IR 05000338/1980016
| ML19323H786 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 04/29/1980 |
| From: | Kellogg P, Kidd M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19323H782 | List: |
| References | |
| 50-338-80-16, 50-339-80-17, NUDOCS 8006160186 | |
| Download: ML19323H786 (13) | |
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- o, UNITED STATES 8 )n,-*j NUCLEAR REGULATORY COMMISSION
r REGION 11
101 MARIETTA sT N.W sulTE 3100
. ATL.ANTA, GEORGIA 30303 ..... Report Nos. 50-338/80-16 and 50-339/80-17 Licensee: Virginia Electric and Power Company P. O. Box 26666 Richmond, VA 23261 Facility Name: North Anna Units 1 and 2 Docket Nos. 50-338 and 50-339 License Nos. WF-4 and CPPR-78 ' InspectionatNorthAnna. shit ela Mineral, VA / Inspector: _/ -/ d,/o /li, * 9/' /[o 'B.' S. Kid .s ' d ector Date igned Approved by: _ / 7/c/ P.
q M Sect N -f, unNS Branch Date Signed . , ". SUMMARY Inspection on March 3 through April 11, 1980 Unit 1 Areas Inspected
This routine inspection by the resident ~ inspector involved six hours onsite in . the areas of licensee event reports and selected short term lessons learned (NUREG-0578) requirements.
t . Unit 1 Findings: No items of noncompliance or deviations were identified.
Unit 2 Areas Inspected i This routine inspection by the resident inspector involved 70 hours in the areas
of implementation of short term lessons learned (NUREG-0578), and THI-2 Action Plan (NP2EG-0660) requirements, reports per 10 CFR 50.5$(e) and 10 CFR 21, and plant readiness for fuel loading.
' Unit 2 Findings: . M ._. . '. 3-- . No items of noncompliance or deviations were identified. - ~ -(i.19Y- , .. _ J.u Y.. '*J~5.A.*-.*5**W.;***'"~~'*.~'..'".~. '.-*~+..N.**.~- ~ ~ ~ ~n~T '.'.N~ W ~ '* "~ T,' '.~ 5. 5.5 "*.* *.*:b. * * - ' ~ ~ ~ ~ ~ ~ ~"'Y - ' mw.- .. vs u.n-c.
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. . DETAILS 1.
Persons Contacted Licensee Employees t -I' R. F. Anderson, Acting Operating Supervisor H. W. Burruss, Associate Engineer
- W. R. Cartwright, Station Manager J. H. Horton, Chemistry Supervisor L. O. Goodrich, Mechanical Maintenance Supervisor C. C. Kelsey, Engineer i
47. D. Kellams, Superintendent Operations J. R. Harper, Instrument Supervisor D. M. Hopper, Health Physics Supervisor
- E. R. Smith, Superintendent - Technical Services D. L. Snodgrass, Assistant Instrument Supervisor D. E. Thomas, Electrical Maintenance Supervisor Other licensee employees contacted during this inspection included four oper-ators and three office personnel.
Other Organisations Stone mid Webster Engineering Corporation (S&W) M. C. Reynolds, Senior Advisory Engineer A. W. Rychaksly, Advisory Engineer . * Attend 6d exit iriterview 2.
Exit Interview , The inspection scope and findings were summarized on March 7,14, and 21 and April 3,1980, for those persons indicated in Paragraph I above.
, 3.
Licensee Action on Previous Inspection Findings Not inspected.
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Unresolved Itens.. a...f.... _ _, _.._. _. .y.
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... - -, Unresolved items were not identified during this inspection.
. 5.
Implementation of Short Term Lessons Learned (STLL)- Unit 2 NUREG-0578, "1MI-2 Lessons Learned Task Force Status Report and Short-Tern . Recommendations", of July 1979 defined certain requirements to be implemented , _.. ._ prior to operating license issuance (Category A) for plants such as North , - - - ? -~. - Anna Unit 2.
- These requirements were clarified by letters issued by NRC on ' October 30 and November 9,1979.
