ML19320D296

Submits Summary of Concerns Stemming from Review IE 1977-79 Insps & Lers.Preoperational Test Program Was Conducted Inadequately.Mgt Controls Over Power Ascension Test Program Is Inadequate

ML19320D296
Person / Time
Site:	Crane, Davis Besse
Issue date:	07/18/1980
From:	Creswell J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:	Ahearne J NRC COMMISSION (OCM)
References
TASK-TF, TASK-TMR NUDOCS 8007210217
Download: ML19320D296 (13)

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SUMMARY

OF CONCERNS

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1.

The Preoperational Test Program was conducted inadequately. This is evidenced by a large number of equipment malfunctions discovered during the Powar Ascension Test Program. Large number of outstanding f?:Ei:7 items 'added as condition of License also are indicative of unresolved Preoperational Test items.

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Management controls over Power Ascension Test Program has been

.1 inadequate. There is evidence of " Jury-rigging" of systems to fj}$~

55iiff conduct tests. There is evidence that testing has been delayed to l?E---]

._....q allow electrical generation. This action results in operation at EE significant power levels with untested systems even in light of

=== =6 events which show evidence of inadequate system performance (e.g.

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November 29, 1977 event).

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3.

There is evidence that the reactor design provides significantly less protection than other PWR reactor designs.

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There have been numerous significant operator errors., The inspector notes that these errors are not being reduced in frequency.

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There is evidence that significant design defects exists in the electrical distribution system.

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There are serious questions about conformance with several General iiij Q@

Design Criteria.

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NEO has conducted management meetings with the licensee identifying e5:5M V:::

=f many repeat concerns. These concerr ; are not being adequately

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A significant number of safety related FCR's (Facility Change

..53 3 Requests) remain outstanding. This item coupled with operator

..;;;.,...l..j crrors mentioned in Item 4 increases both the risks and consequences,

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of accidents at the facility.

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There is evidence that when defects are identified in safety related m=ii==

(ii:f systems the defects are not analyzed and corrected in a timely

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g EXAMPLES IN SUPPORT OF CONCERNS 1.

References:

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Paragraph 8, Report 50-346/77-06

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Appendix A, to Report 50-346/78-19, Page 3(a) f"."5""""

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Paragraph 4, Report 50-346/78-17, Item H March 7, 1979 transmittal

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Memo, Keppler to Mosely dated August 14, 1979

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Reference a refers to how the SFAS testing was performed.

- r=1 Reference b refers ta the inadequacy of the SFAS testing.

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Reference c refers to the conduct of electrical testing during

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preoperational testing and power ascension testing.

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personnel.

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References:

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a.

IE Report 50-346/78-06, Appendix A

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IE Report 50-346/78-17, Appendix A and Paragraph 13 i~"fiff

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IE Report.50-346/78-30, Paragraph 9 c.

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IE Report 50-346/79-04, Appendix A (T4.c+ Q,4kh M*W*I g.....].

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Commentary:

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Reference a refers to management control over rod drop testing.

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The test deficiency is still unresolved.

Reference b refers to management control over determination of

?ikbfb worst case core peaking factors and corporate management

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overview of the Startup Test Program.

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Reference e refers to use of supplemental air supply during 222' testing.

Reference d is to management controls over testing.

Some items

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are repeat of Noncompliance issued in 50-346/78-06.

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Some critical testing such as Natural Circulation Test, Loss

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of Offsite Power Test, Shutdown Outside Control Room Test and Load Rejection Test have only recently been completed after

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approximately one and one-half years after initial criticality.

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References:

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IE Report 50-346/78-30, Paragraph 13 E

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Rancho Seco event, March 20, 1978

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c.

Licensee report on November 29, 1977 event (Item G, March 7, 1979

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transmittal) f '-

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Additional Safety Evaluation of Transient Resulting from Inability

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of Operator to Control Steam Generator Level at 35 Inches

"EE E (Docketed --Serial No. 475, December 22, 1978)

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Reference a refers to no direct abtomatic triaiassociated

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with loss of heat sink such as the Westinghouse low steam i::

generator level reactor trip.

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....q Refsrence b refers to the loss of power to non nuclear-I"'*" *1 h:: : rq instrumentation which can result in severe thermal transients lFE={

and extreme difficulty in controlling the plant.

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=..::q Reference e and d refer to loss of pressurizer level indication

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,,...q and possible voiding of the pressurizer dur.ng anticipated r, ",] operational transients. g:......j ... q .a

===d gi =4 There is a concern that large positive moderator temperature .::- j J... (:iik: coefficients produce difficultly in coccrolling the pir.t. r This item will be addressed further in Item 8.

