ML20149C505

Amend 35 to License NPF-29,changing Definition of Core Alteration in Tech Specs to Include Certain Exceptions & Changing Footnotes in Tech Specs to Be Consistent W/New Definition

ML20149C505
Person / Time
Site:	Grand Gulf
Issue date:	09/10/1987
From:	Berkow H Office of Nuclear Reactor Regulation
To:
Shared Package
ML20149C511	List: ML20149C505 ML20234B764
References
TAC-65511, NUDOCS 8709180465
Download: ML20149C505 (19)

Text

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UNITED STATES

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NUCLEAR REGULATORY COMMISSION wAssiNoTow. o. c. nosss a

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MIS $15$1PPI POWER A LIGHT COMPANY SYSTEM ENERGY RESOURCES, INC.

l SOUTH MIS $15$1PPI ELECTRIC POWER ASSOCIATION DOCKET NO. 50 416 l

GRAND GULF NUCLEAR STATION,tlNIT 1 f

AMENDMENT TO FACILITY OPERATING LICENSE I

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3 Amendm nt No. 35 f

License No NPF-29 i

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The Nuclear Regulatory Comission (the Commission) has found that J

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The application for amenhnt by Mississippi Power & Light Company, i

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System Energy Resources, Inc. (fomerly Middle South Energy. Inc.)

7 and South Mississippi Electric Pcver Association. (the licensees) i dated June 3. 1987, as supplemented June 22, 1987, complies with the j

standards and requirements of the Atomic Energy Act of 1954, as

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amended (the Act), and the Comission's rules and regulations set

l forth in 10 CFR Chapter It 8.

The facility will operate in conformity with the application, the f

provisions of the Act, and the rules and regulations of the Comission; 3

C.

There is reasonable assurance (1) that the setivitie.= authorized by l

this amenhnt can be conducted without endangering the health and i

safety of the public, and (ii) that such activities will be conducted i

in compliance with the Comission's regulations.

D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the publict and 1

J E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

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Accordingly, the license is. amended by changes to the Technical Specifications

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j as indicated in the attachment to this license amendment, and paragraph 2.C.(2) 1 of Facility Operating License No. NPF-29 is hereby amended to read as follows:

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1 Technical Specifications 1

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i The Technical Specifications contained in Appendix A and the E.vironmental Protection Plan contained in Appendix R, as revised through Amendment No. 35, are hereby incorporated into this license.

j System Energy Resources, Inc. shall operate the facility in accordance i

j with the Technical Specifications and the Environmental Protection i

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Plan, t

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This Itcense amendment is effective as of its date of issuance.

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FOR THE NUCLEAR REGULATORY COW !$$!ON f

i ktT b b i

I erbert N. Berkow Director i

Project Directorate 11-2 l

i Division of Reactor Projects-!/II 1

Office of Nuclear Reactor Regulation 1

Attachment:

j Changes to the Technical Specifications s

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j Date of Issuarce: September 10,1g87 1

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ATTACHMENT TO LICENSE AMENDMENT N0. 35 j

FACILITY OPERATING LICENSE No. NPF 29 DOCKET NO. 50 416 t

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Replace the following pages o' the Appendix "A" Technical Specifications with i

the attached pages. The revised pages are identified by Amendment number and i

contain vertical lines indicating the area of change, The corresponding i

overleaf page(s) have been provided to maintain document completeness.

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3/4 3-4 3/4 3-4 3/4 7-11 3/4 7-11 3/4 9-3 3/4 9 3 3/4 9-7 3/4 9 7 83/4 7-3 B3/4 7 3

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1.0 DEFINITIONS The following terms are defined so that unifom interpretation of these sy cifications may be achieved. The defined terms appear 1r capitalized type and shall be applicable throughout these Technical Specifications.

ACTION 1.1 ACTION shall be that part of a Specificatian which prescribes remedial sensures required under designated conditions.

AVERAGE PLANAR EXPOSURE 1.2 The AVERAGE PLAMAR EXPOSURE shall be applicable to a specific planar j

height and is equal to the sum of the exposure of all the fuel rods in the specified bundle at the specified height divided by the neber of fuel rods in the fuel bundle.

. AVERAGE PLANAR LINEAR HEAT GENERATION RATE 1.3 The AVERAGE PLAMAR LINEAR NEAT GENERATION L Mi (APLHGR) shall be applicable to a specific planer height and is equal to the sum of the i.INEAR HEAT GENERATION RATES for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fue.1 bundle.

CHANNEL CAL 18 RATION 1.4 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the chsnnel sonitors. The CHANNEL 1

CALIBRATION shall encompass the entire channel including the sensor and alarm i

and/or trip functions, and shall include the CHANNEL FUFCTIONAL TEST. The CHANNEL CALIBRATION say be performed by any series of sequential, overlapping or total channel steps such that the entiro channel is calibrated.

