IR 05000454/1986045

Safety Insp Repts 50-454/86-45 & 50-455/86-41 on 861018- 1114.Violations Noted:Failure to Perform Adequate 10CFR50.59 Safety Evaluations & Inadequate Corrective Actions Re Primary Process Sampling Preoperational Test

ML20212M383
Person / Time
Site:	Byron
Issue date:	01/21/1987
From:	Ring M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20212M063	List: IR 05000454/1986045 ML20212M061 ML20212M068
References
50-454-86-45, 50-455-86-41, NUDOCS 8701300105
Download: ML20212M383 (25)

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t j U.S. NUCLEAR REGULATORY COMMISSION

REGION III

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Report flo: 50-454/80045(DRS)

50-455/86041(DRS)

Docket flo: 50-454 License No: f4PF-37 50-455 License No: NPF-60 Licensee: Conmonwealth Edison Con'pany

Post Office Box 767 l

Chicago, IL 60690 Facility Name: Byron Station, Units 1 and 2 Inspection at: Byron Station, Byrcn, IL i

Inspection Conducted: October 18 - November 14, 1986 Inspectors: M. L. McCormick-Barger D. A. Beckn.an P. G. Brochman

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S. G. DuPont P. L. Eng J. M. Hinds W. E. Gunther

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I R. M. Lerch l

J. A. Malloy

. J. H. Neisler i

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W. Shier Approved By: i Ri g f /[7 /'[M _

Test Programs Section Date l I

I_n, spec _tio_n Sununary Inspection on October 18 - Nevenber 14 1986, 50-454/~86045(~DPS)1

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50-4 5 5/ 86'O'41]DRS)T ' ' ' ~' ' ' ' ' ' ' " ~ ~ ~ ~ ~ ',' ' " ' , ,(, Rep ~ orts No .

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_reas _In,spected: Routine, announced safety inspection of licensee action on K l

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previous inspection fiodings; evaluations of preoperaticnal test results; j

preoperational test results verification; preoperational test program I

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implen.cntation review; startup test procedure review; an electrical train 8701300105 DR 070121 ADOCK 05000454 l

PDR

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separation issue; and a diesel generator breaker issue. Reactive, anncunced safety inspection to review actions regarding the independent technical review of preoperational tests.

I Resul ts: Of the eight areas inspected, no violations or deviations were identified in six areas; two violations were identified in the remaining two areas: (failure to perform adequate 10 CFR 50.59 safety evaluations -

Paragraph 2j and inadequate corrective actions pertaining to a primary process 1 sampling preoperational test - Paragraph 4a). In addition, one violation was

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discussed in Paragraph 7 which was docun.ented in a Braidwood Inspection Repor Furthermore, three unresolved items were identified, one of which is considered a violaticn; however no violaticn is being issued in this report pending consideration for escalated enforcement action.

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DETAILS 1. Persons Contacted Conmonwealth Edisen Company R. Querio, Station Manager B. Shelton, Connonwealth Edison Projects Engineering Manager L. Sues, Assistant Superintendent, Operations

1. R. Ward, Services Superintendent,

1. R. Tuetken, Startup Superintendent H. Campbell, Onsite Project Engineering Supervisor

1. D. Peters, Project Engineering - Onsite

1. E. Falb, Unit 2 Testing Supervisor F. Hornbeak, Technical Staff Supervisor P. Devine, Unit 2 Assistant Technical Supervisor, Test Review Board

1. E. Zittle, Regulatory Assurance Staff

1. J. Langan, Regulatory Assurance Staff

1. W. Pirnat, Regulatory Assurance Staff i K. Yates, Nuclear Safety

1. F. Willich, Quality Assurance Additional station technical and administrative personnel were contacted by the inspectors during the course of the inspectio # Denotes those present during the exit interview on November 14, 198 . Action on Previous Inspection Findings (92701)

_ (Closed) Unresolved item (455/84054-01): verify that the Main Steam Isolation Valves (MSIV) are preoperaticnol tested to denonstrate the override capability of the steam line isolation signal during the Fast Exercise Mode of the active and standby trains of the MSIV The inspector verified by reviewing Preoperational Test 2.51.67,

"ftain Steam - Integrated Hot Functional," Section 9.5 that the Fast Exercise Mode of all active and standby trains of the MSIVs was tested, (Closed) Unresolved Item (455/84054-02): verify that acceptance criteria are properly annotated on preoperational test steps. The inspector verified by reviewing Preoperational Tests RC 63.60,

" Integrated Hot Functional (IHF)," RC 63.67, "Reacter Coolant System IHF," MS 51.67, " Main Steam IHF," RY 69.67, " Pressurizer IHF," and CV 18.67, " Chemical Volume Control - IHF," that test steps and data acquisition points where acceptance critei ta bEre verified or evaluated were ider.tified with a cent sign notation. The inspector also verified that appropriate instructions were contained in the Startup Manual, Revision 26, Section (Cicsed) Open Item (454/85033-02(DRS)): verify that appropriate precauticns are included in Byrcn cooldown procedures to preclude a recurrence of the evcnt docuniented in Inspection Report 454/EEC33 in I

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which a control rod was mechanically stuck due to differential '

thermal expansion during a plant cooldown while the ccntrol rods were energized on the core bottua. The inspector reviewed Byron Procedure EBGP 100-5, " Plant Shutdown and Cooldown," Revision 52, Stcp 11.L which opens the Reactor Trip Breakers to prevent thermal

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lcckup or binding of the control rods.

4 (Closed) Safety Evaluation Report item (455/83000-08) and Open items

(454/85055-01;455/85045-01): containraent sump vortex control. The

licensee performed an analysis, of containment sump geometry, as discussed in the proposed revision to Regulatory Guide 1.82, for Braidwood Units 1 and 2. Byrcn also performed the calculations and verified that Byron's sump geometry was similar to Braidwood's,

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showing that the designed sump geometry was within the parameters demonstrated in NRC studies to be indicative that the sump is capable of performing its intended function without being degraded by the development of a vortex.

j (Closed) Open Item (455/86005-01(DRS}): two testing deficiencies

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were identified during Preoperational Test AB 01.60, " Boric Acid,"

- and were closed to Action Item Record (AIR) No. 6-86-2038. Deficiency 61197 involved the capability of the Byron Unit 2 boric acid filter to handle the cepacity from simultaneous two pump operation as defined in Final Safety Analysis Report (FSAP) Section 9. ;

(150 + 5 gpm ficw). The maximum flow obtained through the filter during testing was 126 gpm. Deficiency 61199 involved slightly excessive vibration of Boric Acid Pump 2AB03 Retest R-2023 i

reperformed the test scctions pertaining to the deficiencies and, through the Project Engineering Department's (PED) review, the vibration signatures were found to be acceptable per the design specifications (expected range of "slightly rough" or better);

however, the recorded filter flow was cutside of the expected range (135spm). To resolve the filter flow deficiency, the Nuclear Steam 1 Supply System Vendor, Westinchouse, was requested to evaluate the deficiency. The inspector reviewed the Westinghouse resolution letter for Byron and Braidwccd (CAW-10189/CBW-5666), which determined that 135 gpm flcw meets the requirements of FSAR Section 9.3.4. i and that the expected range of 145 to 155 spm is not a requirement of the design basis. The design basis of the system was a flowrate of

