ML19345E390

Safety Evaluation Supporting Proposed Change 7 Revising Tech Specs

ML19345E390
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	07/09/1965
From:	Boyd R US ATOMIC ENERGY COMMISSION (AEC)
To:
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ML19345E389	List: ML19345E388 ML19345E390
References
NUDOCS 8101150836
Download: ML19345E390 (11)

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SAFETY EVALUATION BY THE RESEARCH AND POWER REACTOR SAFETY BRANCH l

DIVISION OF REACTOR LICENSING CONSUMERS POWER COMPANY PROPOSED CHANCE NO. 7 DOCKET No. 50-155 Introduction The Consumers Power Company has requested by application dated March 26, 1964, supplemented by letter dated May 21, 1964, and 1WX dated July 13, 1964 that certain changes be made to the Technical Specifications of License No. DPR-6 for the Big Rock Point re actor.

Forty-nine items were contained in the proposal, all but nine of which are designated Proposed Change No. 7.

Two of the proposed changes relating to natural circulatic flow testing, items 13 and 46, were reviewed separately by the Commission ud authorized as Change No. 1, dated Jure 15, 1964.

Three others relating to high power density tests, items 47, 48, and 49 were re-viewed and authorized as Change No. 2, dated July 31, 1964. The request for Item No. 12 was withdrawn by Consumers as indicated by its letter dated May 21, 1964.

In addition, items 6 and 38 are not included in this evaluation as Consumers has indicated that approval of these items need not be granted at this time.

Item 25 was superseded by the request authorized in Chnnee No. 3.

Evaluation of the remaining proposed changes is presented herain.

Discussion of Editorial Changes Most of the proposed changes to the Technien1 Snecifications in no way affect the safe operation of the facility. This category consists of changes in wording of the Technical Specifications to more clearly express limitations originally imposed by the Commission, and items involving the principal calculated characteristics of the reactor which were appropriate for the initial operation of the facility but are considered obsolete now that the operat.ing characteristics have been identified. Accordingly, we believe that the uhnical Specifications relating to these items which do not af fect safety may be changed as follows:

(1)

Item 1 - Delete the present Section 1.2, " DEFINITIONS", and substitute the following:

"1.2 DEFINITIONS j

Various provisions of these Technical Specifications set forth limitations and restrictions which depend upon medes of operation.

The following modes of operation are defined to :larify the intent of such provisions, and are not the same as, nor should they be confused with, the positions of the mode selector switch described in Section 6.1.3.

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1.2.1 Power Operation - is any operation other than shutdown or cold shutdown with the reactor vessel closure bolted in place.

1.2.2 Core Alteration - is any completed planned sequence of movements or core components resulting in either a net change in the con-figuration of the reactor core or a net gain in core reac-tivity.

1.2.3 Refueling Operation - is any operation with any of the reactor sessel closures open during which a core alteration, or other operation which might increase core reactivity, is in progress.

1.2.4 Major Refu;j; ig - is any refueling operation with the head of f during w"f eh four or more fuel bundles are added, exchanged or reposit A-oed in the reactor core.

1.2.5 ph,utdown - is any reactor condition meeting the following re-quirements :

(a) All or all but one of the control rods are fully inserted in the reactor core; and (b) Primary system coolant water temperature is less than 212*F.

1.2.6 Cold Shutdown - is a reactor condition involving no fuel in the core, or a reactor condition meeting the follosine require-ments:

(a) All of the control rods are fully inserted in the core and withdrawal prevented by means of the keylock selector switch, the key to which is in the possession of the Shift Supervisor; and (b) The reactor coolant system is at atmospheric pressure; and (c) The core shutdown reactivity control margin requirement has been verified in the manner set forth in Section 5.2.2(b)."

l (2)

Item 2 - Change the first sentence of Section 3.0, " REACTOR CONTAINMENT",

to read as follows:

" Reactor containment shall consist of an externally insulated spherical steel vessel, hereinafter referred to as the containment sphe re."

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(3)

Item 4 - Change Section 3.5.2(c) to read as follows:

"The proper operation of the automatic valves and associated controls of the spray system shall normally be functionally tested during each major refueling shutdown, but not Jtss fre-quently than once every 12 months."