CHANNEL CHECK 1.5 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall. include, where possible, comparison of the channel indication and/or stat'us with other indi-cations and/or status derived iros independent instrument channels measuring the same parameter.

CHANNEL FUNCTIONAt. TEST 1.6 A CHANNEL FUNCTIONAL TEST shall be:

a.

Analog channels - the injection of a simulated signal into the-channel as close to the sensor as practicable to verify OPERABILITY 4

including alare and/or trip functions and channel failure trips.

b.

Sistable channels - the injection of a simulated signal into the sensor to verify OPERA 31LITY including alare and/or trip functions.

The CHANNEL FLECTIONAL TEST may be performed by any series of seguential.

everlapping or total channel steps such that the entire channel is tested.

F GRAND GULF UNIT 1 1-1

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DEFINITIONS CORE ALTERATION 1.7 CORE ALTERATION shall be the addition, removal, relocation or movement of fuel, sources, incore instruments or reactivity controls within the reactor pressure vessel with the vessel head removed and fuel in the vessel.

Normal movement of the SRMs, IRMs. LPRMs, TIPS, or special movable detectors is not considered to be CORE ALTERATION.

Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe conservative position.

CRITICAL POWER RATIO i

1.8 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the assembly which is calculated by application of the XN-3 correlation to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

DOSE EQUIVALENT I-131 1.9 OOSE EQUIVALENT I-131 shall be that concentration of I 131, microcuries per gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132,1-133, I-134, and I-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites."

ORWELL INTEGRITY 1.10 DRWELL INTEGRITY shall exist when:

a.

All drywell penetrations required to be closed during accident conditions are either:

1.

Capable of being closed by an OPERABLE drywell automatic l

isolation system, or j

2.

Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, except as provided in Table 3.6.4-1 of Specification 3.6.4.

b.

The drywell equipment hatch is closed and sealed.

j c.

The drywell airlock is in compliance with the requirements of Specification 3.6.2.3.

d.

The drywell leakage rates are within the limits of Specification j

3.6.2.2.

e.

The suppression pool is in compliance with the requirements of Specification 3.6.3.1.

f.

The sealing mechanism associated with each drywell penetration; e.g., welds, bellows or 0-rings, is OPERABLE.

GRAND GULF-UNIT 1 1-2 Amendment No. 35 l

i 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN l

LIMITING CONDITION FOR OPERATION 3.1.1 The SHUTDOWN MARGIN shall be equal to or greater than:

a.

0.38% delta k/k with the highest worth rod analytically determined, or b.

0.28% delta k/k with the highest worth rod determined by test.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3, 4 and 5.

ACTION:

With the SHUTDOWN MARGIN less than specified:

a.

In OPERATIONAL CONDITION 1 or 2, reestablish the required SHUTDOWN MARGIN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, b.

In OPERATIONAL CONDITION 3 or 4, immediately verify all insertable control rods to be inserted and suspend all activities that could reduce the SHUTDOWN MARGIN.

In OPERATIONAL CONDITION 4, establish SECONDARY CONTAINMENT INTEGRITY within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

c.

In OPERATIONAL CONDITION 5, suspend CORE ALTERATIONS and other I

activities that could reduce the SHUTDOWN MARGIN and insert all insertable control rods within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Establish SECONDARY CONTAINMENT INTEGRITY within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

SURVtILLANCE REQUIREMENTS 4.1.1 The SHUTDOWN MARGIN shall be determined to be equal to or greater than specified at any time during the fuel cycle:

a.

By measurement, prior to or during the first startup after each refueling, b.

By measurement, within 500 MWD /T prior to the core average exposure at which the predicted SHUTDOWN MARGIN, including uncertainties and calculation biases, is equal to the specified limit.

c.

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after detection of a withdrawn control rod that is immov:tble, as a result of excessive friction or mechanical interfer-ence, or is untrippable, except that the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod.

GRAND GULF-UNIT 1 3/4 1-1 Amendment No.35

1 REACTIVITY CONTROL SYSTEMS 3/4.1.2 REACTIVITY AN0MALIES LIMITING CONDITION FOR OPERATION 3.1.2 The reactivity difference between the monitored core k,ff and the predicted core k,ff shall not exceed 1% delta k/k.

APPLICABILITY: OPERATIONAL CCNDITIONS 1 and 2.

ACTION:

With the reactivity diffarence Greater ther) 1% t's'lta k/k:

I Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, perform an analysis to deterni,ne and enlain the a.

cause of the reactivity difference; operation may continue if the difference is aglained and corrected.

b.