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30 to 85 gpm to the charging pbmp suction. The upper limit (85 gpm)

was established to prevent possible runout (100 gpm) and for Het

' Positive Suction Head consideraticns. This upper limit also represented the original design basis of the system. This design basis was related to the possibility of using the emergency boration line as an alternate shutdcwn system in the event that the control rods were inoperable. The purpose was to insert sufficient negative i reactivity to shutdchn the reactor from 100% power to 1% power in 90 l mir.utes with equilibrium xenon. This capability is described in l

Section 9.3.4.1.3 of the FSAR. Westinghcuse had determined that a flowrate of 57 gpm of 4 weight percent of boric acid wculd acccmplish this function. Westinghcuse also determined that the actual recorded test flew of 55 9pm was within acceptable tolerance and, as such, should meet the design basis of the FSAR. Also, the Technical

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Specifications require a flowrate of 30 gpm for operability of the

boric acid transfer pumps (with either one or two pumps). Since the design basis and Technical Specifications are based upon a one pump cperation and Westinghouse had determined that the only impact of 135 spm versus 145 to 155 gpm (two pump operation) is on the boric acid transfer and recirculating capabilities, which is not safety or l

accident related, PED accepted the preoperational test results. The

, inspector fcund these evaluations to be adequate and agrees with the conclusions. Since this flowrate does not affect safety, accident related systems, or design bases and is within 101; of the expected .

range, the inspector has no further concerns.

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r (Closed) Unresolved Item (455/86010-01(DRP)): verify that diesel generator preoperational testing, consisting of a minimum of four cold diesel generator (CG) starts from a single air receiver withcut recharging the air receiver between starts, was accomplished as required by Safety Evaluation Report Section 9.5.6. The inspector

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verified by reviewing preoperational data associated with Preoperational Tests DG 22.60, "2A Diesel Generator," and DG 22.61, "2B Diesel Generator," that, for each air receiver, four starts were accomplished without recharging the air receiver. In addition, the inspector verified that the testing conditions included initial air receiver I

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pressure at or below 175 psig which is the minimal operating and l Technical Specification allowed pressure. Regarding the requirement

that the diesel generator starts be from cold ccnditicns, flRR

provided the interpretation that the DG should be at ambient

conditions (lube oil and jacket water temperature 120-130*F as l deternint.d by the licensee) prior to each start, tiRR and Region III concluded that a single air receiver can provide sufficient pressure to perform four cold starts of a DG based on the following: A) the DG's could be considered to be at ambient conditions at the start of i

the testing since they had r.ot been operated for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> previous to the start of the testing, B) although no lube oil or jacket water temperature measurcments were taken during the air start tests in question, temperature data taken during other air start

testing shcwed that these temperatures tend to decrease on each successive DG start for the case in which the diesel generator is started at ambient conditions, run for a short period of time, a then stopped and restarted and, C) the CG was only run for short

periods between starts as evidenced by the fact that all 9 starts (4

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for one air receiver and 5 for the other) for DG 2A were accomplished

, within 75 minutes and all 11 starts for DG 2B (5 for one air receiver

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and 6 for the cther) were accomplished within 63 minutes. This verified that the designed air receiver volume would be adequate in  !

the worst allowed condition to start the diesel generators four  ;

consecutive tirces without needing to be recharged.

I The above paragraph addresses only the Unit 2 diesel generator The same concern applies to Unit 1 and will be followed as an

- unresol ud item (454/86045-01(CRS)) pending NRC review of Unit I diesel generator air start receiver testing dat .

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g. (Closed) Open Item (455/86010-02(DRP)): verify that new computer software is installed for RTD cross calibration function. This problem was identified during Braidwccd Unit l's Integrated Hot functional (IHF) testing. The inspector serified that the software deficier.cy was identified f or Byron Unit 2 (Deficiency No. 64827)

and resolved by installing "E-Proms" software and tested in Preoperational Test IT 47.E0, "Incore Thern.ocouples".

h. (Closed) Unresolved Item (454/85017-01(DRS)): For Unit 1 Startup Test RY 69.30, a deficiency was not generated as required by the Byrcn Startup Manual for pressurizer heater phase currents and powers This item is similar in whichwereoutsidetheirexpectedranc(es.CRS))

nature to the violation (455/86034-01 discussed in Paragraph 3 of Inspection Report 455/8003 In Inspection Report 455/86034, the licensee was asked to address the adequacy of corrective actions for Unresolved Item 454/85017-01. Therefore, Unresolved Item 454/85017-01 is closed since the NRC's review of the licensee's response to Violation 455/86034-01 will enccrr. pass the licensee's response concerning Unresolved item 454/85017-0 i. (Closed) Unresolved Item (455/86034-02(DRS)): Various preoperational test deficiency forms did not provide justification or engineering basis for accepting the deficiency as is. For each of the varicus deficiencies (identified in Paragraph 3.e.(1) of Inspection Report 455/86034), the inspector verified through interviews during this inspection that the design bases were discussed and evaluated by the Test Review Board, Project Engineering Department, and the System Test Engineers. Based on these interviews, the use as is disposition appeared to be acceptable. In addition, the licensee revised the Startup Manual to require the Test Review Board to ensure that a technical justification be provided or referenced for all " accept as is" draf t deficiency dispositicn (Closed) Unresolved Item (455/86034-05 (DRS)): Acceptability of 10 CFR 50.59 Safety Evaluations for preoperational and startup test During the referenced prior inspection, eight of ten safety evalu-aticos reviewed were fcund inadequate in that the evaluations were inconsistent with Byrcn Administrative Prccedure BAP 1210-5,

"'O CFR 50.59 Safety Evaluation Procedure".

10 CFR 50.59 permits a licensee to conduct tests not described in the FSAR or revise procedures described in the FSAR so long as the licensee has determined that the activity dces not involve an unreviewed safety question. 10 CFP,50.59 further requires that the records of these determinations include a written safety evaluation which provides the bases for the determinaticn that an unreviem d safety questicn does not exis BAP 1210-5 is applied to Byron Unit 2 precperational and startup tests in accordance with the Byrcn Startup Manual (Sections 3 and 2.2) to evaluate Byron Unit 2 tests which could have a safety irrpact en the licensed operation of Byrcn Unit . .

i BAP 1210-5 and Form BAP 1310-T19 (ircplemented by BAP 1210-5) require that checklists be cortpleted to document the evaluation. Form BAP 1310-T19 requires that checklist questions be answered with a "yes" cr "no" and that the " specific reasons justifying the decision" be recorded. DAP 1210-T5 requires evaluation of system interactions and review of key docurc.cnts (FSAR, Technical Specifications, license conditions, etc.) for potential irrpact by the tes Typical responses to the BAP 1310-T19 questions were reiteration of the procedure's evaluation questions with no additional supporting information provided. The BAP 1210-T5 review items were all answered as either "ncne", or N/A (not applicable).

