IR 05000454/1987026

Insp Rept 50-454/87-26 on 870624-0708.Violations Noted.Major Areas Inspected:Action on Previous Findings,Core Power Distribution Limits,Core Thermal Power Evaluation & Shutdown Margin/Estimated Critical Condition Calculation

ML20236C095
Person / Time
Site:	Byron
Issue date:	07/23/1987
From:	Azab B, Wright G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20236C062	List: IR 05000454/1987026 ML20236C059 ML20236C071
References
50-454-87-26, NUDOCS 8707290336
Download: ML20236C095 (6)

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U. S. NUCLEAR REGULATORY COMMISSION REGION'III Report No. 50-454/87026(DRS) -)

i Docket No. 50-454 License No. NPF-37 1 l

Licensee: Commonwealth Edison Company l Post Office Box 767 Chicago, IL 60690 y

Facility Name: Byron Station, Unit 1  ;

i Inspection At: Byron Site, Byron, Illinois

Inspection Conducted: June 24 through July 8, 1987 A Q& \

Inspector: B. A. Azab 7/J73/89 l Date/ /

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Approved By:

CLkN$b.[

G. C. Wright 7[7Sk7 l

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TestPrograms(Ch'e7-Section Date' '

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In,spection Summary Inspection on June 24 through July 8, 1987 (Report No. 50-454/87026(DRS))

Areas Inspected: Routine unannounced, safety inspection of licensee action on previous inspection findings (92701), core power distribution limits (61702),

core thermal power evaluation (61706), shutdown margin / estimated critical condition calculation (61707), isothermal and moderator temperature coefficient determinations (61708), and control rod worth measurements (61710).

Results: Of the six areas inspected, no violations or deviations were identified in five areas, and one violation was identified in the remaining area (failure to implement an abnormal operating procedure - Paragraph 2b).

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DETAILS 1. Persons Contacted

• R. E. Querio, Station Manager

• D. W. Berg, Nuclear Safety

• P. Brochman, Resident Inspector R. Choinard, Nuclear Group Leader

• M. Farr, Nuclear Engineer

• F. A. Hornbeck, Technical Staff Supervisor

• B. Klinger, Senior Quality Assurance Engineer

• Snow, Regulatory Assurance Supervisor

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• Stauffer, Assistant Technical Staff Supervisor

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• L. A. Sues, Assistant Superintendent of Operations

• R. M. Williams, Primary Group Leader

• E. M. Zittle, Regulatory Assurance Staff The inspector also interviewed other licensee personnel during the course of the inspection including members of the. technical staf * Denotes persons attending the exit meeting on July 8, 1987.

2. Licensee Actions on Previous Inspection Findings (92701 and 92702) (Closed) Violation (50-454/87016-01(DRS)); Operating Surveillance 1805 0.1-6 and Procedure B0P 199-T6 were not properly implemented. The licensee has taken the following actions to correct the deficiencies and to clarify the instructions: (1) issuance of a daily order on April 8, 1987 to notify personnel of the difference in the reference water level for the reactor cavity and spent fuel pit, (2) revision of procedure B0P 199-A70 to clarify the reference level difference and (3) distribution of a memo, dated June 3, 1987, to all operating personnel re-emphasizing the requirements of Technical Specification 6.8.1.a. The inspector has no further concerns in this are (Closed) Open Item (50-454/87016-02(DRS)): The low level Spent Fuel Pit (SFP) alarm was continuously enunciated during the. refueling, thereby defeating its purpose, to alert the operators of SFP level loss. The only other SFP level indication during this time was a ruler bolted to the SFP wall which was bent inward, therefore having questionable accuracy. The licensee performed a geometric analysis to determine the error in the readings of the SFP ruler, which proved the error was negligible and in the conservative directio This portion of the open item is considered close Further review of this item identified that the SFP low level alarm setpoint was raised from 423' 6" to 424' 2" in February,1987 before the SFP was flooded. The change was in response to a letter from Sargent & Lundy dated December 19, 1985 which recommended the l