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There is a concern regarding the reduncancy and diversity of the g:dll.:."=..! g.... t= auxiliary feed water system. Usually there are combinations of {' .... ] st.eam and electric driven auxiliary feed pumps. The Davis-Besse a =L.?ii. facility has two steam driven auxiliary feed pumps. There are

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..b".'.....~ indications that during some of the events steam pressure dropped to levels which affected the pumps operability. .==; ' E:.: 5- ~~:. u....... ..?.7.. ~ ::::: E:: lE:Iss: - if ~ E.-

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s Eii "~! I:: (]i: 4. R_eferences: 2955E= V= W* a. Vendor Report ot. September 24, 1977 event (Item i', March 7, 1979 [. ... 4 _. =.~f Transmittal) - "'" ' 1 C" "":q b. Vendor Report on April 29, 1978 transient (Item P, this transmittal) I,...-.; c. Reportable Occurrence 78-066 (in PDR) ...glg d. Reportable Occurrence 78-067 -s .._.....{ Commentary: "9EE

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l ':== :.s ..j Reference a refers to operator shutting oft the Emergency Core '"~~""". .4 Cooling System during the LOCA. l Reference b and c refers to poor operator performance particularly ' ~ (h in view of withdrawing control rods when they should have been "=" inserted. iisiliiss Reference d is to operator repeating errors. I>me

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References:

EEE 555.5 l ~ "' a. Memo, Streeter to R. W. Woodruff dated June 9, 1978 (AITS F30385H2) iE& b. Daily Staff Notes, October 30, 1978.- I"=== ['19i"=u c. Licensee Report on November 29, 1977 event (Item G of March, 1979 E"' transmittal).

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e. IE Report 50-346/78-17, Paragraphs 4, 5, and 12. (=.;...... ~~" f. IE Report 50-346/78-30, Paragraphs 3, 9, and 10. Ed $ETE ..q [.t'~;,'~'g.. Commentary: EEEE5 %=: y_y Commentary on these items will be delayed until results of ~~~~ EE1 recent events regarding loss of offsite power are reviewed Eg sj further. =*l .=======q ..) =::: =3 9.....". =..1 4

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o p::..._... 6. (a) Comunents on GDC 11 p=eg .=- Criterion 11 - Reactor inherent protection. The reactor core and

== '"":.i, associated coolant systems shall be designed so that in the power .====4 I-operating range the net effect of the prompt inherent nuclear feedback characteristics tends to compensate for a rapid increase g ;.[.)

===. in reactivity. r~g;; ;.g Commentary: It does not appear to the inspector that the Davis Besse core J."":..._ - design conforms to this minimum design requirement. A moderator temperature coefficient of + 0.7 x 10-4 O % AK/k F was measured during the Cycle IB startup. In addition, the inspector has a concern that the reactor was operated above 95% power with a positive i temperature coefficient (see 50-346/79-04). Some effects noted from operation with the positive coefficient are noted below. 8/2/78 In preparation for 40% reactor physics testing, the six

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second rod. insertion step for differential rod worth 1r.;3;

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measutzement was attempted. The rod movement resulted in a

==; EU Reactor Coolant System upset. The positive temperature coefficient caused feedwater control of Tave to be unstable. 1-.=== A divergent oscillation in feedwater lead to overfeeding oi ~r ; the steam generators, and resulted in an RPS low pr'ssure j 7. trip. [:=53 . :== s [;l. ;;._ [

Reference:

Supplement 3 of the Davis Besse Unit 1 Initial ...... ~. fl5== Startup Report Dated February 8, 1979. }ry_ =---.y

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O 2...... 8/31/78 Tech. Spec. 3.1.1.4 regarding minimum temperature for

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s{ criticality exceeded due to positive moderator temperature b.. 1 coefficient. 5 E.0.5: E

Reference:

R.O. 78-088

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TP ST.0B 4 9 ..y E: The inspector notes that the Zion (Westinghouse PWR) Technical !Ezs= Es=-I."4 Specifications require: _,.._ q .......a 3.2.1.C. Unit Startup ..._.. ~. =.-...

==d 1. Immediately prior to startup, the reactor coolant temperature .==1

=====i shall be shown to be greater than the temperature above which =J;g the moderator temperature coefficient is always negative and ..... -~ greater than 5000F, except during low power physics tests.

