ML19340A258

Forwards Responses to Violations Noted in Insp Rept 50-269/74-04 & Informs That Info in Rept Nonproprietary

ML19340A258
Person / Time
Site:	Oconee
Issue date:	06/28/1974
From:	Thies A DUKE POWER CO.
To:	Moseley N NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
Shared Package
ML19340A255	List: ML19340A254 ML19340A258
References
NUDOCS 8001310578
Download: ML19340A258 (13)

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June 28, 1974 Mr. Norman;C. Mo'seley,JDirector-

.-Directorate of Regu atory. Operations l

U.1S. Atomic Energy Commission-Region II - Suite 818 230 Peachtree Street, Northwest Atlanta, Georgia' 30303 E

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.Re: RO:II:TNE 50-269/74.3

Dear Mr. Moseley:

Please find attached our responses.to Items I.A.l.a and I.A.3.a and b 50-269/74-4.

contained in RO Inspection Report i

i d in.

_ Duke Power; Company does.not consider any information conta ne i-50-269/74-4 to be proprietary.

I R0 Inspection ~ Report Very truly yours,

'A.:C.'Thics'

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o DUKE POWER COMPANY OCONEE UNIT 1 RESPONSE TO AEC/R0 REPORT NO. 50-269/74-4

.I.A.l.a Asymmetric Rod Monitoring System' 10CFR50.59 requires that proposed changes to the facility involving an un-reviewed safety question shall not be. carried out unless authorized by the Atomic Energy Commission.

Contrary-to the above, during the period from Ja'nuary 18, 1974 to April 26,

1974,; Unit 1 was operated with' as many as six asymmetric rod monitors for -

individual control rods turned off.

RESPONSE' In a February 28, 1974 letter to Mr. Angelo Giambusso, Deputy Director iar.

Reactor Projects, Duke Power Company reported problems with the control rod drive absolute position indication.(APl) system at Oconee Nuc1 car Station.

As identified in that report, most of the problems with the API have con-sisted of fluctuating or erratic indications.

The effect of the erratic indications-is to give an asymmetric rod signal to the integrated control

. system (ICS) which initiates a runback in power as long as the indicated fault exists provided the ICS is operated in the automatic mode.

In order to avoid spurious runback due to erratic position indication signals, station personnel switched out of service the signal from the asymmetric Records

, rod-monitoring bistabic in the control rod drive system (CRD).

indicate that six of these bistables were placed in the inoperative position.

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-Switching off the bistables had-the following effect on the operation of the unit:

1.

For those bistables switched off a runback in power would not have been

-initiated if those. rods were -asymmetric. However, in many cases, if a bypassed rod were dropped, runback would be initiated by other rods

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in the= group due to.the change in group' average.

-2.

Rod withdrahallinhibit was bypassed on the particular rods for which bistables were switched off.

3..

The statalarm window indicating " asymmetric ' rod" would have been in-operable for those rods which had the. bistable switched off. However, position indication on the -control rod ~ drive PI panel would be available Lto the operator as well as an asymmetric alarm light on the control rod

-drive station.

This light indicates a seven-inch asymmetric rod.

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should-be:noted when the. control. rod drive station is in' manual, the reactor runbacki and rod ~ withdrawal inhibit features of the control rod-r Ldrive;and: integrated control systcas are blocked.

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f a dropped control rod, assuming no ICS or control roi drive

.The effect o system action, has been analyzed and reported in BAW-1387, "Oconce 1 Fuel Densification Report," January, 1973.

Referring to Page 33, "The sod drop nralysis results in a decrease in power initially after which the power it has been shown previously that neither the returns to 100 percent.

withdrawal nor the drop of a single control element will cause perturbation of the flux shape suf ficient to exceed design conditions at 112 percent.

this _ occurrence still does not ' present any thermal problems."

Therefore, AEC/ Directorate of Licensing (DOL) in Supplement 3 to the Oconce Nuclear Station Unit 1 Safety Evaluation issued July 10, 1973 stated in Section 3.3 that the results of the control rod drop incident have been reviewed by the staff and this transient, "taking into account'the effects of fuel densifi-cation concludes that they would not result in a reduction of core thermal i.e., a DNB, less than 1.3."

