U-601650, Requests Changes to License NPF-62,correcting Typos, Refs to Fsar,Testing to Be Done During Initial Plant Startup & Incorporating Revised Operating Power/Flow Map for Two Reactor Recirculation Loop Operation

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Requests Changes to License NPF-62,correcting Typos, Refs to Fsar,Testing to Be Done During Initial Plant Startup & Incorporating Revised Operating Power/Flow Map for Two Reactor Recirculation Loop Operation
ML20043H502
Person / Time
Site: Clinton Constellation icon.png
Issue date: 06/19/1990
From: Spangenberg F
ILLINOIS POWER CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20043H503 List:
References
L30-90(06-19)LP, L30-90(6-19)LP, U-601650, NUDOCS 9006260032
Download: ML20043H502 (10)
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U 601650-1 o

L3 0(06 19) LP CLINTON POWER STATION, P.O. HOX 678. CLINTON. ItLINOIS 61727 0678. TELEPHONE'(2171935 8881 -

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. j June'19.1990 i

10CFR50.36,

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Docket No. 50 461 Nuclear Regulatory Commission Document Control' Desk.

Washington, D.C.

205551 j

Subject:

Clinton Power Station Technical Soccification Bases-

Dear Sir:

The purpose of this letter is to request changes to. Illinois Power 3-Company's Clinton Power Station Technical Specification Bases. -The l

Bases contained in the Clinton Power Station Technical' Specificattor.s

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summarize the reasons for the Specifications in Sections 2.0, 3.0 and 4.0, but in accordance with 10CFR50.36, are not part of the Techni.nl Specifications. -Therefore, thes'e Bases changes are not being sub91cted as an application for amendment of. Operating License NPF 62. since the requirements of 10CFR50.90 are not applicable to such changen.

The purpose or nature of these changes is as follows: ' t) correct typographical errors and references to the Final Safety. Analysis Report j

(FSAR) and testing to be done during initial plant:startup, 2) 1 incorporate a revised operating power / flow map for two roactor V

recirculation loop operation,.3) delete information that is no longer applicable to the current = ACTION statements: associated with. reactor coolant specific activity, 4) updato guidance on the.uso'of,the.-

_ i containment purge-and exhaust lines to agree with recently docketed

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j correspondence, 5) clarify the basis for-drywell internal. pressure limits, 6) provide additional guidance on OPERABILITY of.tho. reactor core isolation cooling. system relative to the ACTION statomonts of related Technical Specifications,' 7) provide additional guidance concerning entry into and compliance with the ACTION-. statements

-associated with the Technical Specifications for diesel. generators, and

8) clarify the Bases for station battery specific gravity. limits to support the use of accoptable replacement battery cells having a j

different nominal specific gravity, j

A discussion of each of these changes and revised (marked up)'

' Bases pages are provided in Attachment 1.

Sincerely yours, F. A. S ngen'crg, II h

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bi"* "I"E ""d 8"I"'y 9006260032 900619 PDR ADOCK 0500046,1 g

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5 DAS/ cam Attachment cc:

NRC Clinton Licensing Project Manager

-NRC Resident Office Illinois Department:of Nuclear Safety

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.Attech:2nt 1 to U.601650 Page:1 of 38 As identified in the cover letter,'the specific Bases changes'have been categorized into eight separate categories based-on the-nature and purpose of the specific change. Table 1 (on page 9 of this Attachment) identifies the specific Bases section, table or figure affected, the associated Bases page number (s), and the associated category number which discusses the specific change.

Each of the' eight categories of:

changes is described below.

1.

Correct Editorial Errors and References to FSAR and Initial Plant Starruo Testine These changes are proposed to correct typographical errors that have been identified in the current Bases for certain Specifications.

In addition, references: to the Clinton Power Station (CPS) Final Safety Analysis Report.(FSAR) have been replaced with the references to the corresponding location in the CPS Updated Safety Analysis Report'(USAR) since the USAR is'the i

current-version of the Safety Analysis 1 Report in accordance with 10CFR50.71(e). These= changes are editorial-in nature and therefore do not result in a reduction in the level of safety

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maintained by the Technical Specifications and their Bases.