(4)

Item 5 - Change Section 3.6, " CONTAINMENT REQUIREMENTS", to read as follows:

" Containment sphere integrity shall be maintained during power operation, refueling operation, shutdown and cold shutdown conditions except as specified by a syster of procedures and controls to be established for occasions uhen containment must be breached during cold shutdown."

(5)

Item 7 - In Section 4.1.3, change the last sentence of the first para-graph to read as follows:

"A removable shield plug of a thickness 4 feet, 6-1/2 inches, consisting of 4 feet, 4 inches of concrete and 2-1/2 inches of lead, shall close the opening above the top of the reactor."

(6)

Item 8 - In Section 4.2.l(a), change the second paragraph to read as follows:

"The reactor safety system and related circuits are fed from four 120-volt a-c buses. Each of two buses is supplied from a dif ferent 480-volt system through its own motor-generator set.

Each motor-generator is equipped with a flywheel to sustain operation during momentary power syst disturbances.

The third bus is supplied from the 125-volt d-c system through a static inverter which supplies power to Neutron Monitoring Channel No. 3.

The fourth bus is supplied from the 125-volt d-c system through a motor-generator set which supplies power ta the control rod position indicating system. The 125-volt d-c battery system also furnishes power for other critical services including:

Liquid Poison System Controls Motor-Operated Automatic Containment Sphere Isolation Valves Containment Sphere Ventilation System Isolation Valves Emergency Condenser Drain Valves Safety System Annunciators Emergency Lighting Swit chgear" l

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l (7) Item 9 - In the third paragraph of Section 4.2.10, change the second sentence to read as follows:

"During turbine start-up and shutdown, and for short periods of time during normal operation, operation on speed control shall be permitted.

During such operation, the turbine load limiter shall be set to limit turbine output to correspond to the planned reactor output."

(8)

Item 10 - In Section 5.1.4, change the second item in the left-hand

column, "Available Quantity of Solution, Gallons," to read " Poison Tank Capacity, Callons."

(9)

Item 11 - (a) In Section 5.1.4, change the fourth item " Initial Injection Gas Pressure, Psia 2080", to read as follows:

" Initial Injection Cas Pressure, Psia Condition Pres sure Head off 500 Head On - Relief Valve Settings 1250 1470 1500 1800 1750 2080" (b) Delete Section 5.2 in its entirety and redesignate Section 5.3 as Section 5.2 (10)

Item 16 - In Section 5.3.2(b), change the third paragraph to read as follows:

"During power operation, if reactivity and control rod motion data indicate a possible loss of poison from a control rod, the reactor shall be shut down and, if any corrective action

'is necessary, shall remain in shutdown condition until such corrective action has been taken."

(11) Item 17 - In Section 5.3.2(d), insert the words "to be utilized" between the words " rod" and "shall" in the first line of the subpara-graph entitled " Routine Coupling Integrity Checks."

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(12) Item 18 - Change Section 5.3.2(e) to read as follows:

" Control Rod Exercising During Sustained Power Operation -

At least once each day during sustained power eperation, the cperator shall exercise,'ach control rod that is not at that time fully inserted in cus wore."

(13) Item 19 - Change the first paragraph of Section 5.3.4 to read as follows:

"During initial loading of a core, the moderator temperature and void coef ficients shal3 be measured."

(14) Item 20 - Change (a) and (b) of Section 5.3.4 and eliminate paragraph (c) to read as follows:

"5.2. 4 Reactivity Coe f ficient s a) The power coefficient shall meet the following requirement at all times during the reactivity life of the core: The control rod movement re-quired to produce a given power change with the reactor pressure constant and the reactor water at saturation temperature with this pressure shall always infer a negative power coefficient.

b) The moderator. temperature coefficient, over the reactivity life of the core, (inferred from critical control rod position) during uniform heating of the core shall be limited such that the potential maximum /\\ K gg/Keff added by heating the e

moderator is always less than one dollar."

(15) Item 21 - Change Section 5.3.5(a) to read as follows:

"All rods

• shall be fully inserted during the reactivity addition.