Consequently, although the words of

margin, Section 14.1.2.7 of the Final Safety Analysis Report have not been revised af ter the analysis was completed for fuel densification, the ef fect of a dropped rod with no credit being taken for the rod withdrawal inhibi-or asymmetric rod runback f eatures of the control rod drive and integre:.ed control systems had been analyzed by B&W and confirmed by AEC/ DOL in 1973.

Also, Scction 7.2.3s3 of the FSAR states the following:

Failure of the ICS does not diminish the safety of the reactor.

None of the functions provided by the ICS are required for The reactor reactor protection or for actuation of the ESPS.

protection criteria', used in the analysis of the acc~idents presented in Section 14, can be met irrespective of ICS action.

In 1by of 1974, B&W was requested by Duke Power Company to reconfirm by analysis that core protection limits are not exceeded assuming no control system action. The results of that analycis are as follows.

The maximum increase in peaking for a dropped control rod is 22.3 percent Startup test with an incore detector quadrant power tia.t of 14.9 percent.

analysis for Oconee 1 of the worst-case dropped rod observed an incore The detector tilt cf 15 percent and consistent peaking relationships.

analysis indicates that if a dropped rod occurred during normal operation The the center fuel line melt and DNBR criteria would not be exceeded.

following is a list of assumptions for the analysis:

1.

The dropped rod is the worst case 2.

Initial power level is 102 percent 3.

Central fuel melt limit is 20.1 kW/ft Fuel densification is included 4.

Power peaking is at the highest level for normal equilibrium operation l

5.

during life (BOL)

There is no initial quadrant tilt 6.

Af ter the dropped rod, the tilt increases to.15 percent (peaking 7.

increases by 22.4 percent)

With the preceding assumptions, the initial heat rate is 15.0 kW/f t.

After the rod drop, the' linear heat rate would increase to 18.4 kW/f t assuning no

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If power control reactivity compensation and a return to 102 percent power.

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.remains constant at 102 percent and the control rods compensate for the reactivity decrease, linear heat rate increases to 18.0 kU/ft.

The minimum margin occurs in the case without reactivity compensation and is 9.2 per-cent.

The loss in DNBR peaking margin for the tilt at 15 percent is 17.8 p e rc ent. For the same conditions as the } receding, the minimum DNBR peaking margin is 19.5 percent at 102 per ant power leaving an ex #ess margin of 1.4 percent.

The above analysis by B&W assumes no action by either the control systems or by the operator.

Ilowever, Technical specification 2.5.2.2e states the following:

If a control rod in the regulating or safety rod groups is declared inoperable per 4.7.1.2, power shall be reduced to 60 percent of the thermal power allowable for the reactor coolant pump combination.

This, in effect, requires the operator to manually runback the pl' ant to 60 percent power assuming no control system action.

Appropriate action by the operator assuces additional conservatism with regard to the above analysis.

The violation as issued alleged that an unreviewed safety question was involved. From 10CFR50.59, the definition of an unreviewed safety question is as follows:

A proposed change, test or experiment shall be deemed to invcive an unreviewed safety question:

(1) if the probability of occurrence or consequences of an accident or a malfunction of equipment important to safety previously evaluated in the' safety analysis report may be increased; or (2) if a possibility for an accident or malfunction of a different type than that evaluated previously in the safety analysis report may be created; or (3) if the margin of safety as defined in the bases of any technical specification is reduced.

In view of the above analysis concerning a dropped control rod without control rod drive or integrated control systea action, Duke does not con-sider that the bypassing of the control rod drive asymmetric rod bistable monitors to be an unreviewed safety question.

Duke, however, is concerned that appropriate station management personnel were not adv'eed 'and did not have an opportunity to review the bypassing of the bistables.

Mr. J. E. Smith, Oconee Nuclear Station Superintendent, in a letter dated June 14, 1974 to all supervisors and engineers stated that the bypassing "of thece functions was performed without written pro-cedures and without proper review." Smith further writes, "while we do not consider the bypassing of these functions to be an unreviewed safety question worthy of a Category I violation, it does point out the potential l

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? hat;a potisibility exintsiin other 'arcas for ' the bypassing of important funct lens which'.could (result in 'a. compromise of saf ety or - the result. in t

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- In the future, such'use of bypass

negating,.of important functions.

switches is not.to"be permitted without proper review and a writte'n approved procedure."