During Inspection 455/86034, a Byron Test Review Board representative indicated that the above deficiencies were licensee identified ard were the subject of additional staff training during late 1985 and early 1986. A limited resampling during Inspection 455/86034 (three exartples after February,1986) indicated some improvemen During the current inspection, additional safety evaluations post dating the retraining were reviewed. Several were found to be satisfactory but six were found to be inadequate. Again, no basis S.as provided for the Form BAP 1310-T19 safety evaluation question responses (cuestions were merely restated at. declarative statements with no specific basis for the determinaticn given), "none" and

"fi/A" were again entered on the BAP 1210-T5 evaluation checklist items for system interactions and source document review These inadequacies were found in:

DG 22.62, " Diesel Generator Lead Test," evaluation dated February 22, 1986 Retest 2067, VQ 94.60, " Containment Purge," evaluation dated September 20, 1986 Retest 2020, DG 22.61, "2B Diesel Generator Test," evaluation dated May 25, 1986 Retest 2021, DG 22.61, "2B Dieccl Generator Test," evaluation dated May 25, 1986 Retest 2049, VD 86.61, " Integrated Diesel Generator, Miscellanecus Electrical, and Switchgear Room Ventilation," evaluation dated August 13, 1986 Component Demonstration 2061, MS 51.61, " Main Steam (Safety-Related Power Operated Relief Valves)," evaluation dated Septerrber 18, 1986 Failure to provide the basis for the determinaticn that an unreviewed safety question does not exist ccnstitutes a violation of 10 CFR 50.59 (455/86041-01(DRS)).

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One violation and one unresolved item were identified during the revicw of this area as discussed abov . Preoperational Test Results Evaluation (70400)

The inspectors reviewed the results of the below listed preoperational test procedures to verify all test changes were identified and approved in accordance with administrative procedures; all test deficiencies were appropriately resolved, reviewed by Iranagar.ent, ar.d retested as required; test results were evaluated by appropriate engineering personnel and specifically compared with acceptance criterie; data were properly recorded, signed, dated and documented as test deficiencies if out of tolerance; test packages were reviewed by (A for adequacy of contents; and test results were approved by appropriate personnel:

AF 3.60, " Auxiliary Feedwater" AP 5.61, " Bus Loadirg and Independency" CS 17.60, " Containment Spray" CV 18.60, " Chemical and Volume Control - Volun.e Cytrol Tank and Charging Pumps" CV 18.61, " Chemical and Volurre Control - Charging, Letdown and Reactor Coolant Pump Seal Injection Logic" CV 18.67, " Chemical and Volume Control - (IMF)"

DC 21.60, "125 Volt DC Power Distribution (Safety Related)"

DG 22.61, "2B Diesel Generator" DG 22.62, " Diesel Generator Load Test" EF 26.60, " Engineered Safeguards Features Actuation" EF 26.61, "Energency Core Cooling System (ECCS) Full Flow" EF 26.62, " Engineered Safeguards Features Logic and Time Respcnse" IP 46.60, " Instrument and Control Power" MS 51.67, " Main Steam Safety Related (IHF)"

NR 52.60, " Neutron Monitoring (Excores)"

RC 63.61, " Reactor Coolant" RC 63.63, " Reactor Vessel Level Indication System (Heated Junction Thertroccuple System)"

RH 67.60, " Residual Heat Rer.. oval"

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RP 68.60, " Reactor Protection Response Time" RY 69.67, " Reactor Coolant Pressurizer - Integrated Hot Functional (IHF)"

SI 73.60, " Safety Injection - Component Checks"

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SI 73.61, " Safety Injection - Accumulators" SI 73.62, "ECCS Flow Balance Test" SI 73.63, " Safety Injection ECCS Check Valve Operability and Leakage (IHF)"

VP 93.61, " Containment Ventilation (IHF)" With respect to RY 69.67, the Project Engineering Division letter dated August 15, 1986, stated that the retest of Sections 9.3 and 9.4 would te tracked and performed per the requirements of Action Item P,ecords (AIR) 6-86-2067 and 6-86-2068, respectively. Review of AIR 6-86-2067 revealed that additional test requirements were imposed to determine the 2PC455A pressurizer pressure controller signal error due to the integral portion of the controller. The system test engineer stated that a procedure incorporating the requirements of RY 69.67 and the additional requirements delineated in AIR 6-86-2067 would be written, reviewed and approved prior to use. Completion of the review and approval of the new test procedure is an open item (455/66041-02(DRS)).

' The NRC inspectors had not completed their review of the quality assurance audits performed for CV 18.60, CV 18.61, CV 18.67, AP 5.61, DG 22.62, and IP 46.60 at the end of this inspection. This will be

! tracked as an open item (455/86041-03(DRS)) pending completien of the

review during a subsequent inspectio Preoperational Test CS 17.60 did not meet Final Safety Analysis Report (FSAR) Table 14.2-18 in that containment spray ring nozzle

, ficw testing did not include the use of infra-red thermography (apparently due to test methodology difficulties). Instead, flow was verified by placing a hand in the flow stream for each flow nozzle. After noting several apparent differences between the FSAR and the preoperational program, the hRC requested that Connonwealth Edison Cor:pany (CECO) identify all instances in which differences existed between the Dyron FSAR and the plant as t;uilt and teste Failing to use infra-red thermography during CS 17.60 was one of eicht differences that CECO notified the NRC of during a telepho conversation with NRR on October 27, 1986. Ceco committed to submit FSAR changes for each of the eight instances. The inspector has no further concerns in this are Subsequent to the corrpletion of Preoperaticnal Test AF 3.60,

" Auxiliary Feedwater (AF)," the licensee made several design changes to the AF system to improve controlability and danipen out some j 9

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dynamic transients. System functions following the design changes will be verified by three retest Retest R-2058 will test the control loop for the 2AF005 valve i

following the installation of breakdcwn orifices downstream of each of the 2AF005 valves and coincident changes to the ficw indicating switches and alarms upstream of cach of the 2AF005 valves. The inspector will review Retest R-2058 upon its ccmpletion and this acticn will be tracked as Open Item (455/86041-04(DRS)).

Retest R-2088 will test the performance of AF pump suction pressure transmitters following their replacement with wide range transmitters, and will include verification of alarm, control, and protective features. The inspectcr will review Retest R-2088 upon its completion and this action will be tracked as Open Item (455/86041-05(DRS)).

Retest R-20E9 will test the ability of both AF pumps to start simultaneously and not have a low suction pressure trip or switchover to Essential Service Water after the installation of the new suction pressure transmitters tested in R-2088. The inspec?.or will review Retest R-2089 upon its completion and this action will be tracked as Open Item (455/86041-06(DRS)).