In most valualbe rod completely withdrawn, the procedur$ ou/K [ned in Section order to verify the core shutdown margin of 0.3% ZL K with the tI 5.2.2(b) shall be utilized before and after any core alteration.which may result in a net gain in core reactivity. During the reactivity addition, i

suberiticality checks, consisting of withdrawal of one control rod in the vicinity of the component diange, shall be made at the intervals indicated in Section 5.2.5(b)."

• " Removal of a control rod from the core by means other than normal control rod drive movement shall require that the four fuel bundles surrounding the control rod first be unloaded and that the control rod be properly reinstalled and checked out prior to reinsertion of the fuel bundles."

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(16)

Item 22 - In the first sentence of Section 6.1.2.1, substitute the words "five decades abovg source level, without moving detectors (approximately 10 to 10-0% of rated power)" for the words

" rated power."

(16) Itcm 23 - In Section 6.1.2.2 change the first and last sentences to read as follows:

" Channels 4 and 5 provide logarithmic neutron flux icvel and period information from approximately 10-5% to rated power for the 84 bundle core."

"The detectors shall be gamma compensated ion chambers with a design sensitivity of at least 2.2 x 10-14 amp e res /nv."

(17) Item 24 - Change the first and last sentence of Section 6.1.2.3 to read as follows:

" Channels 1, 2 and 3 shall provide linear neutron flux level information from approximately 10-7% to 125% rated power for the 84 fuel bundle core."

"The dete ctors shall be gamma compensateg ion chambers with a design setsitivity of at least 2.2 x 10-ampe res /nv. The amplifier output shall be connected to the reactor safety system.'

(18)

Item 26 - Change Section 6.1.2.5, "NEITIRON MONITORING RANGE SWITCH", to read as follows:

"The range switches on the three power level instruments contain resistor-capacitor feedback circuit combinations which can be switched to provide nine decades of ove.-lapping power level indication. The unit also contains interlock contacts used in conjunction with the upscale-downscale trip units of the pico-amme t e rs. "

(19)

Item 27 - In Section 6.1.3 change the second sentence of the introductory paragraph to read as follows:

"A key-lock reactor mode switch shall be provided, having

' Shutdown', ' Refuel', ' Bypass Dump Tank' and 'Run' positionu."

Also, under the tabulation of functions, delete the position

' Start-up' and its trip function "None (e)'. Add the reference

"(e)" to the trip function for "Run."

(20)

Item 30 -In Section 6.1.5(d), chanRe the second sentence to read as follows. j "For reactor operation above approximately 5% of rated power, the logarithmic neutron flux level information and period scram protection are not required (see Section 6.1.2)."

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1 (21)

Item 31 - Change Section 6.1.5(h) to read as follows:

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" Minimum nuclear instrtraentation in operation during shutdown operation shall be the same as that required for refueling operation, except that only one start-up range monitor shall be required."

(22)

Item 39 - In Section 7.0, "0I'ERATING PROCEDURES", change the introcuctory paragraph to read as follows:

"This section describes those basic operating principles ano procedural safeguards which have a potential effect on safety.

Operating principles and procedures are presented for normal and emergency operation of the plant, for Phase II testing within the Research and Development Program, and for operational testing of the nuclear safeguards systems of the plant."

(23)

Item 40 - In the first line of Section 7.2.3, delete the words " Review of."

(24)

Item 41 - In the second sentence of Section 7.3.2(i), delete the phrase "and the reactor reaches rated pressure."

(25) Item 42 - Change Section 7.4(a) to read as follows:

" Detailed written procedures shall be available prior to each refueling operation."

Safety Considerations The remaining items, relating to changes in component operating dharacteristics, operating procedures, or tecting frequencies have significance from a safety standpoint. These items and an evaluation of the safety considerations are presented below.

Item 3.

In Section 3.4.3(c), change the closing time for " Main Steam Drain (MO 7065)" from "7.5" to "60" seconds.