-While Duke does not consider the action performed to be an unreviewed safety question,- it is-realized L that the description in -Section 14.1.2.7

.of the. Final Safety-Analysis Report may-not lead one to the proper con-

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'clusion concerning the ' required action of, the control rod drive ^ and inte-specifically,'that core protection criteria are

. grated control: systems:

There-met without asymmetric rod. runback or control ' rod drive inhibit.

fore,'a. revision to'Section 14.1.2.7 has been forwarded to AEC/ DOL'and is-

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attached _for information.

Duke is concerned that this matter was not -identified to station manage-ment.as n' violation during the exit interview by-the RO inspectors on Nor was Duke advised that RO considered this to be a April-26, 1974..

50-269/74-4 dated.May 28, violation. prior to receipt of Inspection Report 1974._ We believe.th'at appropriate identification of apparent. violations-at the earliest possible date will_ provide sufficient opportunity for dis--

i We believe

- cussion. prior to written issuance of the apparent violation.

its

' that a full discussion of this incident would not have resultcd :ht Although Duke believes that

.being identified as a Category I violation.

proper procedural steps were not taken at the station, we respectfully request. that the Category'I nature of this violation be rescinded.

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s 1.'A. 3. n Specific Conductance and pil Monitoring As'noted'in the details section of the inspection report, there was one day during the period from December 1,1973 to April 18,1974 (and it has been subsequently verified from April 18, 1974 through June 1, 1974) when pH and specific conductance monitoring uas not properly done.

It has not been determined why this sample was not taken.

It is considered that appropriate action has already been taken to prevent recurrence of this violation. A status board has been prepared which gives the last date on which the' sample was taken and the date on which the next sample is due.

In addition, additional emphasis has been given by the station chemist to the regular gathering of these samples for analysis or the re-porting of basin status if samples cannot be taken.

Since monitoring has been done regularly each day since December 22, 1973, no further corrective action is anticipated.

I.A.3.b Waste Collection Basin Effluent pH The proposed revision to Technical Specification 1.2B is under review by the Directotate of Licensing.

Furthermore, work is in progress to enlarge the wastewater handling and collection facility at Oconee.

Full compliance vill require satisfactory completion of both of these items.

O 1

i 14.1.N.7

'StucN0uti ' Stuck ;In, or Dropped-in Control. Rod Accident -

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14.1.2.7.1-Identification. of Cause s

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'In the event'that a control rod cannot be moved, localized power. peaking'and-suberitical marginLmunt be considered.

lf a control rod is ' dropped into the core while operating, a rapid decrease L

in-neutron power would occur, accompanied -by a decrease in core average coolant temperature.

In addition, the power dist'ribution might be dist6rted-due to the.new control rod pattern.

In the presence of a-distorted power distribution, the return -to full power-might11 cad to localized power densities

- and heat fluxes in excess of design limitations.

14.1.2.7.2 Protective Basis Adequate hot suberitical-margin is provided by requiring a-suberiticality of

- 1 percent ak/k with-the control rod of greatest worth fully withdrawn f rom the

-The nuclear-analysis reported in 3.2.2 demonstrates that this' criterion

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can be satisfied.

This criterion has been analyzed in terms of the' minimum l

. tripped rod worth available in the loss-of-coolant-flou, startup, rod with-

'drawal, and steam-line-failure accidents.

In all cases, the available rod worth is_ sufficient to provide margins below any damage threshold.

l For protective purposes a dropped control rod.is defined as the deviation of a control rod from its group reference position by more than a maximum of 9 inches.

This definition then covers both the action of dropping a rod and 3

sticking a rod while moving a group.

The action taken by the ICS is:

All rod-out motio'n is inhibited.

a.

b.

The steam generator load demand is run back to 60 percent of rated load at 5%/s.

The details of these actions arc described in Section 7.2.2 and 7.2.3.

Although tliesc '1CS actions are ~ available-to mitigate the consequences of the accident, they are not required' functions for safe plant operation as the results of the accident analysis demonstrate.

l Tae criterion'f'or-plant protection during this transient is that the DNB ratio

-will' not be less than.l.3 and the system pressure will not exceed ' code limits.

14.1.2.7.3 Method of: Analysis

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TThe transient response to a dropped contro1 rod has been analyzed using a detailed B5U digital model.