The Bases for Technical Specification 3/4.2.3, Minimum Critical Power Ratio (MCPR), state that an-evaluation of the MCPR at 25% of rated thermal power with minimum recirculation-pump speed will be performed during initial startup testing of-the. plant. -The purpose of this evaluation is to demonstrate that a large MCPR margin exists below 25% of rated thermal power so that future MCPR' evaluation below this power level is unnecessary.

This MCPR evaluation was performed during startup tests STP-19-1 and STP-19 2.

These tests demonstrated a'MCPR of 3.495 at 18.5% power and 27.2% core flow and 3.344 at 21.6% powerfand 33% core flow-The operating limit MCPR for these plant conditions was l'.711and 1.77, respectively. These tests-demonstrated that a large'MCPR i

margin does exist below 25% of rated thermal power.

Therefore, i

the Bases for Technical Specification 3/4.2.3 have been revised to reflect that this confirmatory testing has been completed.

2.

incoroorate Revised Power / Plow Mao for Two Reactor Recirculation Loon Oncration I

Bases Figure B 3/4.2.3-1, " Reactor Operating Map for Two Recirculation Loop Operation", was previously revised by Amendment 18 to the CPS Operating License. Amendment 18 was issued in response to Illinois Power Company's amendment request dated j

September 6, 1988 (IP letter U-601239) to support the first refueling of CPS with new fuel types. This request also supported subsequent operation in the Maximum Extended Operating Domain (MEOD) and operation with reduced feedwater temperatures.

Figure B 3/4.2,3-1 was revised in Amendment 18 to show the boundary of operation allowed by MEOD. However, the figure provided in Amendment 18 was based on a generic BWR/6.

Figure B 3/4.2.3-1 has now been revised to reflect the CPS-specific characteristics of i

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c' Att:chment 1 4

di to U 601650 Page 2 of 38 the 105 percent rod _line.

In addition, the 60 percent and'40 percent rod lines have been reincorporated'intorthe, figure.

(These two rod' lines were previously deleted in Amendment 18.)

The figure has also been rotated 90 degrees to' facilitate easier use by plant personnel.

3.

Delete Information That is No Longer Anolicabla

'The BWR/6 Standard Technical Specifications (STS)_and early drafts of the CPS Technical Specifications contained additional ACTION requirements for limiting plant operation with' reactor coolant specific activity levels exceeding the limits. With specific activity greater than 0.2 microcuries per_ gram' DOSE' EQUIVALENT I-131 but less:than 4.0 microcuries per gram DOSE EQUIVALENT-I 131 for greater than 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> in any consecutive 12 month period, the plant was required to be shut dnwn. Additionally, if this time exceeded 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> in any consecutive 6 month period, a Special-Report was_ required to'be submitted, Purther a Special Report l

was required to be submitted every.92 days following.a THERMAL POWER change of more than 15 percent in one hour,'an offgas.

3 activity level increase of~more'than-10,000 microcuries por second, or an offgas activity level increase of moreLthan 15:

percent.

These cumulative time limits and additional' reporting.,

i requirements were subsequently removed from the CPS < Technical' Specifications in accordance with NRC'Ceneric1 Letter'85 19 prior.

to issuance of the Low Power Operating License;. However, the discussion of these items in-the Bases was not deleted.

l Therefore, the Bases information associated with these additional ACTIONS has been deleted to be consistent with the: current ACTIONS of-. Technical Specification 3.4.5.

4.

Uodate Guidance on Use of Containment Purge and Exhaust Lines As identified in Supplement 5 to the NRC stafffs Safety Evaluation l

Report-(SSER-5), Section 6.2.4.1, three programs were implemented

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at CPS to assess the need for operation of the ' containment purge

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system during Operational Conditions 1, 2 and;3 and to investigate l

methods to minimize its use consistent'with ALARA (as low as reasonably achievable) considerations.