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During the review of AF 3.60, the inspector identified a

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with the test method for the Essential Service Water (SX)prcblen Booster Pump 2SX04P. IE Bulletin 83-05 Action 1.C required the licensee to conduct a pump performance test for ASME Nuclear Code Pumps manufactured by the Hayward Tyler Pump Company. The licensee stated in a letter from P. L.- Barnes to J. G. Keppler, dated August 12, 1983, that only two Hayward Tyler pumps were used at Byrcn Station. The pumps are identified as Essential Service Water Booster Pump ISX04P and 2SX04P. They are driven by an accessory shaft on the diesel engines which power the 18 and 28 AF pumps. The purpose of the SX04P pumps is to circulate cocling water through the AF diesel during a complete icss of all AC power at Byron Station, when normal SX is not running. Part V of the Attachment to the P. L. Barnes letter stated that the results of tests on the Unit I and Unit 2 pumps would be provided after testing was completed. The Unit 1 response was

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provided in a letter from P. L. Barnes to J. G. Keppler, dated January 10, 1984. Inspection Report 454/83061 documented NRC's

! review and acceptance of the Unit I response.

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IE Bulletin (IEB) 83-05, Action 1.C states that as a minimum the performance test should contain the criteria of Attachment 2 to the IEB. Attachment 2, Paragraph B requires that five measurements be obtained: head vs. flow, vibration, temperature, motor current, and leakage; these measurements are to be taken at normal, minimum, and

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runout flow conditions. Paragraph C. requires that purt.p rundown be observe A review of AF 3.60 indicated that the five measurements had only been taken at normal flow, r.ot minimum and runout flow condition Additionally, pump rundown was not cbserved. The inspector discussed l

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these concerns with the licensee management and staff and the licensee agreed with the inspector that the AF test did not verify that the 2SX04P pump would meet all the requirements of IEB 83-05. The licensee believed that the satisfactory review of IEB 83-05 in Inspection Report 454/83063 for Unit I created a reasonable belief

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that the test method for the Unit 2 test, AF 3.60, was adequate. The inspector agreed with the licensee's position, however, the licensee

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is still obligated to comply with the requirements of the IEB. The licensee connitted to perform the additional testing cn 2SX04P and submit the results prior to initial entry into Mode 3 in Unit 2 and this is discussed further in Inspection Report 455/86040. The licensee con.mitted to perform additional testing on ISX04P at the next available opportunity and this is discussed further in Inspection Report No. 454/8604 The inspector expressed a concern to the licensee that Sections 9.7.35 and 9.7.53 of AF 3.60 did not provide adequate evidence that contacts 5-9 on relay K8 and contacts 5-9 cn relay K10 had functioned properly. Relays K8 and K10 are protective relays for the diesel engine which drives the 2E AF pump. The inspector discussed these concerns with PED. PED believed that the test in its entirety adequately demonstrated the functioning cf the K8 and K10 relays.

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However, the licensee agreed to do an additicnal functional check of these contacts during a retest of AF. This connitment is documented in AIR 6-86-2228. The inspector agreed with this method and will review the results of the retest as previously state e. With respect to VP 93.61, the inspectors had the following consents:

As part of the Containment Ventilation System, the Reactor Containment Fan Coolers (RCFCs) provide two functions: 1) maintenance of containment air temperature during normal operations (high speed fan operation) and 2) post-accident heat removal (low speed fan operation). Final Safety Analysis Report (FSAR) Table 6.2-56 and the test's acceptance criteria require Reactor Containment Fan Coolers (RCFCs) to have capacities of:

High Speed 94,000 cfm (cubic feet per minute)

Low Speed 59,000 cfm Per the FSAR, high speed capacities are based upon maintaining the bulk centainment temperature below 120 degrees Fahrenheit during normal operation (Technical Specification Limit). The icw speed capacity is based upcn one of two trains (two fans per train)

providing post-accident containment heat removal of 132 Million BTU /hr at ccntainment conditions of 271 degrees Fahrenheit, 50 psig, and 0.189 lb/cu. ft. density of the steam-air atmosphere (about 3 times the pre-accident estimated air density of 0.0754 lb/cu. ft.).

Per a Contronwealth Eoison Company Project Engineering Department

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letter of July 17, 1986, initial preoperational test data, which was taken in the upstream duct at ambient conditions, could not provide \

a positive determination of fan performance due to inconsistencies

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in the data obtained during the execution of the test and, therefore, retesting was requircd for Fans 2VP01CA, B, C, and D at both high and low speeds. The retesting was performed in Retest f2046 after minor system tuning. The retest data was also unsatis-factory as indicated below since it was less than the acceptance criteri DATA _ FR_0M, DRAF_T__EhGINEERING EVALUATION

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FAN 2VP01CA 2VP01CB 2VP01CC 2VP01CD INITIAL PREOPERATIONAL TEST DATA (cfm)

HIGH SPEED 94,276 93,891 93,122 94,276 LOW SPEED 63,245 63,415 62,722 65,070 RETEST 2C_4_6_ DAT_A_,(cfml HIGH SPEED 88,645 86,566 83,997 86,475 LOW SPEED 53,744 56,620 55,207 54,439

" ADJUSTED" RE_TE,S_T__20_46_ D_AT A (_c_fm)

HIGH SPEED 90,752 94,042 84,164 89,104 LOW SPEED 54,992 58,238 55,277 55,810 VARIATION - ADJUSTED RETEST DATA TO ACCEPTANCE CRITERIA HIGH SPEED -3.5% -1% -10.4% -5.2%

LOW SPEED -6.7% -1.3% -6.3% -5.4%

• NOTE: For each data set, the licensee extrapolated the test data frcm an:bient test conditions to the design basis conditions, high density, temperature, pressure, et A Comrcnwealth Edison Ccmpany (CECO), Project Engineering Department (PED) letter of October 14, 1986, approved the results of Retest 2046 based on:

(1) High speed capacity will be verified to be adequate during Startup Test 2.093.80, " Heat Capacity Verification for Primary Containment Ventilation System" at the 5% reactor power testing plateau by verification of containment temperatur The inspector verified that the objectives of that test will acccmplish the above by measurement of containment tenperature This is considered an open iten (455/86041-07(DRS)) pending performance of the test and subsequent NRC revie (2) Low speed capacity was evaluated by PED and the project Architect Engineer and found acceptabl . _