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licensee raise the minimum SFP water level to 424' 2". The 424' 2" water level corresponds with an FSAR commitment which states that ten feet of water be maintained above the top of the fuel assemblies ;

during all handling operations. The Sargent & Lundy letter further ,

recommended that the licensee modify all fuel handling tool racks such that the tool. handles would remain above the higher water l level. The licensee did not perform any modifications on the fuel ?

handling tool racks. Discussions with members of the licensee's -

staff indicated that the SFP water level was kept below the low {

1evel alarm setpoint to keep the fuel handling tools from being submerged and fuel handling personnel from working with their hands ;

in the wate f On March 25, 1987 the SFP low level alarm came in and remained in continually during the refueling activities. Annunciator response procedure, BAR 1-1-C1, " Spent Fuel Pit Level High Low," directs the !

l operators to immediately enter the abnormal operating procedure, 180A Refuel-3, " Spent Fuel Pool Level Loss - Unit 1," when the level goes below the low level alarm setpoin BOA Refuel-3 gives the following instructions:

• Empty transfer system of all fue ,

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• Place fuel in a safe positio '

• Stop all fuel movemen * Isolate spent fuel pool from reactor cavit l l i The licensee neither performed the steps listed in 1 BOA Refuel-3, !

nor took remedial actions to modify the procedure such as writing a :

temporary procedure change or performing an evaluation of the lower SFP water level to permit continued operation at a lower setpoin Technical Specifications, Section 6.8.1 states that written procedures shall be established, implemented and maintaine Contrary to this, the licensee failed to adhere to abnormal operating procedure 1 BOA Refuel-3, " Spent Fuel Pool Level Loss - Unit 1," during the Byron Unit 1 core reloa This is considered a violation (454/87026-01(DRS)).

No additional violations or deviations were identified.

l 3. Core Power Distribution Limits (61702)

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The inspector reviewed a number of procedures and test results concerning core power distribution limits, and verified that they were technically adequate, petformed with the proper frequency and complied with Technical Specification requirements. The following documents were used during this review:

• IBVS XPT-3, " Reload Startup Physics Tests Following Refueling,"

Revision 0, dated May 15, 198 _ _ _ _ _ _

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• IBVS 3.3.2-1, " Moveable Incore Petector's Operability Check,"

Revision 4, dated June 8, 198 {

• B0P IC-3T1, "Incore Moveaole Detectors Checkoff Sheet," Revision 52, dated June 18, 198 .i

• 1BVS 3.1.-5, "Incore - Excore Axial Flux Quarterly Calibration," l Revision 1, dated June 12, 1987.

l * 1805 2.1.1.a-1, " Axial Flux Difference Weekly Surveillance," J Revision 53, performed June 17 and June 23, 1987.

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• Byron Curve Book - 1, Figures 19A through 19 I

• IBVS 2.2.2-1, " Heat Flux Hot Channel Factor Checkout Using Peaking {

Factors," Revision 5, performed June 13, 15 and 19, 198 j

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• IBVS 2.3.2-1, " Nuclear Enthalpy Rise Hot Channel Factor Check," !

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Revision 4, performed June 13 and June 19, 1987.

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• Incore Flux Maps: "BY10219," dated June 18, 1987 and "BY10204,"

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dated June 12, 1987 l .

