procedure TSS 15.6.55, Revision 2, " Rod and Boron Worth Measurements," dated August,15, 1984, and performed for Unit 2 Cycle 9 on January 27, 1986.
Letter from Henry E. Bliss, CECO Nudlear Fuel Services Mana G. J. Plim1, Zion Station Manager, " Zion 2, Cycle 9 Rod Ex'ger.to change
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Data," Z2C9/049, dated December 13, 1985.
Procedure TSS '15.6.53, Revision 2, " Boron Endpoint Measurement,"
dated January 13, 1986, and performed for Unit 2 Cycle 9 on January 27, 1986.
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Procedure TSS 15.5.51, Revision 16, "Zero Power Physics Measurements Following Refueling," dated January 17, 1986, and performed for Unit 2 Cycle 9 on January 27-29, 1986.
.Theinspectorhadthefollowingconcernsbaseionthereviiw:
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Procedures TSS 15'.6.51, TSS 15.6.53, TSS 15.6.54 (refer to Paragraph.4),
and TSS 15.6.55 are required to be performed during zero power physics testing, that is, at less than or equal to 5%.of rated thermal power, during the initial startup following a refueling. The zero power physics
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testing for Unit 2 Cycle 9 was performed at. the end of January 1986.
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This inspection began slightly over three months later and, in reviewing
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the aforementioned procedures, the inspector observed the following: -
(a) The following procedure steps, which in each case were the.last
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steps in the procedures, were not signed:
(1) Steps G.1, G.2, and G.3'of Procedure TSS 15.6.54 dealing with data sheet signoffs related to establishing initial conditions.
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(2) Steps A.5.6, A.S.6.1 and B.5.11 ofe Procedure.TSS 15.6.55' ",
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dealing with completing data sheets-(Step A.5'.6) and'with
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evaluating test data in accordance with the, method specified
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by the associated step (Steps A.5.6.1 and B.5.11.1).
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these procedure steps were not actually at. the end of the precedure, these steps were related to results evaluation and a note was present prior to Part A of this procedure that stated that all parts will be evaluated together in Section 8.0.
Although Section 8.0 did not exist in the
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procedure, the licensee explained that the. intent was for all parts to be evaluated in the last section, the test evaluation section.
With respect to item (1), the licensee stated that the initial conditions for the isothermal / moderator temperature coefficient test were appropriately established and similar steps at "
another point in the procedure had been initialed.
With respect to item (2), the licensee stated that tihe work associated with the unsigned steps had been performed prior to proceeding with the initial power escalation, the associated
. procedure steps simply were not signed.
(b) The following procedure steps were skipped (not signed nor performed before proceeding to the next step):
Steps 7.7.b and 7.9.b of Procedure TSS 15.6.51 deali'ng with
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evaluating rod worth test data.
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Contrary to the situation discussed in Paragraph (a)(2), this procedure did not provide for performing the above steps out of sequence, although.in discussions with the licensee, the licensee stated that it was always their intent to perform these evaluation steps at the end of the procedure since there was no need for the evaluation to be performed before proceeding to the next step. Hence, performing these. steps.
out of sequence had no impact on the test or test results,.
With respect to the lack of signatures, the licensee stated that the evalu.ations associated with the steps were performed prior to proceeding with power escalation,"the associated steps a
simply were not signed.
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(c). Procedure TSS 15.6.51, Step 7.23.a dealt with review of a zero
power physics test, specifically it stated, " Based on Rod Vorth measurements. adequate shutdown margin exists.'-
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Theinspectorobservedthattherodworthmeasuhementdata sheets attached to the master copy of Procedure TSS 15.6.55,-
" Rod and Boron Worth Measurements," were blank.
Per discussions with the licensee, the' inspector learned that the
signature associated with Step 7.23.a was-based on data fro'm a data sheet that was largely in pencil, was dated-but not signed,
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and was still not attached to the master copy of the* procedure
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at the time of the inspection which was over three months after
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did not have a position for a signature, there were places for
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signatures associated with rod worth measurement data evaluation in Procedures TSS 15.6.51 (steps 7.7.b and 7.9.b),
.TSS 15.6.53 (Step 5.13)., and TSS 15.6.55 (Steps A.S.6 and B.5.11) that had not been signed at the time of the.
inspection. -(The steps in the previous sentence correspond to steps already discussed in items (a)(1) and (a)(2) above.)
Items (a), (b), and (c) above, specifically, skipping procedu're, steps,'failing to sign the final steps of procedures before
proceeding -to an operational phase for which performance of th,e procedure was required, and utilizing test results calculations
'to make decisions that could affect plant safety which had not.
been signed by anyone 'and in addition, were not cl.early a part
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of the master procedure either by documented reference or by attachment to the procedure, are considered examples of a violation of 10 CFR 50, Appendix B, Criterion XI (295/86012-02b(DRS)).
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Several examples of a violation were identified. No deviations were identified.
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Shutdown Margin / Estimated Critical Condition Calculation In the area of shutdown margin and estimated critical condition calculations the inspector. reviewed various ' procedures, procedure
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results, and Zion in.ternal memoranda for technical adequacy and against'
Technical Specification requirements. The inspector utilized the following documents during the reviews:
General Operating Procedure G0P-2, Revision 13, " Plant Startup,"
Steps 3,.16.e, and 16.u (steps related to estimated critical boron concentrat, ion calculations).