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The inspector requested the engineering data used to substantiate Item (2) above, met with the coanizant PED engineer on October 29, 1986, and was provided a draft { unapproved) Architect Engineer evaluation. This evaluation stated that the reduced low speed fan capacities would have " minimal impact on the containment heat removal ar.d the flow rate obtained is technically acceptable" flo additional bases for the October 14, 1986 approval letter was availabl The licensee was advised that the above evaluation was inadequate in that it failed to quantitatively establish the acceptability of the RCFC accident mode (low speed) performanc CFR 50, Appendix B, Criterion XI, " Test Control," requires that test results be documented and evaluated to assure that test requirements have been satisfie CECc Quality Assurance Manual, Quality Requirement 11.0, " Test Control," requires, in Section 11.3, that test data that does not meet test requirements be recorded and, if necessary, reported in accordance with Quality Requirement 15.0, " Nonconforming Materials, Parts or Con.ponents". CECc Quality Assurance Manual, Quality Requiren.ent 15.0, requires, in Section 15.2, that when items are accepted "as-is", a technical evaluation be performed to assure that the final condition of nonconforning items will not adversely affect Code requirements, safety, operability, or maintainability of the items or of the compcnent or system in which it is installed and that technical evaluation actions relating to nonconforming items will be documented and such documentation will be retained and traceable to each item. Accepting test data "as-is" without performing an analysis of the affects of the demonstrated RCFC fan flow rates being lower than those taken credit for in accident analyses is considered a violation of 10 CFR 50 Appendix B, Criterion XI. However, a Notice of Violation will not be issued until the licensee completes further evaluation and/or testing (refer to the discussicn below), at which time the NRC will review this matter to determine if escalated enforcement action is appropriate. Until that decision is made this matter will be tracked as an unresolved item (455/86041-08(DRS)).

The licensee representative advised the inspector that several additicnal factors had affected the test data and had been considered in the conclusions reached, although not specifically addressed in the evaluatio (1) Initial construction phase RCFC test data was taken with pitot tubes and manometers sensing directly at the fan inlets and at the same upstream point as the preoperational test. The licensee advised that the fan ir.let data showed higher (acceptable) fan flows than did the subject preoperational test, VP 93.6 (2) For the preoperaticnal test the pitot tube taps were installed about 30 - 40 feet upstream of the fans in the vertical RCFC inlet headers. The licensee stated that this point of measurement resulted in ductwork pressure drop reducing the observed fan velocity pressures and calculated flcw . . - - . . .- -. . - - - _ _ - . - -

, . .

(3) The RCFC duct and plenum sections are also subject to air i

inleakage because the system's joints are not sealed or caulked (per design requirements prohibiting joint sealant). The licensee stated that air inleakage thrcugh the ductwork was downstream of the point of preoperational test flow measurement and upstream of the fan inlet. Such inleakage would result in a reduction in the cbserved fan flows but would be a component of actual fan flow. The licensee stated that the construction phase tests with measurements made both in the upstream duct i and at the fan inlet confirmed this positio (4) The licensee further stated that the inherent inaccuracies and variabilities in pitot tube flow measurements further contributed to the measured flow data error as illustrated by the variation

'

in data between the tests and retests.

No calculations or analytical bases were provided nor available to j support the above conclusion The inspector requested that the licensee quantify the above effects, i if possible, to substantiate their position and RCFC acceptabilit The licensee advised the inspector that the RCFC unit test results were still considered acceptable by CECO with no further analysis or

! FSAR amendment necessary based en the above and " engineering l judgement," but that the requested information would be provided.

! When advised that the absence of analytical basis was unacceptable, the licensee contacted the NSSS and RCFC vendor and the cognizant PED engineer advised the inspector on October 30, 1986, that the

! design of the RCFCs is based upon circa 1978 laboratory testing of prototype units in near/ actual post accident environment This laboratory testing reportedly demonstrated that the RCFC cooling

- coils had substantial excess capacity (initially reported as 400 -

l 500%, later revised by the licensee to 10 - 20% excess capacity).

However, the licensee's RCFC coil vendor had advised that correlation of the laboratory test data with preoperational test data taken at ambient conditions was not practical. The laboratory data was taken with high tu:perature and density two-phase flow conditions which could not be correlated with conventional calculational method On November 4,1986, the inspectors again met with the Startup Superintendent requesting the status of the licensee's evaluation of the above. The inspectors were informed that a reanalysis by the NSSS of containment heat loads was in progress and would be complete on or about November 25, 1986. Further, additional confirmatory testing was being considered for post-core heatup phase operation _. ._ _ _ _ _ . - _ _ _ . . _ _ . _ - - _ _ __

_ . . _ - _ . _ . _ _

. _ _ . - . _ - =- - . - - -

. .

During subsequent conference calls ar.d n.eetings on November 5 and 6, 1986, the licensee provided additional discussion of their rationale (above) and, on Noven.ber 7,1986, provided another draft supplenent to the original October 14, 1986, test evaluation and acceptance letter.

The November 7 docunient accepted the preoperational test results

'

based cn applying correction factors for inlet duct inleakage determined from the construction phase testing. The construction i

test data measured at the fan inlet and in the upstream duct showed that the total A Train flow was about 103.6% of the FSAR requirement and the B Train about 101%. The inspector advised the licensee that the use of construction phase data to meet preoperational test acceptance criteria would require the licensee to demonstrate that

the construction phase test methods and controls were at least

equivalent to those required for preoperational testing. However, this became a moot issue, since adding inleakage factors to the l preoperational test data only increased the total fan flows to about 98.4% of the FSAR values for the A Train and to abcut 97.5% for the B Train as shown below

FAN: 2VP01CA 2VP01CB 2VP01CC 2VP01CD

,

CONSTRUCTION TEST DATA (cfm):

AT FAN: 61,324 58,718 60,938 60,496 AT DUCT: 59,432 56,632 57,758 58,352 CALCULATED

I ENGl

1,892 ?_,086 1,180 2,144 PREGP DATA ADJUSTED FOR LEAKAGE f

FLOW: 56,691 57,239 59,368 57,788 i

The November 7, 1986, licensee document also stated that, based upon the laboratory test data (discussed above) the excess cooling coil l performance of 10 - 20% provided at least an additicnal 7% margin on i overall heat removal even wher, considering the reduced flows of the preoperational test data. The inspector advised the Startup

, Superintendent that such a relationship could not be considered l valid unless the correlation between lab test and preoperatfor.al

test conditions could te definitively established.

During the various discussions with licensee representatives, the inspector was advised that the Byrcn Unit 2 RCFC fan performance was

similar to that experienced on Byron Unit I during the same tests i (seebelcw). The Unit I reactor containment fan coolers were l considered acceptable by the licensee on bases similu to those ,

applied to the Unit 2 reactor contairment fan coolers. The inspector reviewed the Unit 1 RCFC test (VP 93.10) which was conducted during

.

l 15

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_- - .. - - . - - _ .... - - - _ _ ._ - _ _ - - -

.