Evaluation : Presently the Technical Specifications require a 60 second closing time for the main steam isolation valve and a 7.5 second closing time for main steam drain valve. Experience has shown that the closing speed on the drain valve from full-open to full-closed exceeds 7.5 seconds. It has therefore been necessary for Consumers to limit the stroke of the valva in order to meet the 7.5 second limit. Since this valve is normally closed during operation at power, and since the line is only 1-1/2" pipe size, Consumers believes and we agree that operation of the valve at the same speed as the main steam isolation valve will not significantly increase any potential hazard.

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Item 14.

Change the first sentence of the second paragraph of Section 5.3.2(a) to read as follows:

"The following tests shall be performed during each major resueling shutdown and at least once every six months during periods of power operation."

Item 15.

Change the last paragraph of Section 5.3.2(a) to read as follows:

" Insertion of each drive over its entire stroke with reduced hydraulic system pressure to determine that drive friction is normal."

Evaluation: Consumer requests that the control rod drive testing frequency be increased from three month intervals to six month intervals. The present frequer.cy of control rod drive testing has been required since 1962. The test data accumulated since that time indicate a slow deterioration of drive operating characteristics, probably due to component wear, but have not indicated any incipient failures. Based upon the statistics thus far accumulated it appears that any significant trend in data 'could be developed with a lower frequency of testing, without increasing potential hazard.

As further experience is accumulated with operation of the drive mechanisms during service life, other tests and dif ferent fre-quencies of testing may become appropriate. We believe that, in view of the testing experience to date, and the operating history of drive performance at Big Rock Point, it is reasonable at this time to change the frequency of required testing to six month intervals.

Item 15, although related to Item 14, only corrects the spelling of " stroke".

Item 28.

Change Section 6.1.5(a) to read as follows:

"Except as otherwise provided in these Technical Specifications, the reactor safety system shall be operable during power operation as indicated in Section 6.1.2.

This system shall be functionally tested during each major refueling shutdown, but not less frequently than once every 12 monthe, and in addition shall be tested not less fre-quently than once a month using the switches provided to simulate sensor trips."

Item 29.

Change Section 6.1.5(b) to read as follows:

"The core spray system and emergency condenser control initiation sensors shs11 be functionally tested not less frequently than once every 12 monens."

l Item 32.

Substitute the following for the last sentence.in Section 6.2.2

" Permissive circuits shall be functionally tested not less fre-quently than once every 12 months. However, the refueling inter-locks will be functionally tested prior to each major refueling."

Evaluation: The changes requested by items 28, 29 and 32 reflect the initial intent of the testing frequency for the safety system and core spray and emergency condenser controls.

It was assumed that the yearly frequency would be limiting; however, due to frequent

" refueling" shutdowns for conduct of the R&D program, the -

refueling shutdown frequency proved limiting. To correct the above sections of the Technical Specifications to reflect the intent, the changes in these items were proposed. We believe the testing frequencies, as -indicated, are acceptable.

Item 33.

In the parenthetical phrases in Sections 6.3.2(a) and (b), delete

", and monorail crane."

Evaluation : Experience has shown that movement of fuel into or out of the core via the monorail crane is neither necessary nor desirable.

Consumers indicates that plant administrative procedures prohibit its use for this purpose.

In our opinion, the prop'osed change is acceptable.

Item 34.

In the last sentence of the second paragraph of Section 6.4.1(a),

substitute "10-4 to 10" f or "10-3 to 100", and change the last paragraph to read as follows:

"A trip circuit in the air ejector off-gas monitor shall have an alarm which shall annunciate in the control room. The air ejector off-gas monitor trip circuit shall also initiate action of a time-delay switch, which in. turn shall trip the off-gas shutof f valve closed after a pre-selected delay adjustable up to 15 minutes.

(Off-gas average holdup time is about 30 minutes.) Alarm and trip settings shall be as specified in Section 6.4.3(a)."

In the fourteenth line of Section 6.4.1(b), substitute "10-4 to 10" for the "10-3 to 100."

Evaluation: Experience has shown that it is impractical to extend the range of the off-gas and stack-gas monitors to 100 curies per second. The proposed upper limit of 10 curies per second is quite adequate and still well above the operating limit of I curie per second. The ur9er limits of the alarm settings have been adjusted downward m.cordingly. The reference to alarm settings has been deleted from Section 6.4.l(a), and the last paragraph reworded accordingly, since these settings are specified in Section 6.4.3(a) under

" Operating Requirements."