Descriptions ~of these I

programs (i.e., the Containment Purge Operational Data Cathering Program, the Containment Access Management Program, and the-Interim Guidelines for, Containment Purge Operations) were provided 3

1 by IP letter U-0716 dated June 29, 1984 and IP letter U 0731 dated September 10, 1984. - As identified in SSER-5,'these programs were intended to be' utilized during the first operating ~ cycle to gather operational data.

Prior to startup from the first refueling outage, IP was required to provide the NRC staff with a-reevaluation of the need to use the containment purge systems l

during Operational Conditions 1, 2 and 3 and criteria for use of l

these systems during the remainder of the plant life. This information was provided by IP letter U 601410 dated April 4, 1989.

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0-Attcchm:nt l'.

to U-601650 Page 3 of 38-The current Bases for Technical-Specification 3/4.6.1.8-reflect the interim guidelinen contained in IP letters dated June 29, 1984-and September 10, 1984.

Since CPS has completed its first.

operating cycle and provided the.information required by SSER 5, the Bases are being revised to reflect the~information contained-in Ip letter U-601410 dated April 4,-1989 The revised.

containment purging guidelines are-consistent with the current.

requirements of Technical Specification 3.6.1.8 and consist of the following criteria which are applicable during Operational-Conditions 1, 2 and'3:

1.

The 36-inch and.12 inch containment purge systems shall be-isolated while.' venting the drywell,-

2.

The 36 inch and'12-inch containment purge syttems shall not be' operated simultaneously, 3.

Operation of the.36 inch containment purge system shall be limited to 250 hours0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br />'per year, q

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Either the 36-inch or 12-inch containment purge system may I

be operated to maintain the containment building air 1

pressure within the;11mits of-the Technical Specifications, j

to reduce airborne radioactivity levels in the containment:

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building to maintain worker. dose A1 ARA,'or for the duration of any welding task or-any task in which chemicals requiring ventilation are used, and j

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Continuous operation'of the 12-inch containment purge system is allowed except while venting the drywell or while operating the 36-inch containment purge system.

Based upon the results of the studies presented in IP letter U 601410 dated April 4, 1989, an interim requirement.has been-deleted regarding keeping the 12-inch containment purge system q

isolated until the containment-building airborne radioactivity.

-q 1evel is greater than or equal to 1/4' Maximum Permissible Concentration (MPC) per 10CFR20 or the I-131 concentration measured in a reactor coolant sataple is greater than 4 x 10 5 microcuries/cc.

Deletion of this requirement will result in lower worker doses and reduced cycling.of the containment purge isolation valves.

Deletion of these requirements continues to be consistent with the requirements of= Technical Specification 3.6.1.8 since continuous operation of the 12-inch containment purge system has been shown to be regttired to support access to

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the containment to perform Technical Specification surveillances during normal conditions while in Operational Conditions 1, 2 and-3.

As stated in IP letter U-601410 dated April 4, 1989, CPS began implementing the new criteria at the start of the second operating cycle.

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Att:chment 1 I

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to U.601650 Page[4of38

-5.

Clarify Bases for-Drywell Internal' Pressure Limits -

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The original design basis accident analysis which produced the calculated peak positive drywell pressure of 18.9 psig was the main steam line break analysis..By IP letter U 601239 dated-September 6,1988, IP requested amendment of the operating license.

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.to allow operation in the Maximum Extended Operating Domain l

(MEOD).c ' Operation in the MEOD was subsequently approved for CPS by Amendment 18. Analysis showed that operation in'MEOD:results in a higher calculated peak positive drywell pressure of 19.7, psig -

following a recirculation line break. Although the-BasesLfor Technical Specification 3/4.6.2.5 were revised in Amendment 18 to reflect the higher calculated peak drywell pressure, they were not revised to reflect that the limiting _ design basis-accident had!

changed from the main steam line break to the recirculation line break. Therefore, the' Bases are being revised to correctly-reflect that the calculated peak positive drywell pressure of 19.7 psig occurs during loss of coolant accident (LOCA) (recirculation-line break) conditions.

6.