Virginia Electric and Power Company (VEPCO) . ~-.' responded to NUREG-0578, as clarified, by submittal of letters to NRC inclu- - ' ding serial r.o. 572 of July 12,1979; no. 681 of August 24,1979; no. 806 of . -_ e : pix- ,.. -... - - ' ' ... ..,. g.
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F . -2- . October 25, 1979; no. 906 of November 26,1979; no. 015 of January 10, 1980; no. 015A of Januar, 31,1980; no. 015C of February 8,1980; and no. 200 of , March 10,1980.
The resident inspector was requested to ascertain degree of implementation of those items which required modifications to existing equipment or instal-lation of new equipment. Lead review of all other items is the responsibility of the NRC Office of Nuclear Reactor Regulation (NRR), or a management inspec-tion team, whose findings were dccumented in IE Report 50-339/80-01. This team reviewed the STLL areas of shift technical advisor (Section 2.2.1.a), operating experience evaluation (2.2.1.b), onsite technical support center (2.2.2.b), and onsite operations support center (2.2.2.c).
STLL areas reviewed by the inspector were as follows: Power Operated Relief Valve (PORV) Redundant Motive Power - (2.1.1.3.2) a.
The licensee's evaluation of this requirement concluded that high pressure nitrogen should be utilized as a backup motive force for the pressurizer PORV's, which are also actuated by instrument air. The following documents were reviewed: (1) Engineering and Design Coordination Report (E&DCR) P-2673-2, as revised; (2) E&DCR P-2004-2 series; (3) Rework Control Forms (RCF) associated with the above E&DCR's; . (4) Instrument (TIP) and Mechanical (TAP) test procedures associated with the above RCF's ana E&DCR's.
Review of this documentation revealed that high pressure nitrogen had
been installed for both PORV's and that most required calibrations and checkouts had been completed, including valve stroke times (no flow).
However, as of March 21, 1980, further calibration checks were to be , made and valve stroke times under dynamic conditions were yet to be completed. Licensee personnel were informed that this function, as f yet part of the reactor coolant system (RCS) overpressure protection system, must be demonstrated operable prior to solid water operation following fuel loading (open item 339/80-17-01)y.
.. , The original RCS overpressure protection system, required by FSAR Supplementary Section 5.81, was previously installed and tested during hot functional testing of Unit 2.
Portions of that testing were witnessed by the inspector as documented in IE Report 50-339/79-01, . Details paragraph 9.c.
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Direct Indication of Pressurizer Valve Position - (2.1.3.a) " pr: ' _ . .... - - ~ Letter no. 015 of January 10, 1980 states that accoustical monitors , i would be provided on the pressurizer safety and relief valves prior to - startup of Unit 2.
The inspector verified physical installation of , ' I a :- . . _.. . u . _. _... - , j ' - -- . _.
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_ _. _ _ _ - _ _ _ _ _. . -3-the monitors and reviewed E&DCR P-2689J-2, involving installation of them. Also reviewed were test records in the form of instrument test procedures (TIP) and electrical test procedures (TEP). As of March 28, 1980, the monitors had been calibrated and tested by Babcock and Wilcox procedure C-010-00, " Installation and Checkout Procedure for Valve Monitoring Systems" as much as possible. Final noise spectrum recording and setting of alarms will be accomplished during post-full load testing. As of March 28, control room vertical board indicators had not been installed due to late delivery from the vendor. Licensee personnel were informed that these indicators must be installed and calibrated prior to initial criticality (open item 339/80-17-02).
c.
Core Subcooling Monitors - (2.1.3.b) . VEPCO's letter of January 15, 1980 committed to installation of a Westinghouse core subcooling monitor on Unit 2 prior to startup. This was accomplished by E&DCR's P-2707-2 through P-2707B-2. These documents were reviewed along with completed electrical and instrument test procedures. Final calibration and checkout of the monitors was per instrument calibration procedure ICP-2-SC-RC200, "Subcooling Monitor System - Channels A and B".
There were no further questions following review of the above documentation and inspection of equipment as installed.
d.