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August 1984 with later retests perfomed to confirm fan fichs lower than the acceptance criteria by 1% to 7%. The final Unit 1 RCFC data was approved by Ccan.onwealth Edison Company, Project Engineering Department letter dated September 4,1984. That letter stated: " Low flows are acceptable based cn the fact that heat removal capabilities

,

are adequate. Fan coolers are capable at (sic) removing design heat loads. No impact on test results." The acceptability of the Unit 1 test evaluation will remain an unresolved item (454/86045-02(DRS))

pending NRC Region III review of the licensee's reevaluatio Discussion among NRC Region III management, NRC Nuclear Reactor Regulation (NPR), and the inspectors detemined that at least the following licensee actions were necessary to support Unit 1 and 2 licensed operations:

(1) If the " actual" RCFC flows are less than the FSAR values, an FSAR change must be processed to permit NRC staff review in that the FSAR values are considered minima. The submittal must include revised fan duty cycle curves or equivalent data, a discussion of the licensee's analysis, and the significance of the chang (2) If the licensee traintains the position that the observed test flows are acceptable, a quantitative basis for their

acceptability must be provided and is subject to further review

by the NR (3) The status of the Unit 1 RCFC units must also be reviewe ! The licensee's Startup Superintendent was advised of the above by

the inspectors and Region III managerrent on October 30, 1986 at which time he advised that this matter would be identified as an open testing item in a letter to NRR listing test program items requiring deferral for Unit 2 operating license issuance. This

letter was provided for the inspector's infortration. Further, he

'

stated that the status of the Unit 1 RCFCs would be reviewed in a i similar ligh During discussions between the inspector and the licensee regarding verification of Byron Cperating Prccedures (60P), an individual frcm the Unit 2 operating staff stated that it was permissible to skip steps in a B0P that are not applicable and that a temporary procedure change would cnly be needed if steps needed to be added to (' a procedure. In addition, the individual stated that this would be permissible even after Dyren Unit 2 was issued an operating licens The inspector discussed this with the Assistant Superintendent of l Operations who stated that unless specifically allowed by the

, procedure in use, the procedure tr.ust be performed as written. To

- ensure that all members of the operating staff were aware of this requirement, a " Daily Orders" form was issued which required plant

operators to revit.w Byrcn Administrative Prccedure (BAP) 300-1,

"Ccnduct of Operations," Paragraph C.4 which identifies requirerrents

, for compliance with plant procedures. The inspector revicwed BAP

!

!

_-_ - - - , -- --,, - - , - - - - - - - , , . - -, , , , - - - - - - - - - - , - , , - - , , - , , , _ , .--,_ . .

300-1, Paragraph C.4 and found the requirements regarding procedural compliance to be adequate. The hRC will continue to mcnitor performance of the Byron operating staff in the area of procedural cctpliance during routine inspection Two unresolved items and six open items were identified during the inspection of this program are . Preoperatic_n_al Test Results Verification (70329)

The inspectors reviewed the following preoperational test procedures and verified that results were reviewed against approved acceptance criteria and an evaluaticn of the test results had been performed in accordance with Regulatory Guide 1.68 and the licensee's Startup Manual:

.

CC 10.60, " Component Cooling" FH 32.60, " Fuel Handling Tools" FH 32.61, " Fuel handling Transfer" FH 32.62, " Fuel Handling Building Crane" FW 34.60, " Main Feedwater - Isolation Valves" HC 39.60, " Polar Crane" MS 51.60, " Main Steam Isolation Valves" PS 61.60, " Primary Process Sampling" PS 61.61, " Process Sampling Hydrogen Monitors" RY 69.60, " Reactor Coolant Pressurizer" SX 76.60, " Essential Service Water" VD 86.60, " Diesel Generator Ventilation" VD 86.61, " Integrated Diesel Generator, Miscellaneous Electrical, and Switchgear Room Ventilation" VQ 94.60, " Containment Purge" VX 99.60, "Switchgear Ventilaticn" VX 99.61, "Switchgear Heat Removal" VE 128.60, "Miscellanecus Electrical Equipment Room Ventilaticn" LM 133.60, " Reactor Lcose Parts Monitor" 17 .

-

. .

Based on the preoperational test results verification review, the inspectors had the following consents: During the review of Preoperational Test PS 61.60, the inspector noted that two problems were encountered during testing that were dispositioned by TRB/ PED through the " rationale of Byron Unit 1" Retests R-252 and R-243. One of the documented problems (test deficiencies BY and CE) pertained to Acceptance Criteria 4.1, " grab sanples from the reactor coolant module and radwaste module should have ... a temperature of no more than 6 F higher than the Component Cooling (CC) water which enters the sample coolers." The other problem (test deficiencies AM, AQ, AS, AZ, BF, BT, BU and BZ) pertained to the sample purge flow rates (documented in inches of water). The inspector found that in both cases, these conditions had previously existed and were resolved during preoperational testing of Byron Unit 1 but the resolutions were not incorporated into the Byron Unit 2 preoperational test. Details regarding each of these cases are provided in the following paragraph (1) Deficiencies BY and CE Both deficiencies were written against Acceptance Criteria 4.1 based en test data which exceeded the 6 F temperature difference between the component cooling inlet temperature and the sample coolers effluent temperature as recorded at test steps 9.16.23(BY) and 9.9.23(CE). (The recorded teraperature differences were 8 F and 7.5 F for Test Steps 9.9.23 and 9.16.23, respectively.) The dispositions of the deficiencies appeared to be inadequate in that accepting the test results based on the sample temperature being

"below a temperature that would scald the sampler" did not fully address the design requirements of the sample cooler In addition to the dispositien provided with the deficiencies, the TRB/ PED acceptance letter documented that these deficiencies were closed based on the rationale of the Byron Unit 1 Retest R-243, which stated that the intent of the sarrple coolers pertained to protecting both the equipment and personnel. The design requirement as provided in the Sentry Equipment Corporation Manual (Vendor Manual No. F551) is that the sample effluent temperature from the reactor coolant sample module is a maximum of 101 F and the sample effluent temperature from the radwaste sample module is less than 96 F. The inspector verified that, in both cases, the recorded sample effluent temperatures were within the vendor manual requirements. Additionally, the rationale of Retest R-243, as contained in the September 6, 1984, letter from G. T. Klopp to R. E. Querio, stated that the acceptance criterion of 6 F differential temperature should be deleted based upon the fact that the combined instrument tolerances of the CC inlet temperature instrument 1T1-0674(+6 F),

the Doric instrument used to measure the sample effluent temperature (0.7 to 1.0 F), and the unkncwn tolerance of the

. .

thermocouple probe used with the Doric is larger than the acceptance criterion of 6 F. The inspector finds the deletion of the 6 F criterion acceptable considering that the vendor requireaents of 101 and 96 F were met since samples were adequately taken with the plant at normal operating pressure and temperature (2235 psig, 557 F) without exceeding 86 F (Unit 1) and 82 F (Unit 2) sample coolant effluent temperature (2) Deficiencies AM, AQ, AS, AZ, BF, BT, BU and B These deficiencies were associated with several test steps whose purpose was to verify that adequate demineralizer flush water would purge the sampling panel. The deficiencies were written because the ficw rates obtained were less than the flow rates specified in the test steps. Actual measured flows ranged from 5 inches of water to 30 inches of water, whereas the test steps specified expected flow rates of 35 to 40 inches of water. The deficiencies were dispositioned by TRB/ PED based upon the rationale of Byron Unit 1 Retest R-252 which concluded that the demineralizer flush system was at a significantly lower system pressure (150 psig)