Provide Additional Guidance on OPERABILITY of the Reactor Core-I Isolation Cooling (RCIC) System Technical Specification 3.3.3, ACTION c, provides'! the required action to be taken in the event that an Automatic Depressurization System-(ADS) trip system is inoperable.

The ADS trip system'is,

j required to be restored'to OPERABLE status within.7 days,when'the-high pressure core spray (HPCS) system and the RCIC' system.are OPERABLE.

If:the HPCS system or the RCIC system is inoperable,-

the ADS trip system is required to be restored to OPERABLE-status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The ADS actuation instrumentation is requiredito be OPERABLE in Operational Conditions 1,-2 and 3 when reactor-1 pressure is greater than 100 psig.

a Technical Specification-3.5.1, ACTION c, provides the required l

action to be taken in the event that the HPCS system-is:

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inoperable. The HPCS system is required to be restored to OPERABLE status.within 14 days when the RCIC system is OPERABLE.

The ACTION statements of Technical Specification 3.5.1 do not address operation with both the HPCS system and the RCIC system inoperable. The HPCS system is required to be:0PERABLE, per 1

Technical Specification 3.5.1, in Operational' Conditions 1, 2 and 3.

The ACTION statement for Technical Specification 3.7.3 provides the required action to be taken in the event that the RCIC system 1

is inoperable. The RCIC system is required to be restored to OPERABLE status within 14 days when the HPCS system is OPERABLE.

Like Technical Specification 3.5.1 Technical Specification 3.7.3 does not address operation with both the HPCS system and the RCIC system inoperable. The RCIC system is required to be OPERABLE in i

Operational Conditions 1, 2 and 3 when reactor pressure is greater 1

than 150 pstg.

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-Att%hment l' to U.601650 j

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The applicability of each of the 'three Tnchnical' Specifications l

-described above.is different.

The HPCS system is required to be OPERABLE.in Operational Conditions 1, 2 and-3;:the RCIC system is.-

required to be OPERABLE in Operational Condition 1, 2 and 3.when.

l reactor pressure is greater than 150 psig; and the ADS is required' to be OPERABLE in Operational' Condition 1,.2 and 3 when reactori pressure;is greater than 100 psig...As a result, there may be situations in which'an ADS-trip system becomes inoperable or the-HPCS system becomes inoperable while in a plant condition which=

does not require the' operability of the RCIC system (i.e.,

s Operational Condition 2 or'3 with reactor pressure less:than or equal to 150 psig).-- Since Wis-condition would result in entry into the respective ACTION 3ement which provides allowable-outa&*. times basedtupon the <perability of the RCIC system,.

clarification is required in order to preclude unnecessary restrictions on plant operation and unnecessary _ entry into Technical Specification 3.0.3.

Therefore, clarification has:beenE

.t added to the Bases for Technical Specifications'3.3.3,'3.5.1 andl 3.7.3 which states that, although the RCIC system may not be a

considered OPERABLE per Technical Specification 3.7.3, for the -

purposes of satisfying the ACTION requirements of other M

Specifications (3.3.3 and'3.5.1), the RCIC system may be considered OPERABLE when reactor pressure is less - than or equal to i

150 psig.

The bosis for this clarification is as follows:

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The function of the RCIC system is to assure adequate core cooling

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in the event of reactor isolation from'its primary heat sink and.

the loss of feedwater flow to the reactor vessel without requiring' actuation of any of the emergency core-cooling system (ECCS)'

equipment.

The Bases-for the RCIC system state that 150 psig is.

substantially below the pressure for which the low pressure:ECCS can provide adequate core cooling for events which could require.

l the RCIC system.

I Similarly, the function of the ADS is to automatically i

l depressurize the reactor so that flow from the-low pressure ECCS l

can enter the core in time to limit fuel' cladding. temperature to l

1ess than 2200 degrees F in the event of a failure of the HPCS i

system to function properly followingta small-break loss of coolant accident (LOCA)..The Bases for the ADS stato that 100 psig is substantially below the pressure atLwhich-the low pressure ECCS can provide adequate core cooling for events which require the ADS.