Auxiliary Feedwater Valve Alarms - (2.1.7.a) The auxiliary feedwater (AFW) systswas already designed for automatic initiation, but the licensee's evaluation determined that alarms should be added to the pump discharge valves to alert operators to malfunctions. E&DCR's P-2669-2 through P-2669D-2 implemented instal- ' lation of the necessary valve limit switches and alarm circuits.
Review of the E&DCR's and completed TIP's and TEP's revealed that alarm circuits had been installed and tested for all nine (9) AFW discharge valves. There were no further questions on this item.
D AW Flow Indication to Steam Ge erators - (2.1.7.b) e.
As noted in VEPCO's letter of January 10, 1980, power supplies for the k AFW flow indication circuits were swapped to safety grade sources.
ELCR's P-2674-2 and P-2674A provided instructions for swapping supplies
and indicators.
The E&DCR's, along with completed TEP's and ICP's were reviewed. Following inspection of the flow indicators there were no further questions on this item. -' -C f.
Improved Sampling Capability - (2.1.8) .. - ~ A description of an interim system to provide the capability of obtain-i. - ing a post-accident sample of reactor coolant was provided in VEPCO's - letter no. 015A. of,1/31/80.. This system, shared with Unit 2, was
installed by Unit 1 cesign change package 79-85. The design change - added provisions to the installed sampling system for collection of a sample, dilution of it under controlled conditions, and transporting ,_ , the sample in shielding to the hot lab for analysis. Testing results - l [., ...,.. ... ,._ - - - - --- --- -~ - - -. - - - - - ~ -
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__ _ _ _. . _. _ _ - - - _ - - _ _ _ _ - _ _ _ _ _ - - . -4- . documented in a memorandum from VEPC0 to Stone and Webster on January 23, 1980, revealed good results on dilution capability. Inspection of the newly installed system revealed it to be installed as prescribed by DC ,
79-85.
The inspector observed, however, that pressure indicator PI-SS-126, used in obtaining the sample was obscured by shielding such that a short individual would have difficulty in using it as required.
Station management stated that it would be raised to allow easier reviewing.
This will be reinspected at a later date (open item 339/80-17-03).
, g.
Increased Range Radiation Monitors - (2.1.8.b) Letter no. 015 of January 10, 1980 stated that high range monitors would be installed on the main steam safety valves, the process vent, and ventilation vents prior to Unit 2 startup. These were to have a
range of up to 10 R per hour. The monitors for the process vent (RM-GW 173), vent stack "A" (RM-VG 174), and vent stack "B" (RM-VG 175), shared wi*.a Unit 1, were installed by Unit I design change DC 79-66.
Review of this completed design change package and instrument calibra-tion procedures (ICP) revealed that these monitors had been in talled and calibrated by January 31, 1980, the commitment date for Unit 1.
Monitors RM-MS 270, 271, and 272 were instal.1.ed on the Unit 2 main ' steam safety headers via E&DCR's P-2706-2, P-2706A-2, OSD-10098-2.
These completed E&DCR's, along with associated instrument test proce-dures were reviewed.
Also reviewed were ICP's which accomplished final calibration of each unit on March 27, 1980. Based on review of the above documentation and inspection of the main steam monitors in + Unit 2, it appeare.d that this function had been cocrpleted as required.
As noted in Section II.F.1 of Supplement No.10 to NRR's Safety Evalua-tion Report for North Anna 1 and 2, interim procedures for quantifying i high level releases utilizing tuese new monitors have been evaluated by NRR and found acceptable.
6.
Near Term Operating License Requirer-nts - Unit 2
NUREG-0660, TMI Action Plan, specifies certain activities to be completed prior to fuel loading on plants such as Unit 2.
The resident inspector was
- j assigned responsibility for followup of some of those items. Findings are
[ presented in the following paragraphs. Task plan item numbers are referenced t with the activity headings.
y i N a.
Shift Manning - (I A.I.3) - g i Task I.A.I.3(a) requires a senior reactor operator (SRO) and a reactor operator (RO) in the control room at all times other than Mode 5 opera-tion. Station Administrative procedure (ADM) 1.0, " Station Organization f - - and Responsibility" was revised January 28, 1980 to ircorporate this ) requirement. Item I.A.1.3(b) places restrictions on use of overtime.