Otherwise', be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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$URVEILLANCE REQUIREMENTS 4.1.2 The reactivity difference between the sonitored core k,ff and the predicted core k,ff shall be verified to be less than or equal to 1% delta k/,k:

During the first startup following CORE ALTERATIONS, and l

a.

b.

At least once per 1000 M/T during POWER OPERATION.

i GRAW GULF-UNIT 1 3/4 1-2 Amenement No. 23 l

OCT 2 41966 r

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REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

ACTION.

(Continued) b.

With a "slow" control rod (s) not satisfying ACTION a.1, above:

1.

Declare the "slow" control rod (s) inoperable, and 2.

Perform the Surveillance Requirements of Specification 4.1 3.2.c at least once per 60 days when operation is continued with three or more "slow" control rods declared inoperable.

Otherwise, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, c.

With the maximum scram insertion time of one or more control rods exceed-ing the maximum scram insertion time limits of Specification 3.1.3.2 as determined by Specification 4.1.3.2.c, operation may continue provided that:

1.

"Slow" control rods, i.e., those which exceed the limits of Specifi-cation 3.1.3.2, do not make up more than 20% of the 10% sample of con-trol rods testad.

2.

Each of these "slow" control rodt satisfies the limits of ACTION a.1.

3.

The eight adiacent control rods surrounding each "slow" control rod are:

a)

Demonstrated through measurement within 12 heurs to satisfy the maximum scram insertio.n time liraits of Specification 3.1.3.2, and b)

OPERABLE.

4.

The total number of "slow" control rods, as determined by Specifica-tion 4.1.3.2.c, when added to the sum of ACTION a.3, as determined by Specification 4.1.3.2.a and b, does not exceed 7.

Otherwise, be in at least HOT SHUTOOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

d.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCEREqU,IREMENTS 4.1.3.2 The maximum insertion time of the control reds shall be demonstrated through measurement with reactor coolant pressure greater than or equal to 950 psig and, during single control rod scram time tests, the control rod drive pumps isolated from the accumulators:

a.

For all control rods prior to THERMAL POWER exceeding 40% of RATED THERMAL POWER following CORE ALTERATIONS

• or after a reactor shutdown that is greater than 120 days, b.

For specifically affected individual control rods ** following mainten-ance on or modification to the control rod or control rod drive system which could affect the scram insertion time of those specific control rods, and c.

For at least 10% of the control rods, on a rotating basis, at least once per 120 days of POWER OPERATION.

"Except normal control rod movement.

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• • The provisions of Specification 4.0.4 are not applicable for entry into OPERATIONAL CONDITION 2 provided this surveillance is completed prior to entry into OPERATIONAL CONDITION 1.

j GRAND GULF-UNIT 1 3/4 1-7 Amendment No.35 l

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REACTIVITY CONTROL SYSTEMS CONTROL ROD SCRAM ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.1.3.3 All control rod scram accumulators shall be OPERABLE:

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2 and 5*.

l ACTION:

a.

In OPERATIONAL CONDITIONS 1 and 2:

I 1.

With one control rod scram accumulator inoperable, within 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />s:

i a)

Restore the inoperable accumulator to OPERABLE status, or b)

Declare the control rod associated with the inoperable j

i accumulator inoperable, Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

2.

With more than one control rod scram accumulator inoperable, declare the associated control rods inoperable and:

f a)

If the control rod associated with any inoperable scram accumulator is withdrawn, immediately verify that at least one control rod drive pump is operating by inserting at least one withdrawn control rod at least one notch or place the reactor mode switch in the Shutdown position.

b)

Insert the inoperable control rods and disarm the associated directional control valves either:

1)

Electrically, or 2)

Hydraulically by closing the drive water and exhaust water isolation valves.

Otherwise, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, b.

In OPERATIONAL CONDITION 5*:

1.

With one withdrawn control rod with its associated scram accumulator inoperable, insert the affected control rod and disarm the associated directional control valves within one hour, either:

a)

Electrically, or b)

Hydraulically by closing the drive water and exhaust water isolation valves.

2.

With more than one withdrawn control rod with the associated scram accumulator inoperable or with no control rod drive pump 1

operating, immediately place the reactor mode switch in the Shutdown position, j

c.

The provisions of Specification 3.0.4 are not applicable.

"At least the accumulator associated with each withdrawn control rod.

Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

GRAND GULF-UNIT'1 3/4 1-8

1 TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION 8

9 APPLICABLE' MINIIRM ij OPERATIONAL OPERABLE CHAfMELS q

FUNCTIONAL INIIT CONDITIONS PfR TRIP SYSTEM (a)

ACTION-9.

Scram Discharge Velease tinter Level - High a.

Transmitter / Trip Unit

1. 2, 2

1 III S

2 3

b.