• 180S 2.4.1.a-1, " Quadrant Power Tilt Ratio Calculation," l Revision 51, performed June 11, 18 and 22, 198 '

• Operator's Log Book No. 4 i No violations or deviations were identifie . Core Thermal Power Evaluation (61706)

The inspector reviewed a two week sample (June 5 through 18, 1987) of the completed surveillance, 180S 3.1.1-2, " Calorimetric Calculation Daily l Surveillance," Revision 52 and verified that the nuclear instrumentation system's power range channels were properly adjusted to equal the calorimetric power when required. The inspector verified the technical adequacy of the licensee's hand heat balance procedure, 1805 3.1.1-2, by independently performing a calorimetric for July 1, 1987 and obtaining the same core thermal power as the license The inspector noted during the review that the licensee has been administrative 1y limited to 98 percent core thermal power due to excessive feedwater flow in the lower nozzle of 1A steam generato The licensee is currently performing an analysis which will permit operation at 100 percent core thermal powe No violations or deviations were identified.

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5. Shutdown Margin / Estimated Critical Condition Calculations (61707) i i

i The insp3ctor reviewed the following licensee's procedures and test j results c1ncerning shutdown margin and estimated critical condition i calculati os: j BV', XPT-4, " Initial Criticality After Refueling and Nuclear Heating Lr/el," Revision The inspector reviewed the results of the i

., surveillance performed May 26 through 29, 1987 and verified that the 1 reactor. achieved criticality within 50 ppm of the predicted critical "

boron concentration as required by procedur B0S 1.1.1.1.c-1, " Predicted Critical Control Rod Position 4 Surveillance," Revision 5 The inspector reviewed the results ]

from the surveillance performed on May 27, 1987 and verified they l complied with Technical Specifications 4.1.1.1.1.c and 3.1. j .1.1.1.e-2, " Shutdown Margin Surveillance During Operation,"

Revision 51. The inspector reviewed the surveillance performed on May 31, 1987 ana verified that it complied with Technical Specifications 3/4.1. No violations or deviations were identifie . Isothermal and Moderator Temperature Coefficient Determination (61708) i i

The inspector reviewed 1BVS 1.1.2.a-1, " Moderator Temperature Coefficient j Low Power - BOL," Revision 4, performed June 2, 1987 and found one '

concern. The Doppler temperature coefficient was recorded improperly as j a positive number; however, it was used correctly, as a negative number, in the moderator temperature (MTC) equation. The inspector performed several independent calculations of the MTC and obtained the same results as the licensee. The inspector verified that the surveillance results complied with Technical Specification 3.1.1.3.

I No violations or deviations were identifie . Control Rod Worth Measurements (61710)

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l The inspector reviewed 1BVS XPT-5, " Rod and Boron Worth Measurements,"

Revision 0, and verified that the results were within the limits specified in the acceptance criteri During the procedure review the folicwing c9ncerns were identifie Step 2.4.3 of the procedure instructed the user to complete Table 2.1, but should have read Table 2.2. The procedure is currently being revised to reference the correct Table, The note preceding Step'1.10 was worded such that it was confusin The licensee is revising the procedure to be technically correct by the use of absolute value sign _ _ - _ _ _ _ _

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, Acceptance Criterion G.2.1 read "The measured / inferred N rod worth must be > 90 percent of the predicted N rod worth." The inspector-did not linow the definition of "N rod worth" and asked the lead-nuclear engineer who was also unsure of the meaning, but subsequently determined it to mean total rod worth. The inspector

,. was concerned that an acceptance criterion existed in a procedure l that was not fully understood by those using the procedur However,. the inspector noted that if the test results complied with l acceptance criterion G.1.3 they would automatically comply with criterion G.2.1. .The inspector also noted that other members of the i nuclear engineering staff claimed to have understood the meaning of

"N rod worth." The procedure is currently being revised to define ;

N rod worth as it is used in the acceptance criteri No violations or deviations were identifie . Exit Interview

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The inspector met with the licensee representatives (denoted in

, Paragraph 1) on July 8, 1987. The inspector summarized the scope and findings of the inspection. The licensee acknowledged the statements made by the inspector with respect to the violation (denoted in !

Paragraph 2b). The inspector also discussed the likely informational ;

content of the inspection report with regard to documents or processes reviewed by the inspector during the inspection. The licensee did not i i

identify any such documents / processes as proprietar ]

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