than the design pressure of the reactor coolant sampling module (2300 psig) or the demineralizer sampling module (1250 psig) and that the expccted range of the flush flow was based upon a design pressure of 1700 psig. These conclusions frcm the Byron Unit I retest were documented in a letter dated September 5,1984, from G. T. Klopp to R. Querio and in the test evaluation for Retest R-252 dated August 14, 198 Although both of the above conditions were resolved in September, 1984, with acceptable dispositions, they were not applied as lessons learned to Byron Unit 2 preoperational testing. Through discussions with the Byron staff, it was determined that the Graidwocd Unit 1 precperational test had become the " basic source document" because it had received offsite PED approval prior to the approval of the Byron Unit 2 preoperational tes This appears to be acceptable based on the F. A. Hornbeak instruction letter " Byron-2 Lessons Learned Program," paragraph 7, which states that "in a few instances we have used the Braidwood test procedure as the source document for our test procedure." However, it is not acceptable to approve a test that has not incorporated corrections from as such the above examples (paragraphs and(1) previous (2)) testing and constitute a violation of 10 CFR 50, Appendix B, Criterion XVI (455/86041-09(CRS)), which requires that conditions adverse to quality be identified and corrected in a manner that precludes repetitio b. During the review of Preoperational Test MS 51.60, " Main Steam Isolation Valves," the inspector noted that the pressure switches

which alarm to alert the operators of low nitrogen pressure in l

!

__

. ._ _ _ _ _ _ _ __ . _ _ _ _ _ _

t e .

the MSIV accumulators had to be recalibrated several time :

Discussions with the System Test Engineer (STE) revealed that these switches have historically exhibited problems with respect to staying within the calibration tolerances. These problems have been noted on Byron Unit I as well as at Braidwood. Review of the calibration records associated with the pressure switches for hydraulic as well as nitrogen cover gas pressures revealed that in some cases, the as found values were in excess of four times the allowed tolerance band. It was also noted that the

'

majority of the pressure switch alarm setpoints had drifted in the nonconservative direction. Discussions with men.bers of the

Instrument and Control Department revealed that the pressure switches in question were calibrated on a 208 week frequenc The licensee stated that they would review the effectiveness of

'

the current calibration frequency and, based on operational experience, would revise it as necessary. The inspector stated that from the observations made during the preoperational tests and subsequent retests, incraasing the calibration frequency was warranted until additional operational history on these switches was obtained. Determination of the adequacy of the calibration frequency of the pressure switches listed below is

,

considered to be an open item (455/86041-10(DRS)).

Pressure Setpoint Last As Alarm Switch Found Value Function 2PS-MS-127 4000 + 48 psi 3930 Hi pressure 2PS-MS-128 4000 + 48 psi 3910 Hi pressure 2PS-MS-129 4000 + 48 psi 3940 Hi pressure

2PS-MS-130 4000 + 48 psi 3945 Hi pressure 2PS-MS-135A 3100 + 30 psi 2920 Lo pressure 2PS-MS-135B 3100 + 30 psi 3040 Lo pressure

,

2PS-f15-136A 3100 + 30 psi 3035 Lo pressure

'

2PS-MS-136B 3100 + 30 psi 3015 Lo pressure 2PS-MS-137A 3100 + 30 psi 3070 Lo pressure 2PS-MS-137B 3100 + 30 psi 3005 Lo pressure 2PS-f15-138A 3100 + 30 psi 3020 Lo pressure 2PS-MS-138B 3100 + 30 psi 2990 Lo pressure At the time that the Byrcn Unit 2 license was issued, portions of

, eight preoperational tests, which are listed below, were incomplete.

'

(One of the eight, VA 84.61, was deferred in its entirety.)

AF 3.60, " Auxiliary Feedwater" i

I CV 18.67, " Chemical and Volume Control System - Integrated Hot

'

Functional (IHF)"

PS 61.60, " Primary Process Sampling" -

RY 69.67, " Reactor Coolant Pressurizer - IHF" SI 73.63, "ECCS Check Valve Operability ard Leakage"

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,.

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i l SX 76.60, " Essential Service Water" VA 84.61, " Auxiliary Building Ventilation" VP 93.61, " Primary Containment Ventilation - IHF" l

With the exception of the completion of AF 3.60, RY 69.67 and VP 93.61 which are being tracked via separate items as discussed in

Paragraphs 3d, 3a and 3e, respectively, the NRC will track the completion of the'preoperational tests via an open item (455/86041-11(DRS)).

-

One violation and two open items were identifie . P_reoperational_ Test _. Program _,Implen.enta_t_ ion Review (70302)

The inspectors reviewed the implementation of the test program contained in the Startup Manual to determine that document control, design change

control and administrative control were effective throughout the Byron '

Unit 2 preoperational test phas Document and Design Change Control

'

The inspectors reviewed several completed and approved

, preoperational test packages, including RC 63.60, " Integrated Hot j Functional," RC 63.67, " Reactor Coolant System," and CS 17.60, l

" Containment Spray," and verified that the design drawings used by

the startup test engineers were the current issues. Additionally,

! the inspectors verified that design drawings contained design changes (ECNs) by reviewing drawing M-129 Revision AE which 1'

contained design change ECN-32844 on the Containment Spray Syste Based on their review, the inspectors found that document control

was adequately implemented, Adniinistrative Control The inspectors verified that the administrative centrols of the Startup Manual were adequately implemented by direct interview of startup test engineers and supervisors. The inspectors verified that the individuals were familiar with the Startup Manual requirements for controlling corrective ar.d preventive maintenance and testing deficiencies. In addition the inspectors reviewed the training records of several test personnel and determined that their training adequately included administrative controls for testing and appropriate technical trainin ;

No violations or deviations were identifie . - . - - . - - - - --. --

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e < .

6. Startup Test Procedure Review (72564, 72566, 72570, 72572, 72576, 72578, 725EO,72582,72583)

The inspectors reviewed the following startup test procedures against the Final Safety Analysis Report (FSAR), Safety Evaluation Report, applicable Regulatory Guides and Standards, and portions of 10 CFR 50:

2.05.80, " Loss of Offsite Pcwer" 2.32.83, " Initial Criticality and Low Power Test Sequence" 2.45.828, "Incore Movable Detector and Thermocouple Mapping at Power

(Partial Core)"

- 2.45.82C, "Incore Movable Detector and Theraccouple Mapping at Pcwer (Quarter Core)"

2.47.80, " Isothermal Temperature Ccefficient" 2.47.81, " Power Coefficient Determination" 2.47.82A, " Thermal Power Measurement and Statepoint Data Collection

,

(Pre-Critical)"

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2.47.82C, " Thermal Power Measurecent and Statepoint Data Collection (For 0%, 30%, 50% or 90% Power Levels)"

2.47.620, " Thermal Power Measurement and Statepoint Data Collection (75% Power)"

.