Fleet Switch 1, 2, 2

1 I8) 5 2-3

10. Turbine Step Valve - Closure 1(h) 4 6

3 11.

Turbine Centrol Valve Fast Closure' ow1(h) 2 6

Valve Trip System 011 Pressure - L

12. Reacter Mode Switch Shutdown Position 1, 2 2

1

3. 4 2

7 5

2 3

13. Menesel Scram

1. 2 2

1

3. 4 2

8 5

2 9

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a INSTRUMENTATION TABLE 3.3.1.1 (Continued) i REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION ACTION 1 Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 2 Verify all insertable control rods to be inserted in the core and lock the reactor mode switch in the SHUTDOWN position within one hour.

ACTION 3 Suspend all operations involving CORE ALTERATIONS *, and insert all insertable control rods within one hour.

ACTION 4 Be in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTION 5 Be in STARTUP with the main steam line isolation valves closed within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 6 Initiate a reduction in THERMAL POWER within 15 minutes and I

reduce turbine first stage pressure to less than the automatic l

bypass setpoint within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

1 ACTION 7 Verify all insertable control rods to be inserted within one hour.

ACTION 8 Lock the reactor mode switch in the SHUTDOWN position within one hour.

ACTION 9 Suspend all operations involving CORE ALTERATIONS *, and insert all insertable control rods and lock the reactor mode switch in the SHUTDOWN position within one hour.

Also suspend replacement of LPRM strings unless SRM instrumentation is l

OPERABLE per Specification 3.9.2.

GRAND GULF-UNIT 1 3/4 3-4 Amendment No.35 l

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) e.

Functional Tests During the first refueling shutdown and at least once per 18 months thereafter during shutdown, a representative sample of snubbers shall be tested using one of the following sample plans for each type of snubber. The sample plan shall be selected prior to the test period and cannot be changed during the test period.

The NRC Regional Administrator shall be notified in writing of the sample plan selected prior to the test period or the sample plan used in the prior test period shall be implemented:

l 1)

At least 10% of the total of each type of snubber shall be functionally tested either in place or in a bench test.

For each snubber of a type that does not meet the functional test i

acceptance criteria of Specification 4.7.4.f an additional 5% l of that type of snubber shall be functionally tested until no more failures are found or until all snubbers of that type have been functionally tested; or 2)

A representative sample of each type of snubber shall be func-tionally tested in accordance with Figure 4.7.4-1.

"C" is the total number of snubbers of a type found not meeting the accept-ance requirements of Specification 4.7.4,f.

The cumulative num-ber of snubbers of a type tested is denoted by "N".

At the end i

of each day's testing, the new values of "N" and "C" (previous day's total plus current day's increments) shall be plotted on Figure 4.7.4-1.

If at any time the point plotted falls in the "Reject"regionallsnubbersofthattypeshallbefunctionally tested.

If at any time the point plotted falls in the "Accept region, testing of snubbers of that type may be terminated.

When the point plotted lies in the "Continue Testing" region, additional snubbers of that type shall be tested until the point falls in the "Accept" region or the "Reject" region, or all the snubbers of that type have been tested; or 3)

An initial representative sample of 55 snubbers shall be func-tionally tested.

For each snubber type which does not meet the functional test acceptance criteria, another sample of at least one-half the size of the initial sample shall be tested until the total number tested is equal to the initial sample size multi-plied bv the factor, 1 + C/2, where "C" is the number of snubbers found which do not meet the functional test acceptance criteria.

The results from this sample plan shall be plotted using an "Ac-cept" line which follows the equation N = 55(1 + C/2).

Each snubber point should be plotted as soon as the snubber is tested.

If the point plotted falls on or below the "Accest" line, testing of that type of snubber may be terminated.

If tw point plotted falls above the "Accept" line, testing must continue until the l

point falls in the "Accept" region or all the snubbers of that type have been tested.

i GRAND GULF-UNIT 1 3/4 7-11 Amendment No.35l

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

Testing equipment failure during functional testing may invalidate that day's testing and allow that day's testing to resume anew at a later time, providing all snubbers tested with the failed equipment during the day of equipment failure are retested.

The representative sample selected for the functional test sample plans shall be randomly selected from the snubbers of each type and reviewed before beginting the testing. The review shall ensure as far as practical that they are representative of the various configurations, operating environ-ments, range of size, and capacity of snubbers of each type.

Snubbers placed in the same locatio,ns as snubbers which failed the previous functional test shall be retested at the time of the next functional test but shall not be included in the sample plan.

If during the functional testing, additional sampling is required due to failure of only one type of snubber, the functional testing results shall be reviewed at the time to determine if additional samples should be limited to the type of snubber which has failed the functional testing, f.

Functional Test Acceptance Criteria The snubber functional test shall verify that:

1)

Activation (restraining action) is achieved within the specified range in both tension and compression; 2)

Snubber bleed, or release rate where required, is present in both tension and compression, within the specified range; 3)

For mechanical snubbers, the force required to initiate or main-tain motion of the snubber is within the specified range in both directions of travel; and 4)

For snubbers specifically required not to displace under continuous load, the ability of the snubber to withstand load without displacement.

Testing methods may be used to measure parameters indirectly or parameters other than those specified if those results can be correlated to the specified parameters through established methods.

g.

Functional Test Failure Analysis An engineering evaluation shall be made of each failure to meet the functional test acceptance criteria to deterstne the cause of the failure.

The results of this evaluation shall be used, if applicable, in selecting snubbers to be tested in an effort to determine the OPERABILITY of other snubbers irrespective of type which may be subject to the same failure mode.

For the snubbers found inoperable, an engineering evaluation shall be performed on the components to which the inoperable snubbers are attached.

The purpose of this engineering evaluation shall be to deter-eine if the components to which the inoperable snubbers are attached were adversely affected by the inoperability of the snubbers in order to ensure that the component remains capable of meeting the designed service.

GRAND GULF %

1 3/4 7-12 l

I REFUELING OPERATIONS 3/4.9.2 INSTRUMENTATION J

LIMITING CONDITION FOR OPERATION

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3.9.2 At least 2 source ran2e monitor * (SRM) channels shall be OPERABLE and inserted to the normal operating level with:

a.

Continuous visual indication in the control room, b.

One of the required SRM detectors located in the quadrant where CORE ALTERATIONS are being performed and the other required SRM detector located in an adjacent quadrant, and c.

Unless adequate shutdown margin has been demonstrated, the shorting the time any control rod is withdrawn.gcuitry prior to and during links shall be removed from the RPS ci i

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APPLICABILITY:

OPERATIONAL CONDITION 5.

l ACTION:

l With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS and insert all insertable l

control rods.

l SURVEILLANCE REQUIREMENTS I

4.9.2 Each of the above required SRM channels shall be demonstrated OPERABLE by:

a.

At least once per 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />s:

1.

Performance of a CHANNEL CHECK, l

2.

Verifying the detectors are inserted to the normal operating level, and l

3.

During CORE ALTERATIONS, verifying that the detector of an OPER-ABLE SRM channel is located in the core quadrant where CORE l

ALTERATIONS are being performed and another is located in an

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adjacentquadrant.

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L "The use of special movable detectors during CORE ALTERATIONS in place of the l

normal SRM nuclear detectors is permissible as long as these special detectors are connected to the normal SRM circuits.

l

1. ot required for control rods removed per Specification 3.9.10.1 and 3.9.10.2.

N 1

GRAND GULF-UNIT 1 3/4 9-3 Amendment No.35 l 1

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REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued) b.

Performance of a CHANNEL FUNCTIONAL TEST:

1.

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the start of CORE ALTERATIONS, and 2.

At least once per 7 days.

Verifying that the channel count rate is at least 0.7 cps":

c.

1.

Prior to control rod withdrawal, 2.

Prior to and at least.once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during CORE ALTERATIONS, and 3.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, l

except that:

1.

During spiral unloading, the required count rate may be permitted to be less than 0.7 cps *,

2.

Prior to and during spiral loading, until sufficient fuel has been loaded to maintain at least 0.7 cps *, the required count rate may be achieved by:

a)

Use of portable external source, or b)

Loading up to 2 fuel assemblies " in cells containing inserted control rods around an SRM.

d.

Verifying that the RPS circuitry "shorting links" have been removed within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during:

1.

The time any control rod is withdrawn," or 2.

Shutdown margin demonstrations.

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• Provided signal to noise ratio >2; otherwise use 3 cps.

"Not required 'or control rods removed per Specification 3.9.10.1 or 3.9.10.2.

i These fuel assemblies may be loaded with the SRM count rate less than 0.7 cps.

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GRAND GULF-UNIT l 3/4 9-4

I REFUELING OPERATIONS 3/4.9.5 COMMUNICATIONS LIMITING CONDITION FOR OPERATION i

3.9.5 Direct communication shall be maintained between the control room and refueling platform personnel.

APPLICABILITY:

OPERATIONAL CONDITION 5, during CORE ALTERATIONS.*

ACTION:

l When direct communication between the control room and refueling platform personnel cannot be maintained, immediately suspend CORE ALTERATIONS.*

SURVEILLANCE REQUIREMENTS 4.9.5 Direct communication between the control room and refueling platform personnel shall be demonstrated within one hour prior to the start of and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during CORE ALTERATIONS.*

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"Except movement of control rods with their normal drive system.

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l GRAND GULF-UNIT 1 3/4 9-7 Amendment No. 35l

REFUELING OPERATIONS 3/4.9.6 REFUELING EQUIPMENT REFUELING PLATFORM LIMITING CONDITION FOR OPERATION 3.9.6.1 The refueling platform shall be OPERABLE and only the main hoist shall be used for handling fuel assemblies.

APPLICABILITY:

During handling of fuel a uemblies or control rods in the primary containment with the refueling platform.

ACTION:

With the requirements for refueling platform OPERABILITY not satisfied, suspend use of any inoperable refueling platform equipment from operations involving the hanJ1ing of fuel assemblies or control rods after placing the load in a safe condition.

SURVEILLANCE REQUIREMENTS 4.9.6.1 Each refueling platform hoist to be used for handling fuel assemblies or control rods shall be demonstrated OPERABLE within 7 days prior to the handling of fuel assemblies or control rods:

a.

In the containment fuel pool, reactor cavity or reactor pressure vessel by:

1.

Demonstrating operation of the slack cable cutoff on the main hoist when the total cable load is 50110 pounds.

2.

Demonstrating operation of the grapple engaged loaded interlock on the main hoist before the total cable load exceeds 535 pounds.

3.

Demonstrating operation of the jam cutoff on the main hoist before the total cable load exceeds 1250 pounds.

4.

Demonstrating operation of primary and redundant overload cutoff on the auxiliary hoists before the load exceeds 550 pounds, b.

In or over the reactor pressure vessel by:

1.

Demonstrating operation of the downtravel cutoff on the main hoist when the bottom of the grapple is 3.5 i 0.5 inches below the top of the fuel assembly handles in the reactor core.

2.

Demonstrating operation of the primary and redundant fuel load interlocks on the main hoist before the total cable load exceeds 600 pounds.

GRAND GULF-UNIT 1 3/4 9-8

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PLANT SYSTEMS BASES 3/4.7.4 SNUBBERS (Continued)

The acceptance criteria are to be used in the visual inspection to determine OPERABILITY of the snubbers.

For example, if a fluid port of a hydraulic snubber is found to be uncovered, the snubber shall be declared inoperable and shall not be determined OPERABLE via functional testing.

To provide assurance of snubber functional reliability one of three functional testing methods is used with the stated acceptance criteria:

1.

Functionally test 10% of a type of snubber with an additional 5%

l tested for each functional testing failure, or 2.

Functionally test a sample size and determine sample acceptance or rejection using Figure 4.7.4-1, or 3.

Functionally test a representative sample size and detensine sample acceptance or re,iection using the stated equation.

Figure 4.7.4-1 was developed using "Wald's Sequential Probability Ratio Plan" described in "Quality Control and Industrial Statistics" by Acheson J. Duncan.

Permanent or other exemptions from the surveillance program for individual snubbers may be granted by the Comission if a justifiable basis for exemption is presented and, if applicable, snubber life destructive testing was performed to qualify the snubbers for the applicable design conditions at either the completion of their fabrication or at a subsequent date.

Snubbers so exempted shall be listed in the list of individual snubbers indicating the extent of the exemptions.

The service life of a snubber is established via manufacturer input and information through consideration of the snubber service conditions and asso-ciated installation and maintenance records (newly installed snubber, seal replaced, spring replaced, in high radiation area, in high temperature area, etc.).

The requirement to monitor the snubber service life'is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions.

These records will provide statistical bases for future consideration of snubber service life.

3/4.7.5 SEALED SOURCE CONTAMINATION The limitation on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium.

This limitation will ensure that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values.

Sealed sources are classified into three groups according to their use, with surveillance requirements commensurate with the probability of damage to a source in that grew.

Those sources which are frequently handled are required to be tested more v ien than those which are not.

Sealed sources which are continuously enclosed within a shielded mechanism, i.e., sealed sources within radiation monitoring or boron measuring devices, are considered to be stored and need not be tested unless they are removed from the shielded mechanism.

GRAND GULF-UNIT 1 B 3/4 7-3 Amendment No.35 l

PLANT SYSTEMS BASES i

3/4.7.6 FIRE SUPPRESSION SYSTEMS l

The OPERABILITY of the fire suppression systems ensures that adequate fire suppression capability is available to confine and extinguish fires occurring in any portion of the facility where safety-related equipment is located.

The fire suppression system consists of the water system, spray and/or sprinklers, CO systems, halon systems and fire hose stations.

The collective capability 2

of the fire suppression systems is adequate to minimize potential damage to safety-related equipment and is a major element in the facility fire protection program.

In the event that portions of the fire suppression systems are inoperable, alternate backup fire fighting equipment is required to be made available in the affected areas until the inoperable equipment is restored to service.