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ADM 3.0, " Hours of York", was revised March 12, 1980, to incorporate ] . the required restrictions. The inspector had no questions in these ~ i , areas.
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Vendor Review of Procedures - (I.C.7) Item I.C.7(b) requires Westinghouse review of Unit 2 low power test procedures. This had not beca accomplished as of March 28, 1980 and ' will be required prior to performance of the test (open item 339/80-17-04).
c.
Training During Low Power Testing - (I.G) This item requires development and approval of a training plan by which shift personnel will conduct " hands-on" training during the low power test program. As of March 28, 1980, a plan had been drafted, but had not been finalized and approved. This must be accomplished pricr to initiation of low power testing (open item 339/80-17-05).
d.
Shift Supervisor Administrative Duties - (I.A.1.2) , This item involved minimization of administrative duties of the shift ' supervisor. ADM 1.0, " Station Organization and Responsibility", was reviewed January 28, 1980 to delegate certain of these duties to other j parties. The inspector had no questions at that time.
Shift Relief and Turnover Procedures - (I.C.2) e.
This item required development of plant procedures for shift relief and turnover. These functions are defined in ADM 29.3, " Shift Conduct", dated December 12, 1979. A checklist for performance and documentation of the activity was approved December 31, 1979.
Paragraph 2.D.(5).g of the Unit 2 draft operating license will require that VEPCO establish a system to evaluate the effectiveness of the turn-over procedures. A North Anna Quality Control Inspection Report check-list has been developed for use in surveillance of this function. The inspector had no questions on either of the above items.
f.
Shift Personnel Responsibilities - (I.C.3) Section 2.2.la of NUREG-0578, as clarified required three actions in this area. Responses to those are as follows:
(1) A letter was issued to all nuclear station shift supervisors by the Vice President, Power Supply and. Production Opertions on December 26, 1979, confirming their responsibilities for safe operation of the plant. Plans are to reissue this annually as required.
(2) The letter referenced above, plus ADM 1.0 as revised on January 26, ~ , 1980, delineate the duties and responsibilities of the Shift Super-visors.
(3) &. ' This item required incorporation of the topic of responsibility for safe plant operation into the shift supervisor's training I ~ & , program. A review of the planned training program on March 21, d'
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f . -6- . 1980 revealed that this topic was not specifically listed.
Station management stated that the program would be revised to explicitly require the training (open item 339/80-17-06).
g.
Control Room Access - (I.C.4) This item required provision of procedures to limit access to the control room. Station.Aministrative procedures ADM 1.0 and ADM 6.0, " Control Room Access" wer upgraded January 28, 1980, and December 29, 1979, respectively to pro.ide more explicit instructions on limitir.g access by defining necessary/non-necessary personnel and total numbers permitted in the control room. There were no questions on this item.
h.
Auxiliary Feedwater Initiation and Indication - (II.E.1.2) The North Anna 1 and 2 auxiliary feedwater systems were initially designed to be automatically initiated, thus no change was required in that regard. Control grade flow indicators were installed as dis-cussed in paragraph 5.e of these Details.
i.
Inadequate Core Cooling Instru'ments - (II.F.2.(a) and (b) Subparagraph (a) required development of procedures for use of existing instrumentation in determining adequacy of core cooling. Review of these procedures by another NRC inspector was documented in IE report 50-338/80-10 and 50-339/80-08.
Subparagraph (b) required installation of a subcooling meter. This has been completed as discussed in paragraph.5.c of these Details.
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NRC Resident Inspector - (I.B.2.2) This condition requires placement of an NRC resident inspector at the site. This was accomplished in July 1978. In December 1979, a second resident inspector was assigned in a training status. This individual is expected to be certified as an inspector in the near future. The current senior resident inspector will be transferred to IE Headquarters in the near future and be replaced by another Region II inspector.