. I item 35.

In the eleventh line of Section 6.4.3(a) and in the eighth line of Section 6.4.3(b), substitute "5 curies" for "10 curies."

Evaluation : The maximum permissible alarm setting should be reduced from

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10 curies /second to 5 curies /second so as to be well within the

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proposed upper limit of 10 curies /second on the off-gas and i

stack-gas instruments.

Item 36.

In Section 6.4.3(e), substitute the following for tht; last two sentences in the second paragraph:

"The films at each station shall be replaced and analyzed at least monthly." Also, change the last two sentences of the third paragraph to read as follows:

"One film at each station shall be replaced and analyzed at least every two weeks. The second film shall be replaced and analyzed monthly."

Evaluation: Experience with the three-month films has shown that latent image f adir.c. occurs.

It is, therefore, necessary to process and replace both films at least monthly. This requirement is more restrictive than previously and is, accordingly, acceptable.

Item 31.

In the last sentence of Section 6.4.3(f), substitute "once every three months" for " monthly."

Evaluation: Experience with this instrumentation shows very little need for monthly calibration.

Consumers' indicates, and we agree, that quarterly calibrations will provide sufficient accuracy and will eliminate needless radiation exposure of calibration personnel.

Item 43.

Cnange the second paragraph of Section 7.4(b), to read as follows:

" Fuel shall be replaced according to the following sequence:

(1)

Removal of selected bundles from core end transfer to spent fuel storage.

(ii)

Reshuf fling of remaining bundles in core as desired.

(iii)

Insertion of new bundles in vacant positions as desired. Shutdown margin verifications and suberiticality checks shall be made as required by Section 5. 2.5."

Evaluation: The refcaling procedure as originally written covered only the normal operations and did not take into account special refueling operations during the R&D period, the special operations fer fuel inspection, etc.

The proposed changes will allow more latitude for fuel mo-es and result in a more efficient operation. " Asse mb ly" has been changed to " Bundle" for sake of uniformity.

Item 44.

Change the second paragraph of Section 7.4(c), to[ read as follows:

"No edditional instrumentation need be placed within the core lattice if the out-of-core instrumentation produces a significant response to the subcriticality check in the' region to be altered. If this criterion cannot be met, a low-level neutron detector, measuring neutron flux, shall be located near the region to be altered."

Evaluation: This change places the requirenent for use of in-core instrumenta-tion during refueling on a functional basis and, we believe, will provide adequate neutron instrumentation during refueling opern-tions.

Item 45.

Change Section 7.5.7, to read as follows:

"It shall be permissible to remove a control rod drive from the core when the reactor is in the shutdown xenon free condition and the moda selector switch is locked on the " Shutdown" or " Bypass Dump Tank" position while the drive is removed. The core shutdown margin of 0.3% ggK,gf/K,f f with the strongest rod out of the core shall have been met prior to the drive removal; and, in addition, the equipment shall be properly tagged. The control rod drive that was recoved shall without delay be replaced by a spare drive or the original drive shall be reinstalled."

Evaluation: This change clarifies the drive removal procedure and allows drive removal under the newly defined " Shutdown" condition. The existing Section 7.5.7 is incompatible with the definition of " Cold Shutdown", in that the latter requires all rods to be inserted.

The control blade must be withdrawn to provide a water seal on the drive nozzle, and then is delatched before the hydraulic uystem of the drive is isolated. When the mode switch is in the " Shutdown" of " Bypass Dump Tank" position, a second control blade cannot be withdrawn, assuring that the shutdown margin is maintained during the drive replacement.

Conclusion On the basis of the foregoing considerations, we have concluded that Proposed Change No. 7 does not present significant hazards considerations not described or implicit in the hazards summary report and that there is reasonable assurance that the health and safety of the public will not be endangered.

Acc7rdingly, we believe the Technical Specifications of License No. DPR-6 may be revised as indicated herein.-

Original signed by:

Roger S. Boyd Roger S. Boyd, Chief Research & Power Reactor Safety Branch Division of Peactor Licensing JUL 91965 Date: