ML20128E620

Proposed Tech Specs Reflecting Current NRC Positions Re Leak Detection & ISI Schedules,Methods,Personnel & Sample Expansion,Per GL 88-01

ML20128E620
Person / Time
Site:	Cooper
Issue date:	02/01/1993
From:	NEBRASKA PUBLIC POWER DISTRICT
To:
Shared Package
ML20128E601	List: ML20128E596 ML20128E620
References
GL-88-01, GL-88-1, NUDOCS 9302100526
Download: ML20128E620 (9)

Text

i REVISED TECilNICAL SPECIFICATION PAGES

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LIMITTiri coffDITIONS FOR OPERATION SURVEILLANCE REOUTREMDITS l

3.6.C.

Coolant Leakaqe 4.6.C.

Coolant Leakane l

1.

Any time irradiated fuel is in the 1.

Reactor coolant system leakage shall reactor vessel and reactor coolant

-be checked _ by the sump flow s

temperature is above 2120F, reactor measuring systems and drywell air 4

coolant leakage into the primary sampling. system and recorded' at least once per shift, not to exceed containment shall not exceed a 5 gpm 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

unidentified leak

rate, 25 gpm identified leak rate, or a

2. gym increase in unidentified leak rate within the previous 24_ hour period.

?

If these limits cannot be met, an orderly SHUTDOWN shall be init.tated and the reactor shall be in a COLD j.

SHUTDOWN CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

I 2.

Each of the sump flow measuring systems - shall be OPERABLE during REACTOR POWER OPERATION.

From and after the date that one 'of these j

systems is made or found to be inoperable for. any reason and the s

sump flow leak rate cannot be

. quantified, REACTOR POWER OPERATION is permissible only during the succeeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, unless the system is sooner made OPERABLE.

If leakage can be quantitatively measured by manually pumping the sump or measuring the difference in i

sump level, then REACTOR -POWER j

OPERATION is permissible during the succeeding 30 days, unless the sump flow measuring system is sooner made OPERABLE.

3.

The drywell air sampling system shall -be OPERABLE during REACTOR POWER OPERATIOtt From and af ter the date that this system is made or found to be inoperable for any reason, REAR"'OR POWER OPERATION is permissible only during the succeeding 30 days unless the system l

is sooner made OPERABLE.

1 4.

If the requirements of specification 3.6.C.2 or 3 cannot be. met, an orderly SHUTDOWN shall be initiated and the reactor shall be in a COLD SHUTDOWN CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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LIMITING CONDITIONS FOR OPERATION SURVEII. LANCE RFOUTREMENTS I

3.6,E.

Jet Pumos 4.6.E, Jet Pum,,p, sg j

1.

Whenever the reactor is in the 1.

Whenever there is recircula.*. ion flow l

STARTUP or RUN modes, all jet pumps with the reactor in the ST,RTUP or J

shall be OPERABLE.

If it is RUN modes, jot pump OPElABILITY determined that a jet pump is shall be checked daily by v rifying inoperable, or if two or more jet that the following conditio.s do not l

pump flow instrument failures occur

' occur simultaneously:

and cannot be corrected within j

l 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, an orderly SHUTDOWN shall a.

The recirculation pump flow differs l

be initiated and the reactor shall by more than 15%

from the be in a COLD SHUTDOWN CONDITION establisied speed flow within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

characteristics.

b.

The indicated value of' core flow rate varies from the value derived from loop flow measurements by more l

than 10%.

)

c.

The diffuser to lower plenum differential pressure reading on an-individual jet pump varies from the mean of all jet pump differential pressures by more than 10%.

F.

Recirculation Pumo Flow Mismatch F.

Recirculation Pumn Flow Mismatch l

l l

1.

Following one recirculation p_ ump 1.

_ Deleted, operation, the discharge valve of the low speed recirculation pump may not be opened unless the speed of the faster pump is equal to or less than 50% of its rated speed.

C.

Inservice Inspection G,

Inservice Insnection l

1.

To be considered

OPERABLE, 1.

Inservice inspection shall-be l

components shall satisfy-the performed in accordance with the i

requirements contained in Section XI requirements for ASME Code Class 1, of the ASME Boiler and Pressure 2,

and 3 - components - contained in Vessel Code and applicable Addenda Section XI of the ASME Boiler-and for continued service of ASME Code Pressure Vessel-Code and applicable Class 1, 2, and 3 components. except Addenda as required ' by CFR 50, where relief has been granted by the

_Section 50.55a(g),

except where Commission' pursuant to 10 CFR 50, relief _ has - been granted by the Section 50.55a(g)(6)(1).

Commission pursuant to 10 CFR 50, Section 50.55a(g)(6)(1).

2.

The inservice inspection program for i

piping identified in NRC Generic Letter 88-01 shall be performed in accordance with the staff positions-on schedule, methods, personnel, and sample expansion included in this generic letter.

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3. 6.C & 4.6.C BASES (cont'd.)

indicates that leakage from a crack can be detected before the crack grows to a dangerous or critical size by mechanically or thermally induced cyclic loading,'or stress corrosion cracking or some other mechanism characterized by gradual crack growth.

This evidence suggests that for leakage somewhat greater than the-limit specified - for unidentified leakage, the probability is small that imperfections or cracks, associated with such leakage would grow rapidly.

However, the establishment of allowable unidentified-leakage greater than that given in 3.6.C on the basis of the data presently available _.- would be premature because of uncertainties associated with the data.

Leakage limits of a L 2 - gpm increase within any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period and a maximum of 5 gpm are specified in i

3.6.C and are also-supported in Generic-Letter 88-01. 4The experimental i

and analytical data suggest a reasonable; margin of safety that such leakage magnitude would not result from a crack approaching the critica11 size for rapid propagation. Leakage _less than the magnitude specified can' be detected reasonably in _a matter of a few hours utilizing the available-leakage detection schemes, and if the origin cannot be determined in a i

reasonably - short time the plant should be SHUTDOWN to allow further l

investigation-and corrective action.

The-total leakage rate consists of all leak' age,

identified 'and unidentified, which flows'to the drywell floor drain and equipment drain' sumps.

3 The capacity of the drywell floor sump pumps is 50 gpm and'the capacity of j

the.drywell equipment sump pumps is also 50 gpm.. Removal of 25 gpm from either of these sumps can be accomplished with margin.

I Reactor coolant leakage is also sensed.by the drywell air sampling system which detects gaseous,. particulate, and iodine radioactivity.. Leakage can j

also be detected by - area temperature detectors, h Adity detectors and-pressure instrumentation.

Due to the many and varied ways of detecting

_ primary leakage, a 30 day allowable repair time is justified.

i D.

Safety and Relief Valves l

.The safety and relief valves. are required to be OPERABLE ' above, the l

pressure (113 psig) at which the core spray system is not _ designed to.

deliver full flow. The pressure relief system for Cooper Nuclear Station i

has been sized to meet two design bases.

First, the total safety / relief j_

valve capacity has been established to meet the_ overpressure protective criteria - of the L ASME code.

Second, _the distribution of this' required capacity between safety valves -and. relief valves has been set Eto meet design basis IV-4.2.1 of - the. USAR which states -that the nuclear system

_l relief valves shall prevent opening of the safety valves -during normal plant isolations and load rejections.

The details of the analysis - which shows compliance with the ASME code requirements-is presented in subsection IV._4 of the FSAR and the. Reactor Vessel Overpressure Protection Summary Technical Report - presented in.

4 question 4.20 of ' Amendment 11 to the FSAR.

Results of the overpressure q

protection analysis are provided in the current reload license' document.

Experience in relief and safety valve operation'shows that a. testing of--

50_ percent of the valves per year is adequate to detect failures or

~ deteriorations.

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LDtITI!(G CONDITIONS FOR OPERATION.

SURVEILLANCE KEOUIREMENTS Coolant Leakate Coolant Leakace

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Any time irradiated fuel is in the 1.

Reactor coolant system, leakage shall l

the reactor vessel and reactor be checked by the sump and ir samp-coolant temperature is above 212*F, ling system and recorded a least nce per #

{demlD reactor coolant leakage into the 3

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(cannot be met, an orderly <; u d vn CO M D'1 T @ w k;% 2tl keta _rt, shall be initiated and the reactor shall be in a(C_old Shutdos % ditiol '

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-tlMITIN.G CONDITIONS-FOR-OPERATION

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SURVEILLANCE REOUIREMENTS 3.6.E Jet Pumps 4.6.E.

Jet Pumps 1.

Whenever the reactor is in the s art-1.

Whenever there is recircu gla ion flow u or et n_ mod e all jet pumps shall with the reactor in startup or I

be operrb M If it is determined run modes, jet pump (o'perab_ility)shall iat a jet pump is inoperable, or be checked daily by v.rifying that the it_two or more jet pump flow in-following conditions do not occur sim-t failures occur and cannot ultaneously:

ALL cAO be correcte ~ within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, an orderly shut own shall be initi red a.