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Plant /NRC Telephone Lines , Direct dedicated telephones have been installeh'at the North Anna Units I and 2 control room, the NRC resident inspector's affice, and the site technical support center. A second network for health physics and envi-ronmental information has been installed with extensions in the control b room, health physics supervisor's office, technical support center, and the plant visitors center. A functional check of these phones was per-p ' formed after installation and periodic checks are made on a continuing - "
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Unit 2 Status - Readiness for Operation IE Inspection Report 50-338/80-11 and 50-339/80-09, Details paragraph 7 delineated activities to be completed and problems to be resolved prior to IE:II making a finding that Unit 2 would be ready for an operating license. A revised listing is given below, which reflects the addition of certain other items based on inspection efforts documented in this and other reports. The listing addresses functfons required to be completed prior to full loading by NUREG-0578, "THI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations" and SUREG-0660, "TMI-2 Action Plan". Items to be completed or otherwise resolved are categorized below by recommended completion milestone: a.
Items to be Resolved Prior to Fuel Loading (1) Completion of preoperaticnal testing and resolution of significant test deficiencies. As of March 24, 1980, test procedures 2-PO-30.4A, 2-PO-69A, 2-PO-69B, and 2-PO-70D had not been completed.
All of these tests had been completed by April 8,1980, and fol-lowing licensee personnel reviews, they were made available for NRC review on April 10, 1980. They were reviewed for conformance to NPSQAM sections 5,11, and 14 and FSAR section 14.1.1.
It appeared that the tests had been conducted as required, and deficiencies noted had been resolved or provisions made for tracking them for completion / resolution. No deficiencies which would impact fuel loading were identified.
, (2) Complete development and approval of operational procedures required by Technical Specifications and R.G. 1.33. As of March 24, 1980, i some eight (8) Surveillance Test procedures needed for early opera-ting stages had not been approved.
Unit 2 Technical Specifications require procedures as recommended by Revision 2 of R.G.1.33, whereas Unit 1 procedures were originally written to meet Revision 1.
In reviewing the Unit 2 procedural system, emphasis was placed on the differences in the two revisions.
As of March 24, it appeared that all procedures required had been t approved except for the eight (8) surveillance functions referred to above.
On April 10, 1980, the inspector completed a brief review of periodic test (PT) procedures *.and surveillance logs ! which had been revised or developed to cover those surveillance - functions. PT's and logs reviewed included 1-LOG-6D, 2-LOG-4A, i 2-PT-104.4, 2-PT-82.1A through 82.4A, 2-PT-82.1B through 82.4B and 2-PT-80.
The inspector had no further questions on Unit 2 . s ~ .,. n t operational procedure availability following this review.
, ,, . .... _ - .. p' ~ Previously identified open item 339/79-09-01 involved another "-- . . .. ! . inspector's observation that no mechanical maintenance procedures (MMP) had been identified for Unit 2.
During the current review, s
' it was confirard that the MMP's are common to Units 1 and 2.
, i' - f.. .., ~ ~ .. _ Licensee personnel were inforwed that this item was considered
closed.,
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- . _. --. ' . -8- . ' , . t (3) Complete implementation of industrial security program in accord-ance with the North Anna Modified Amended Security Plan.
Plant readiness to implement full security on Ur;c 2 was ascer-tained by other inspectors as documented in Ir; Report 50-339/ 80-16.
Actual implementation will be ver. tied as part of the routine inspection program.
(4) Resolve quality assurance program items involving instrument calibration control (339/79-11-13 and 339/79-11-24).
These items were inspected and closed in IE Report No. 50-339/ 80-14.
b.
Item to be Resolved Prior to Two Unit Operation in Mode 4 or Above Modify the electrical transfer buses to eliminate overloading during two unit operation (LER 79-57/01T-0 for Unit 1; Unit 2 open item 339/ 79-28-10).
This item was inspected and closed in Report No. 50-339/80-15.
Items to be Resolved Prior to Filling and Venting the Reactor Coolant c.
Coolant System After Fuel Loading (1) Completion of calibration and checkout of-the overpressure pro- , tection system including the high pressure nitrogen system for the power operated relief valves.
(See paragraph 5.a of these Details - open item 339/80-17-01.)