Generai Operating Procedure ~GOP-4, Revision 17, " Plant Shutdown
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and Cooldown," Steps 25, 26, 30, 31, 40, 62, and 73 (steps related
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Mem.orandum,- dated September 6,1985, from the Zion Technical Staff Supervisor aod Station Nuclear Engineer to the. Unit 2 Operating"
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Engineer concerning required reactor coolant system boron
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concentrations'to assure a 10% shutdown margin throughou't the refueling outage.
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Memorandum, dated January 10, 1986, from the Zion Technical Staff
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Supervisor and Station Nuclear Engineer to the Unit 2 Opera. ting Engineer, " Required Reactor Coolant ~ System Baron Concentrations,"
ZNG:86-004.
Startup Boron Concentration. Calculation, performed by the Nuclear-Group for' the period of, time between fuel moves and startup testing
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for Unit 2 Cycle 9 (no.date given).,
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Periodic Test Procedure, PT-0, Revision 89, " Surveillance Check-
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lists and Periodic Tests," Appendix G, " Operating Surveillance
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Checklist," Item No. 2, " Verify Control Rod Insertion Limit, Technical Specification 4.2.1.D,1," completed for Unit 2 from February 2,1986', through March 1,1986.
Technical Staff Surveillance Procedure TSS 15.6.29; Revision 3,
" Reactivity Anomaly Check," dated January 17, 1986, and performed for Unit 2 on January 27, 1986.
Data Sheet titled, " Calculation of Estimated Critical Condition,"
associated with General Operating Procedure G0P-0, " Plant Startup
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Documentation Requirements," Revision 13, and performed for Unit 2 Cycle 9, Startup No. 3 on May 19, 1986.
Report NFSR-0037, Revision 0, " Nuclear Design Report for Zion Unit 2
Cycle 9," dated December 1985.
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Although the-inspector did not review the results of the following calculations, the inspector noted that the licensee performed PT-0,
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Appendix B-1, Sheet 2, " Shutdown Margin Calculation," for Unit 2 on March 24,1986, and May 19, 1986.
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(a) With respect to the estimated critical condition calculation performed for Unit 2 Startup No. 3, the inspector noted that.the rod limits in Step E were not, indicated as negative numbers as
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they should have,been. This error had no impac.t on the calculation results. The error was discussed with the Assistant Superintendent of Operations who stated that the calculation would be reviewed to
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see if any clarification would be beneficial.
(b) With respect to Sheet 2 of PT-0, Appendix B-1, " Shutdown Margin Calculation," the inspectof observed that the second note on the
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96/86 definition of shutdown margin which stated, " SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritichl or would be subcritical fro'm its present conditioh assuming all control and shutdown banks are. fully inserted, except'
for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn."
In discussions with the licensee, the licensee agreed and explained that Technical
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Specification Amendment 96/86 was approved by the NRC but would not.
become effective until September 24, 1986, and, although previous Technical Specifications did not contain a definition of Shutdown Margin, the plant pers.onnel's understanding was that it was that I
amount of reactivity by which-the reactor is subcritical. With that
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definition, 'it would not be possible to approach criticality and,
maintain a "sub" criticality margin at the same time.
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Note: Historically, shutdown margin was typically defined as that amount of reactivity by which the reactor is subcritical. More
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recently, shutdown margin has taken on the definition specified in Zion's. Technical Specification Amendment 96/86.) The licensee intended-to review surveillance procedures and revise them in accordance with Technical Specification Amendment 96/86 prior to September 24, 1986, and accordingly, the second note on Sheet 2 of
~PT-0, Appendix B-1 will be removed prior to September 24, 1986.
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The inspector had no further concerns based on the review.
No violations or deviations were identified.
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General Inspection Observations The inspector observed that numerous nuclear group staffing changes had taken place since the last core physics related inspection in August 1985.
There were three individuals in the nuclear group (including the group
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leader) in August 1985 that 'were no longer in the group at the time of this current inspection. Om individual hadsjoined the group just a few weeks prior to the current inspection and the licensee had plans to add
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additional personriel to the nuclear engineering group in the near future.
The conibination of considerable personnel turnover, coupled with examples'
of failure to adhere to or impl,ement administrative controls for test performance (refer to paragraphs 4 and 5), indicates the need for close management oversight of nuclear engineering group activities including, but not limited to, the area of administrative controls for test performance.
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8.
Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, open items, deviations, or violations. An unres.olved item disclosed during the inspection is discussed in Paragraph 3.
9.
Exit Interview r
The inspector met with licensee representatives (denoted in Paragraph 1) -
-
on June' 5,1986, to discuss the -scope and findings of this inspection and the likely informational content of the forthcoming inspection repgrt.
The licensee acknowledged the statements made by the inspector and ' tated s
that the CECO Nuclear Design Report,and the Westinghouse Small Break LOCA
,
~
Bounding Analysis, referenced within this report, were considered l
proprietary, but references to these documents would not be considered proprieta ry. The insp.ector performed some additional reviews following the June 5,1986, exit interview related to. negative moderator temperature coefficient control and shutdown margin and discussed the results of this review with the S,tation Nuclear Engineer via tele. phone
,
on June 11, 1986.
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,
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