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g=g.._.g. p. fjs:khh and the bases state: 1:.1 During the early part of a fuel cycle, the moderator _,j = temperature coefficient may be slightly positive at coolant i;; "J g.. ~ '5! temperatures below the power operating range. (1)(2) The f moderator coefficient at low temperatures will be most positive at the beginning of life of the fuel cycle, when '"E== the boron concentrations in the coolant is the greatest. Later =El in the cycle, the boron concentrations in the coolant will be ~l1--i:1 lower and the moderator coefficient will be either less positive or will be . _. -] 4 negative., At all times, the moderator coefficient is negative

g. g in the power operating range. (1)(2) The maximum temperature

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.J ({f of life of any fuel cycle with all control rods withdrawn, is is fjf determined during the lower power physics tests for that cycle. [;:,.. ~ . := =?... ; ~ ~.:- = f

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o .== (b) Comments on GDC 12 = [5 .).. ~::2 Criterion 12 - Suppression of reactor power oscillations. The reactor core and associated coolant, control, control, [._?.!?5 and protection system shall be designed to assure that f= power oscillations which can result in conditions ~ p-=::==: exceeding specified acceptable fuel design limits h=gs... p=... are not possible or can be reliably and readily riEEEi detected and suppressed. ~~~

Reference:

I.E. Report'50-346/78-06 Paragraph 4 855 ... ~:: ... i 'The reference refers to power oscillations observed at Davis Besse -1. The Oconee facility has experienced power oscillations of "..E up to 7% power peak to peak. The inspector can find no definitive }~![

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553 statement regarding the safety implications of these oscillations. y,

== : w. (c) Criterion 13 - Instrumentation and Control. Instrumentation. shall be provided to monitor variables and systems over their .==; anticipated ranges for normal operation, for anticipated operational ~ 50 occurrences, and for the accident conditions as appropriate to assure EE !L - = adequate safety, including those variables and systems that can Eh!==- affect the fission process, the integrity of the reactor core, the _ _.y= = reactor coolant pressure boundary, and the containment and its =q 5 f:] associated systems. Appropriate controls shall be provided to .......3 2 maintain these variables and systems within prescribed operating =fd EE5$d ranges. 2.f.;..l. = = =i Ep!E E 'i .21:. ~~. = =.

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Reference:

I.E. Report 50-346/78-06, Paragraph 2 i=~ "J:1:

== [E.. During the November 29, 1977 event there was a loss of pressurizer F '={ r.

==1 level indication for five minutes. In addition there is concern E:I about monitoring makeup flow and T cold during a thermal transient y.,,.ll. q ~ a of this type. ~

===j iEni'E{ Furcher information on GDC 17 and GDC 33 will be furnished at a E I... M ~ later date when more information becomes available. _j 7.

References:

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Licensee minutes of October 1977, Management Meeting (Item E, ~' March 7, 1979 transmittal)

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==--.3 b. NRC Report on August 1978, Management Meeting (Item J, March 7, ~ ~, c=. [" D 1979 transmittal). f ""

== c. Licensee Minutes of August 1978 Management Meeting (Item K, is iE March 7, 1979 transmittal) E=5 Commentary: The inspector feels that a review of the references ..t :. =f address the concern but if further information is needed the lp:l.s 5:. E inspector will gladly discuss the matter. ...) ll.1 ,.3) lnd EE1 ._===; 55h

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Reference:

1.E. Report 50-346/79-05 Paragraph 6.d y.:. (g! E===.]. Commentary: i:...-... t: ;====

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Supplementary informat%n obtained from another inspector regarding ]" y-the status of the FCK's (Facility Change Requests) indicates the

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following: = = " ' o =:' ...2===. a. 516 FCR's are not ready for implementation ~~TZ b. 145 are in the implementation stage

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c. 245 are in the followup stage -- Z~ d. 162 are closed = = -

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Additional information is forthcoming as a result of my request for an investigation.

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References:

h.... a. LER 77-11 "=f b. LER 77-53 .:====. c. LER 77-61

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d. LER 77-17 e. LER 77-80 t:- f. LER 77-83 g. LER 77-110

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LER 77-113 -I. LER 77-116 bb

j. LER 78-05

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Commentary: !="-- (=2.... b=:: =: f:===== The auxiliary feed pumps have had extensive difficulties in speed 7.. ...... 4

=== =d control. In July and August, 1977, re;i ated speed control relay

-- i failures rendered the auxiliary feed pu..ps inoperable. On August t

10, 1977, a design modificatica was implemented which added a ~-"~ ~~ t..._.. 3 second set of identical speed relays in parallel to reduce the t E-current carried by each relay. This did not totally eliminate the [ . t.... ..q speed control failures and in January, 1978,.the relays of the speed t. E:=uc-:.i circuit were replaced with relays of a larger current carrying ~~~~~""1 capacity. (Exerpt from startup test report)

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