Based upon the above, adequate diversity and redundancy.in the core cooling systems exist to provide adequate core cooling at l

reactor pressures less than or equal to 150 psig. Thereforo, considering the RCIC system OPERABLE for the purpose of satisfying the ACTION requirements of the Specifications preserves the licensing basis for assuring the availability of redundant core cooling systems during all plant operating conditions, even when further considering an active single failure.

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Attcchment l' to U 601650 Page-6 of:38

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Provide Additional Guidance Concerninn Entry into and Como11ance:

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with the ACTION Reouirements Associated with the Technical' l

Soecifications for Diesel Generators Tenhnical Specification 3.8.1.1, ACTIONS'b.1e and d, provide the.

required actions to beitaken when.a diesel generator is inoperable, alone or in combination with an inoperable offsite AC.

power source. These ACTIONS require the operability of the remaining offsite AC source (s) to be'demonstr#,ted.~. Additionally, these ACTIONS require the operability of the remaining diesel generators to be demonstrated by starting and loading them for one

-i hour, unless-the diesel generator became inoperable as a result of

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" preplanned preventive maintenance'or testing".

At times the. diesel generators may be removed from' service (i.e.,--

made inoperable)=in order.to perform routine preventive maintenance. This also provides an' opportunity to correct' minor conditions. that do not. impact diesel generator operability (e.g.

i minor leaks).- Good housekeeping practices would. dictate that'such=

3 minor leaks-should be corrected at the earliest convenience.

Since this work would normally be identified on a corrective maintenance-document, the performance of this work.could be construed to be outside the scope.of " preplanned. preventive i

maintenance or testing" enul result in unnecessary testing-of-the remaining diesel generators.- Per the intent.ofLthis Specification, preventive maintenance or testing does not.

p constitute a condition that, by itself, warrants additional.

4 testing of the remaining diesel generators.

Consistent with this intent, the performance of work to correct a condition which by itself would not cause the associated diesel generator'to be inoperable, likewise should not result in'a requirement to

-demonstrate the-operability of the remaining diesel generators.

Industry and the NRC recognize that, unnecessary' testing of: die'sel generators results in increased wear and.potentially reduces their reliability. Therefore, a clarification has been.added'to the Bases which clearly identifies that "proplanned maintenance" includes maintenance which if not performed would'not result in the diesel generator being declared' inoperable.

This will allow the correction of minor conditions which are not severe enough to require the diesel generator to.be declared inoperable-(and thus can be scheduled for the next available diese1~ generator outage)c l

without.resulting in unnecessary testing of the remaining diesel l

generators.

L Additionally, when.a diesel generator is demonstrated operable, as required by these t.CTIONs, the diesel generator _must undergo preparations for testing. These preparations include such items.

as verifying cooling water and lubricating oil levels, checking o

for excessive condensation in the lubricating oil and air start-systems, and barring the engine over to check for leakage into the cylinders. These checks are intended to verify the readiness of.

the' diesel generator to operate and to detect abnormal conditions that could result in damage to the diesel generator during the I;

_Attcchment'1 to U.601650!

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test. As such,;they are necessary in order.to prevent serious, damage to the engino should an adverse condition exist.7 However, barring the engine over requires the diesel engine control-switch to be placed in the lockout position in order to preclude serious 1 injury to personne1' should the ' diesel generator receive a start signal during the barring operation.

If the Division I or Division II diesel generator becomes.,

inoperable, the remaining diesel generators must be demonstrated-operable per Technical Specificatian 3.8.1.1 ACTION b.. As stated above, when a diesel generator is demonstrated operable,:the.

diesel generator must undergo preparations for testing.; If a diesel generator (Division I or Division II) is considered l inoperable during the performance of these pre start. preparations, entry into Technical Specification-3.8.1.1. ACTION g, would be= j required.

Technical Specification 3.8.1.1, ACTION g, requires theD operability of the offsite AC sources-to be demonstrated.