The recirculation pump flow differs

( and the reactor shall be in a old by more than 15% from the established wnConditi_on]vithin24 hours.

speed flow characteristics.

[ g3 b.

The indicated value of core flow rate varies from the value derived from loop i

flow measurements by more than 10%.

c.

The diffuser to lower plenum differen-tial pressure reading on an individual jet pump varies from the mean of all jet pump dif f erential pressures by more than 10%.

l F.

Recirculation Pump Flow Mismatch F.

Recirculation Pump Flow Mismatch l

goe8 1.

Deleted.

Following one recirculation pump operation, the discharge valve of h

the low speed recirculation pump j

may not be opened unless the speed of the taster pump is equal to or less than 50% of its rated speed.

G.

Inservice Inspection G.

Inservice Inspection pgCM' oo L

Inservice inspection shall be per-1 To be considered o rsble, com-formed in accordance with the ponents shall satisfy the require-requirements for ASME Code Class 1, ments contained in Section XI of 2, and 3 components contained in the ASME Boiler and Pressure Vessel Section XI of the ASME Boiler and Code and applicable Adde da for Pressure Vessel Code and applicable continued service of ASME Code Addenda as required by 10 CFR 50, Clasc 1, 2, and 3 components except Section 50.55a(g), except where where relief has been granted by the relief has been granted by the commission pursuant to 10 CFR 50 Commission pursuant to 10 CFR 50, Section 50.55a(g)(6)(11 Section 50.55a(g)(6)(1).

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Ilml i A 3fm Mcramas usMin en M fop pe LA I-3.6.C &a4.6.C BASES (cont'd.)

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indicates that leakage from a c a' can be1 detected before the crack grows t P i "

a dangerous or critical size by mechanically or thermally induced cyclic loading, i

or stress corrosion cracking or some other mechanism characterized by gradual crack growth. This evidence suggests that for leakage somewhat greater than the limit specified for unidentified leakage, the probability is small that imperfections or cracks, associated with such leakage would grow rapidly.

However, the establishment of allowable unidentified leakage greater than i

that given in 3.6.C on the basis of the data presently available would_b premature because of-uncertainties associated with the data.

F r lerEns

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ifle au ^e.C,/heexperimentalandanalytical

~w data suggest a reasonable margin of safety that such leakage magnitude would not result from a crack approaching the critical size for rapid propagation.

Leakage less than the magnitude specified can be_ detected reasonably in a matter of a few hours utilizing the available leakage detection schemes.-

and if the orinin_cannot be determined in a reasonably short time the plant shouldbe{ shutdown)toallowfurtherinvestigationandcorrectiveaction.

i 1 Ali ca n i

The total leakage rate consists of all leakage, identified and unidentified, j

which flows to the drywell floor drain and equipment drain sumps.

4 i

The capacity of the drywell floor sump pumps is 50 gpm and the capacity of the drywell equipment sump pumps is also 50 gpm, Removal of 25 gpm from cither of these sumps can te accomplished with margin.

el ale **=rhaa 3)3 b Reactor coolant leakage is also sensed by the cer ri:n nt redirti

it n ihg 3

-ea4-tr which : ::: gr r bete. ;- perticulrte end 1 din:

rell 2: 57 cryger m d M drrger' - lycere. Leakage can also be detected by area. temperature detectors, humidity detectors an pressure instrumentation.. Due to the many and l

varied ways of detecting primary eakage, a 30 day allowable repair time is

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D.

Safety and Relief Valves

( All CnM.,

The safety and relief valves are required to be

'above the. pressure-(113 psig) at which the core spray system is not designed'to deliver full-flow. The pressure relief system for Cooper Nuclear. Station has been sized I

to meet two design bases. First, the total safety / relief _ valve capacity has g g been established to meet the overpressure protective criteria of the'ASME code. Second, the distribution.of this_ required capacity between safety' u R valves and relief valves has been set to meet design basis IV.4.2.1 cf cut 4

" '- which states that the nuclear ' system relief valves _ shall pre -

vent opening of the safety' valves during normal plant isolations and load.

rej e_ctions.

1 The details of the analysis which shows compliance with the ASME-code' require-ments is presented in subsection IV.4 of the-FSAR and the Reactor Vessel Over-pressure Protection Sammary Technical Report presented in question 4.20 of Amendment 11 to the FSAR. =Results of the overpressure protection analysis:

are provided in the current reload license document.

Experience in relief and safety. valve operation shows that a testing'of 50 4

l percent of the valves-per year is adequate to detect failures or deteriorations.

-149.

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