This' program includes fuel. pin, point kinetics, pressurizer, and loop models, including'the steam generators.

The reactor.is: assumed;to be operating at 100 percent of rated-power when the Tcontrol' ro'd. is : dropped.- In. order to achieve-the most adverse res'ponse the

-most negative values:of moderator coefficient [3.0 x 10-4 (6k/k)/F] and-g

.DopplerLcoefficien't ~[-1.3 x 10-5 ( Ak/k) /F} occurring at end-of-core. lif e ~ were.

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uaed. :The. maximum r' wortha expectcd to occur duri-full operation were used to examinc the etfects of ICS protective action.

Thesa rod worths

. correspond to operation at full power without xenon (0.46% Ak/k) and with xenon (0.36% Ak/k).

It was assumed that the eteam generator load dcmand was reduced linearly to 60 percent at 5%/s.

The effects of a dropped rod without ICS action were also examined for a very conservative rod worth of 0.65% Ak/k.

The rod was assumed to drop to 2/3 insertion in 1.4 seconds.

14.1.2.7.2 Results of Analysis The results of the analysis with ICS action are presented in Figures 14-18 and 14-19.

Figure 14-18 shows the response to a 0.46% Ak/k dropped rod.

The neutron power decreases rapidly to about 55 percent of rated power.

This causes rapid decreases in the core moderator temperature and fuel temperature.

These temperature decreases over-coupensate for the worth of the control rod, and the power rises until the reduced steam generator demand begins to increase the inlet temperature and decreases the power.

The thermal power levels out briefly at about 78 percent of its initial value but soon begins to decrease in response to the decreased steam generator demand.

The pressurizer pressure swing is about i 60 psi before returning to equilibrium.

Figure 14-19 shows the results of the 0.36% Ak/k rod drop.

The initial neutron power decrease is slightly less in this case, resulting in the thermal power leveling off at 83 percent, a slightly higher value than in the 0.46% Ak/k case.

The pressurizer pressure peaks at a higher value due to this higher thermal power.

Figure 14-19a shows the results of a 0.65% Ak/k dropped rol analysis con-servatively bcsed on no ICS action and operation at higher than rated power level of 2772 nwt.

The neutron power decreases causing a rapid decrease in both the core moderator temperature and the fuel temperature.

These tem-perature decreases overcompensate for the worth of the control rod, and the neutron power rises slightly above the initial neutron power level.

The neutron power then decreases to below the initial power level and eventually, levels out at the initial power level.

The thermal power response is similar to the neutron power; however, the thermal power level never exceeds the initial powe,. value.

Both the core moderator temperature and pressurizer pressure decrease during the transient and level out at a value lower than the initial value.

Several cases have been run for rod drops at beginning of life conditions.

These transients yield new power levels that are lower t,han the end of life conditions and may result in reactor trip.

These are therefore not included in this discussion because they represent less severe conditions.

14.1.2.7.5 Conclusions Control rod malfunctions are accommodated by the core' design without ICS action. Since the most' severe case analyzed for the dropped rod does not result in reactor trip nor does the thermal power exceed its initial value, core and reactor coolant system boundary protection'is assured.

Additional protection for the dropped rod accident is provided through the ICS which detects a dropped rod and inhibits out-motion of the control rods.

The ICS 14-14 Entire Page Revised Rev 34 6/28/74

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is ;denigned' to' run,3 9c'k the ; steam generator ! oad deu. sid-upon receiving the :

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~ dropped rod signal f rom the rod: drop detection circuitry.

thehnal. power-will assume la -lower value that' matches the ' load dernand and will

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l. provide : additional l margin t'oward not exceeding any design limit..

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' -Loss of E3cetric Power 14.1.2.8~

14.1.2.8.1.

Identification'of Cause e

Each unit. is designed to dithstand the effects of a loss of electric load or 8.2.3..Two~ types

-. electric power. ' Duergency' power systems are described in of power losses are considered:

a.- A loss of: load condition, caused by separation of the unit from the x

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.b...A hypothetical condition which results in 'a complete loss of..all system and station-power.

The reactor. protection criteria for those conditions are that fuel damage will -

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not occur from an excessive power-to-flow ratio nor will'the reactor coolant :

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systera pressure exceed design prassure.

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