When the inoperable fire fighting equipment is intended for use as a backup means of fire suppression, a longer period of time is allowed to provide an alternate

~ means of fire fighting than if the inoperable equipment is the primary means of fire suppression.

j The surveillance requirements provide assurances that the minimum OPERABILITY requirements of the fire suppression systems are met.

An allowance is made for ensuring a sufficient volume of halon in the halon storage tanks by verifying the weight and pressure of the tanks.

In the event the fire suppression water system becomes inoperable, immediate corrective measures must be taken since this system provides the major fire suppression capability of the plant.

The surveillance requirements for spray and sprinkler systems provide for periodic visual inspections to ensure that temporary structures / objects do not impair the spray patterns which have been established in accordance with the GGNS fire protection design requirements.

3/4.7.7 FIRE RATED ASSEMBLIES The OPERABILITY of the fire barriers and barrier penetrations ensure that fire damage will be limited.

These design features minimize the possibility of a single fire involving more than one fire area prior to detection and extinguishment.

The fire barriers, fire barrier penetrations for conduits, cable trays and piping, fire windows, fire dampers, and fire doors are periodically inspected to verify their OPERABILITY.

3/4.7.8 AREA TEMPERATURE MONITORING The area temperature limitations ensure that safety-related equipment will not be subjected to temperatures in excess of their environmental qualification temperatures.

Exposure to excessive temperatures may degrade equipment and can cause loss of its OPERABILITY, The temperature limits include allowance for instrument error.

GRAND GULF-UNIT 1 B 3/4 7-4 l

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UNITED STATES

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[

g NUCLEAR REGULATORY COMMISSION t

l WASHINGTON, D. C. 20555

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September 10, 1987 Docket No.: 50-416 Mr. Oliver D. Kingsley, Jr.

Vice President, Nuclear Operations System Energy Resources, Inc.

Post Office Box 23054 Jackson, Mississippi 39205 N

SUBJECT:

CHANGES TO TECHNICA SPECIFICATIONS REGARDING CORE ALTERATIONS AND SNUBBER SAMPLE SIZE (TAC NO. 65511)

RE: GRAND GULF NUCLEAR STATIO UNIT 1 The Comission has issued the enclosed Ameiidment No. 35io Facility Operating

~

License No. NPF-29 for the Grand Gulf Nuclear Station,- Unit 1 This amendment consists of changes to the Technical Specifications (TSs) in response to your i

appiteation dated June 3, 1987, as supplemented June 22, 1987.

This amendment changes the definition of core alteration in the TSs to include certain exceptions and changes footnotes in the TSs to be consistent with the new definition. This amendment also changes a snubber surveillance test sample plan in the TSs by decreasing the number of additional snubbers required to be tested # rom 10% to 5% for each snubber in the initial test sample that fails to meet specified functional test criteria, A copy of the Sa'ety Evaluation is also enclosed. The Notice of Issuance will be included in the Comission's bi-weekly Federal Register notice.

Sincerely, rL h W AMA ;

Lester L. Kintner, Project Manager Project Directorate !!-2 Division of ReMtor Projects-!/II Office of Nuclear Peactor Regulation

Enclosures:

1.

Amendment No. 25 to NPF-29 2.

Safety Cvaluation ec w/ enclosures:

See next page 3 7 0 9 8 8 M (jf-

Mr. Oliver D. Kingsley, Jr.

System Energy Resources, Inc.

Grand Gulf Nuclear Station (GGNS) cc:

Mr. Ted H. Cloninger Mr. C. R. Hutchinson Vice President, Nuclear Engineering GGNS General Manager and Support System Energy Resources, Inc.

System Energy Resources, Inc.

Post Office Box 756 Post Office Box 23054 Port Gibson, Mississippi 39150 Jackson, Mississippi 39205 Robert B. McGehee, Esquire The Honorable William J. Guste, Jr.

Wise, Carter, Child, Steen and Caraway Attorney General P.O. Box 651 Department of Justice Jackson, Mississippi 39205 State of Louisiana Baton Rouge, Louisiana 70804 Nicholas S. Reynolds Esquire Bishop. Libeman, Cook, Purcell Office of the Governor and Reynolds State of Mississippi 1200 17th Street, N.W.

Jackson, Mississippi 39201 Washington, D. C.

20036 Attorney General Mr. Ralph T. Lally Gartin Building Manager of Quality Assurance Jackson, Mississippi 39205 Middle South Utilities System Services, Inc.

P.O. Box 61000 Mr. Jack McMillan Director New Orleans, Louisiana 70161 Division of Solid Waste Managerent Mississippi Department of Natural Mr. John G. Cesare Resources Director, Nuclear Licensing and Safety (creau of Pollution Control System Energy Resources, Inc.