(2) Replacement of upper internals control rod guide tube support ,.
pins. Westinghouse informed VEPCO and the NRC (letter no. NS-TMA-2214 of March 13,1980) of a potential for broken guide tube support pins due to stress corrosion cracking. This matter was reported under 10 CFR 50.55(e) and 10 CFR 21 by VEPCO on March 13, , 1980.
Replacem nt of the support pins was in progress at the conclusion of the inspection period (open item 339/80-17-07), j . d.
Items to be Resolved Prior to Initial Criticality
(1) Verify that no mechanical interferences exiIt between incore guide , tubes or modified insulation supports and the reactor vessel lower head insulation due to thermal growth during precritical testing
(78-24-02).
t (2) Visual verification of hydraulic snubber operability at operating , temperature (79-26-02).
~ i ., . i;' (3) Verify adequate seismic stress analysis of safety related piping ' per IE Bulletin 79-07 (79-28-07).
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., ' (4) Verify adequacy of service water and component cooling water pipe , supports to resist thermal loadings over service water temperature . range (78-36-01).
, t (5) Determine if atypical weld material was used in the reactor vessel seam welds (IE Bulletin 78-12A).
(6) Verify that piping and support loads are within limits for systems where erroneous Velan check valve weights were used (79-21-03; also
see IE Bulletin 79-04).
(7) Resolve error found in DNBR analysis for single dropped control rod event (78-28-01).
(8) Resolution of potential degradation of electrical penetrations due to overcurrent conditions (79-46-01).
(9) Verify adequacy of emergency diesel generator fan cooling systems q over all ambient temperature ranges (79-44-02).
(10) Replacement of cracked disc for containment isolation valve MOV-2380 (79-54-04).
(11) Upgrade safeguards building exhaust system to seismic class I (79-54-06).
i (12) Verification of adequacy of piping and supports for those systems ' wherein incorrect seismic response spectrum curves were used in - stress analyses (79-56-02).
(13) Modification of circuitry to assure that engineered safety features equipment do not return to normal operating modes following reset of ,: containment depressurization actuation (CDA), safety injection (SI), or main steam isolation (MSI) signals (79-56-04 and 80-17-12; ree also IEB 80-06).
. , (14) Resolution of potential nonconservative feature of Westinghouse large break LOCA-ECCS evaluation model and its impact on peak
p clad temperatures (79-58-03).
(15) Resolution of nonconservatism involving the error analysis of the ' nuclear instrumentation negative rate trip for dropped control rods (79-58-04).
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(16) Resolution of nonconservatism in the method used to generate power ' ! I distributions and rod worths for dropped rod events (79-58-05).
l i ! -- - (17) Completion,of the seismic ' analysis of as-built safety-relating ! , piping system (IEB 79-14/80-09-01).
I l ' _e (18) Verification of seismic qualification of control panels for the
, control room air conditioner chillers (80-05-01).. - - ' l
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I . , -10- . (19) Rerouting of the steam generator blowdown monitor sensing lines which were found to be cross-connected (79-54-02).
(20) Completion of Westinghouse review of low power test procedures (near term 0.L. requirement I.C.7(b)/open item 80-17-04).
(21) Development of a training plan for shift personnel during conduct of low power tests (near term 0.L. requirement I.G/open item 80-17-05).
(22) Completion of installation and checkout of relief and safety valve accoustical monitoring instrumentation (Short Term Lessons Learned requirement 2.1.3a/open item 80-17-02).
, (23) Completion of installation and checkout of high range radiation monitors (Short Term Lessons Learned item 2.1.8b).
As of March 27, 1980, the date on which this IE list was finalized , for incorporation into the Operating License (Technical Specifica- " tions Section 7), calibration of certain of these monitors had not been completed. Later in the inspection period, as noted in para- { graph 5.g of these Details, the completed calibration procedures ' were reviewed. This item is considered closed.
(24) Completion of rework and retests of the Safeguards Building valve pit sump pumps (Unit 2 Master Deficiency list (MDL) item - 339/80-17-08).
, (25) Completion of retests of volume control tank level control valve LCV-2115A following limit switch replacement (MDL item - 339/80-17-09).