Additionally, ACTION g requires the operability.of the Division-III diesel generator to be demonstrated.

Footnote "*"'further specifies that the Division III diesel generator,must bei demonstrated operable regardless of when the Division I-and-Division II diesel generators are restored to operable status; j

This requirement would result in running the Division III diesel:

g generator even if it had just been run in response. to an entry l

into ACTION b.

In order to avoid this conflict, a clarification j

has been added to the Bases which clearly identifies that if a diesel generator is temporarily made inoperable' solely; to demonstrate its operability in accordance with the ACTION statements, it need not be considered inoperable.

However, if'an adverse condition that affects the diesel generator's operability i

is observed during the performance of the pre start preparations, entry into ACTION g for inoperability of both Division I and II.

l diesel generators would be required.

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Clarifv Bases for Soecific Gravity Limits for Station' Batteries-l Technical Specification Table 4.8.2.1 1 identifies the specific i

gravity limits for the battery cells of the safety-related Division I, II, III and IV station batteries. At least once per seven days, the battery parameters listed in Tablo 4~.8.2.1-1..must.

be verified to meet the Category A limits.,At least once per 92 days, the battery parameters listed in Table 4.8.2.1-l'must boy.

verified to meet the Category B limits.

If the Category A limits are not satisfied, the Category B limits must be verified to be met and compliance with the Category A limits must be restored 1

within seven days; otherwise, the affected battery must be declared inoperable.

Further, if the Category B limits are not satisfied, the Category B allowable values must be verified to be a

met and compliance with the Category B limits must be restored

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within seven days; otherwise, the affected battery must be declared inoperable.

If at any time the Category B allowable values are not met, the battery must be declared inoperable.

Attcchm3nt 1'

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.Page 8 of 38 As stated:in'the Bases, the float voltage'and specific gravity of j

1 a battery cell provide an indication of the charge state and 1

available' capacity of the cell.- The Category A limits of Table.

sci 4.8.2.1-1 for the pilot cells are characteristic of.a charged cell j

with adequate capacity and are indicative of the state-of the entire battery.

The Category B limits of Table 4.8.2.1-1 for each j

cell and.the entire battery are slightly less restrictive.than the Category A limits for'the pilot c611s but are selected to ensure, that the entire battery has: sufficient capacity to supply the required loads,: The Category B allowable values of Table 4.8.2.1

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1 for the battery are'slightly less restrictive than the Category B limits for the battery,1but still ensure that the reduction in-battery capacity is less than the margin provided.in sizing, and-

j therefore, that the battery is still'capableiof supplying its design loads.

i The current Bases for the specific gravity limits of Table.

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4.8.2.1-1 provide a basis for each of the-Category A and B limits-and Category BJallowable values based upon the amount'the specific.

1 gravity.'is "below'the manufacturer's full charge-specific gravity". The battery cells originally installed at CPS are type KC-11 and LC 17 manufactured.by C and D Power Systems.

However, C

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and D Power Systems no longer. manufactures these cells. The

I replacement cells are type KCR-11 and LCRL17 which.have a slightly 3

different nominal specific gravity. ' Design and sizing calculations performed for CPS demonstrate.that these cells are adequate replacements for the KC-11 and LC 17 cells.. Since no i

loads have been;added to the= batteries and;the required capacity determined,in accordance with IEEE 485 remains unchanged,=the' j

i replacement cells continue to maintain the original margin of safety provided in the sizing:of the batteries.

Since the limits currently identiflod in-Technical Specification ~ Table 4.8.2.1 1 4

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continue to serve as adequate acceptance criteria:for determiningc I

battery' cell operability-and capability with no reduction in the a

required safety-margin, no changes to Technical Specification 3

4 Table 4.8.2.1-l'aro required.

However, because the replacement cells utilize a.different nominal specific gravity,-the Bases must be revised to reflect absolute specific gravity-lim?ts as opposed 1

to values "below the manufacturer's full' charge spdfic gravity."

i-The remaining changes to these Bases are being made to make the l

Bases consistent with Technical Specification Table 4.8.2.1-1.

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