! 2.47.82E, " Thermal Power Measurement ar.d Statepoint Data Collection (100% Power)"

2.52.86A, " Axial Flux Difference Calibration"

'

2.52.86B, " Axial Flux Difference Calibration" 2.52.88, "100% (Power) Plant Trip"

! 2.63.85, " Shutdown Outside Nain Control Room" 2.68.80, " Reactor Protection Trip Testing" 2.69.80, " Pressurizer Testing" I With respect to 2.47.80, the inspector noted the following:

(1) Steps 7.2.2 and 7.2.3 indicated that two strip charts would be used during the test, one of which would record reactivity and l

r.eutron flux and the other of which would record average l temperature and pressurizer level. Hcwever, Steps 7.3.12 and

7.3.13 ir.dicated that one strip chart wculd record reactivity l

l l

l

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< .- ,,

and average temperature and the other would record neutron flux and pressurizer level. During discussions between the inspector and the licensee, the licensee stated that Steps 7.3.12 and 7.3.13 were correct and that the procedure would be changed to correct Steps 7.2.2 and 7.2.3 via Test Change Request No. 1 in conjunction with the "30-day" procedure review which is required to be completed prior to, but within 30 days of, the test procedure executio (2) In their approval letter for 2.47.80, the Project Engineering Department identified an apparent contradiction between the 10 CFR 50.59 evaluation and the test procedure in that the 10 CFR 50.59 evaluation stated that "All rods are above their rod insertion limits," whereas Step 7.3.25 of the procedure states that Special Test Exception 3.10.3, which permits the rods to be below the rod insertion limits, may be in effect during the conduct of the test. During discussions between the inspector and the licensee, a representative of the licensee's Unit 2 startup crganization stated that the statement n.ade in the 10 CFR 50.59 evaluation was incorrect and that the 10 CFR 50.59 evaluation would be corrected via Test Change Request No. 1 in conjunction with the

"30-day" procedure revie The items discussed in Paragraphs (1) and (2) above are considered an open item (455/E6041-12(DRS)) pending licensee action to make the stated test procedure changes during the "30-day" review and subsequent NRC revie During the perfomance of Startup Test 2.47.81, " Power Coefficient Determination," it is important to maintain a constant reactor coolant system (RCS) boron concentration. Although the test procedure required that baron concentration measurements be taken at the beginning and end of the test, it did not require that these values be compared. The licensee's Project Engineering Department felt that since a precaution statement in the procedure requires that RCS boron concentration changes be avoided during the test, it was not necessary to specify the amount by which the boron concentration treasurements could change without adversely affecting the test results. NRC's review of the post test results of 2.47.81 to ensure that the licensee adequately addressed the irrpact of any boron concentration changes on the test results is considered an open item (455/86041-13(DRS)).

Two open items were identified which will require further action by the licersee and the NR . Electrical Train Separa_ti_no Issue Paragraph 2.a. of Braidwood Inspection Report No. 456/86052(DRS);

457/86040(DRS) identified a violation regarding failure to fully evaluate a deficiency related to safety injecticn pump instrurents (discharge pressure and flovi) being powered from the opposite electrical system (e.g. "A" train punip instruments powered by "B" train electrical system).

. . -.. .. .. - -

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i i

As indicated in the Braidwood Inspection Report the lack of complete independency of electrical trains existed not only for Braidwood, but for Byron Units 1 and 2 as well. The cover letter for the Braidwood

. Inspection Report requested that Corrwor. wealth Edison Company's (CECO)

i response to the violation encompass Byron Units 1 and Additional details are provided in the Braidwood Inspection Repor One violation was identified which was documented in a Braidwood Inspection Report as discussed above.

Diesel Generator Breaker Issue This issue concerned an apparent design deficiency in the diesel generator (CG) breaker. During the periodic testing of the DG with it paralleled to the bus, a LOCA signal coincident with Loss-of-Offsite j power will cause the DG breaker to trip in order to shed the loads on

! the bus, but the DG breaker will not automatically reclose on the undervol tage. This does not meet the intent of the FSAR and Regulatory Guide 1.108, position C.1.6.3. The licensee's position was that the

design is adequate because an operator would be stationed at the electrical panel in the train control room during testing and the DG procedure states

,

the actions that need to be performed in case of this event. Discussions have been held between CECO, Region III, and NRR. NRR has the lead in i terms of following this issue. NRC Region III will track this issue with an open item (454/86045-03(DRS)); 455/86041-14(DRS)) which will be closed when NRR considers this issue to be satisfactorily resolve One open item was identified which will require further action by the licensee and hRR to resolv . Independent Technical Review of Preoperationa_1_ Tests

Due to a concern raised by the NRC on October 8, 1986, concerning the i adequacy of the engineering evaluation associated with certain preoperational tests at Byron Unit 2, the licensee initiated a review of a sample of those tests to re-confirra the integrity of the original engineering evaluaticn. The review was performed by Westinghouse testing personnel wh'o were not involved in the original review of the tests used

in the sample. The review was n.onitored by the CECO Manager of Nuclear Safet The review was to assure that discrepancies, if any between accepted test results and the applicable acceptance criteria in the test were identifie In addition, the review compared the acceptance criteria in the test with

. the FSAR test abstract for inconsistencies. The review also included a i revicw to identify any differences between the Test Sunirary and FSAR test i abstract The licensee reviewed 10 out of the 28 tests that were in question.

,

These tests are as follows: ,

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.. . _ _ _ _ _ _ _ . _ _ . , . . _ _ _ _ . _ _ _ _ _ _ _ _ _ __ _ . _ _

e CV 18.61, "CVCS and RCP Seal Injection Logic" RC 63.62, " Reactor Coolant - B0P" CC 10.60, " Component Ccoling" fiDG 22.61, "2B Diesel Generator"

1. EF 26.61, "ECCS Full Flow"

1. Fil 32.62, " Fuel Handling Building Crane"

1. PS 61.60, " Primary Process Sarrpling"

1. RH 67.60, " Residual Heat Removal" SI 73.61, " Safety Injection Accumulators" SI 73.62, "SI Flow Balance" The inspectors reviewed a san.ple of these tests (as r.cted by a (f) to deterraine the adequacy of the independent review. Specific convents related to the procedures are addressed in Sections 3 and 4 of this inspecticn report and Sections 3 and 4 of Inspection Report N /E6043(DRS);455/86034(DRS).

No violations or deviations were identifie . Open Items Open itens are rnatters which have been discussed with the licensee, which will be reviewed by the inspector and which involve some action on the part of the NRC or licensee or both. Open iteras disclosed during the inspection are discussed in Paragraphs 3a, 3b, 3d, 3e, 4b, ac, 6a, 6b and . Unresolved _ Items Unresolved items are tratters abcut which more inferr..ation is required in order to ascertain whether they are acceptable items, open items, deviations, or violaticr.s. Unresolved items disclosed during the inspecticn are discussed in Paragraphs 2f and 3 . Exit Interview The inspectors met with licensee representatives denoted in Paragraph 1 at the conclusion of the inspection on November 14, 1986. The inspectors summarized the purpose and scope of the inspcction and the findir.gs. After discussions with the licensee, the inspectors have determined there is no proprietary data contained in this inspection repor .