(26) Completionof seismic analysis of low head safety injection dis-charge lines for temperatures below 70'F and any necessary modi-fications resulting.
On March 11, 1980, VEPCO reported per 10 CFR 50.55(e) and 10 CFR 21 , that the LHSI discharge lines had not been stress analyzed for j temperatures below 70'F, This must be completed prior to initial i criticality (open item 339/80-17-10).
N . (27) Completion of closure gaps in the contain5ent recirculation sump grating where piping and piping supports pass through the grating.
(open item 339/80-17-11).
' e.
Iten.s to be Resolved Prior to Exceeding 25% Power J ~ ~ . _ - (1) Develop surveillance program for fiberglass spray pond piping and - supports (R.G. 1.72).
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_ _ _ .__ _ _ _ _ _ _ _ _ _ _ . -11- . (2) Complete environmental qualification testing and resolve any deficiencies identified for electrical components inside contain-ment (IE Circular 78-08/78-28-01 and IE Bulletin 79-01).
During discussions with the NRR Project Manager on March 24, 1980 the inspector was informed that followup of this area would be required by the NRR list of open items. By mutual agreement, this area will be dropped from this IE tracking list. (See Unit Techni-cal Specification 8.5) ' 8.
Licensee Event Reports - Unit 1 The following events or problems were reported as prompt (24 hours) LER's per Unit 1 Technical Specifications during the report period: Piping Not Analyzed for Possible Fluid Temperatures (LER 80-34) a.
On March 11,1980, VEPC0 reported that Stone and Webster (S&W) had j discovered that the discharge lines for the low head safety injection ' pumps had not been analyzed for temperatures below 70'F.
In that the refueling water storage tank contents are maintained between 40'F and 50*F, temperatures below 70*F are possible during the ECCS initial injection phase.
Reanalysis and resultant modifications were in i progress at the conclusion of the inspection. This will be reviewed in more detail at a later date (open item 338/80-16-01). As noted in paragraph 7.d.(26) of these Details, this problem also applies to Unit 2.
b.
Guide Tube Support Pins (LER 80-35) On March 13, 1980, VEPCO reported that Westinghouse had discovered that the upper internals control rod guide tubes might be susceptible to stress corrosion cracking. Evaluation of support pins by Westing-house at other facilities will determine corrective actions needed on Unit 1.
This matter will be reviewed further (open item 338/80-16-02).
' As noted in paragraph 7.c.(2) of these Details, this problem also applies to Unit 2.
c.
Safety Injection Actuation (LER 80-37) 5p On April 3,1980, during conduct of a main steam trip valve (MSTV) partial closure test, MSTV "A" closed resulting in a safety injection (SI) and reactor trip. Upon receipt of the SI signal, emergency diesel generator 1H automatically started, but tripped on overspeed. This will be reviewed further (open item 338/80-16-03).
m* -- During recovery,e operations personnel noted that the MSTV's reopened ' when the main steam isolation signal was reset. This is inconsistent , - with FSAR Supplementary Section 7.4 in that two operator actions are - required to return a safety system to a non-safety mode (open item , 338/80-16-04). The MSTV reset problem also applies to Unit 2 and was - . .e e mm,a% -a ap.te e e'*
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. -12-reported under 10 CFR 50.55(e) on April 11, 1980. This problem is similar to those previously reported and tracked as open item 339/ 79-56-04, which are discussed in paragraph 7.d.(13) of these Details.
This open item (339/80-17-12) will require resolution by initial criticality also.
9.
Auxiliary Feedwater Motor Fans - Unit 2 During extended runs of the Unit 2 auxiliary feedwater (AFW) pumps as required by NRC, it was found that the inboard motor bearings overheated.
Investigation revealed that the inboard and outboard fans had been reversed on both motors for the Unit 2 AFW motor driven pumps. The fans were rein-stalled properly and the motors retested. Similar motors are used elsewhere in Units 1 and 2, but the problem was unqiue to the Unit 2 AFW motors. This - will be reviewed in more detail at a later date (open item 339/80-17-13).
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