Post Office Box 10385 P.O. Box 23054 Jackson, Mississippi 39?09 Jackson, Mississippi 39205 Alton B. Cobb, M.D.

Mr. R. W. Jackson, Project Engineer State Health Officer Bechtel Power Corporation State Board of Health 15740 Shady Grove Road P.O. Box 1700 i

Gaithersburg, Maryland 20877-1454 Jackson, Mississippi 39205 Mr. Ross C. Butcher President Senior Resident Inspector Claiborne County Board of Supervisors i

U.S. Nuclear Regulatory Comission Port Gibson, Mississippi 39150 Route 2, Box 399 Port Gibson, Mississippi 39150 Regional Administrator, Pegion !!

U.S. Nuclear Regulatory Comission 101 Marietta Street, N.W., Suite 2900 Atlanta, Georgia 30323 Mr. James E. Cross i

GGNS Site Director l

System Energy Resources, Inc.

P.O. Box 756 Port Gibson, Mississippi 39150 i

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UNITED STATES e-

-* *t NUCLEAR REGULATORY COMMISSION h*

.y WASHINGTON, D. C. 20555 e

\\,,,,,/

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 35 TO FACILITY OPERATING LICENSE NO. NPF-29 MISSISSIPPI POWER 8 LIGHT COMPANY SYSTEM ENERGY RESOURCES, INC.

SOUTH MISSISSIPPI ELECTRIC POWER ASSOCIATION GRAND GULF NUCLEAR STATION, UNIT 1 DOCKET NO. 50 416 j

INTRODUCTION By letter dated June 3,1987, as supplemented June 22, 1987, System Energy Resources, Inc., (the licensee) requested an amendment to Facility Operating License No. NPF-29 for the Grand Gulf Nuclear Station, Unit 1 (GGNS-1). The proposed amendment would (1) change the definition of core alteration in the i

Technical Specifications (TSs) to include certain exceptions and change foot-notes in the TSs to be consistent with the new definition; and (2) change a i

snubber surveillance test sample plan in the TSs by decreasing from 10% to 5%

the number of additional snubbers required to be tested for each snubber in the initial test sample that fails to meet specified functional test criteria.

EVALUATION (1) Definition of Core Alteration The following changes to the TSs would be made:

a.

The definition of core alteration would be modified to exclude normal novement of the source range monitors (SRMs), intermediate range monitors IRNs, local power monitors (LPRMs), traversing in-core probes TIPS or special movable detectors, b.

The "*" footnote to Specification 3.1.1 on shutdown margin would be deleted. This footnote provides an exception to the core alteration definition for movement of IRMs, SRMs or special movable detectors, c.

The "*" footnote to Surveillance Requirement 4.1.3.2.a would be modified by deleting the exception to the core alteration definition for the movement of SRMs, IRMs or special movable detectors. The exception for normal control rod movement remains and is not j

affected by this proposed change.

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2 d.

The "*" footnote to Table 3.3.1-1 would be modified by deleting the exceptions to the core alteration definition for IRMs, SRMs or special movable detectors.

The part of the "*" footnote requiring operable SRM instrumentation for replacement of LPRM strings would be retained.

e.

The "**" footnote to Specification 3.9.2 on refueling operations instrumentation would be deleted.

This footnote provides an exception to the core alteration definition for movenent of IRMs, SRMs, or special movable detectors.

f.

The "*" footnote to Specification 3.9.5 would be modified by deleting the exception to the core alteration definition for incore instrumen-tation. The part of the "*" footnote that allows an exception for con-trol rod movement with their nomal drive system remains and is not affected by this proposed change.

The present definition of core alteration is:

"Core alteration shall be the addition, removal, relocation or novement of fuel, sources, incore instruments or reactivity controls within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of core alterations shall not preclude completion of the movement of a component to a safe conservative position."

The proposed change would insert the following after the first sentence:

"Nomal movement of the SRMs, IRMs, LPRMs, TIPS, or special movable detectors is not considered a core alteration."

The exception to the present definition of core alteration for the nomal movement of the SRMs, IRMs LPRFs, TIPS, and special movable detectors is needed in certain specifications related to refueling operations in order to preclude unnecessary suspension of the nomal movement of these detectors.

During a refueling outage, maintenance or modification of equipment can result in TS limiting conditions for operation which require that core alterations be suspended.

In the present TSs, exceptions to the definition of core altera-tion for nonnal movement of detectors are provided by footnotes in those TSs where a need for the exception was foreseen.

However, some TSs that recaire suspension of core alterations do not presently have a footnote excepting normal movement of detectors.

For exam)1e, Specification 3.8.1.2 requires suspension of core alterations wit 1 diesel generator 11 or 12 inoperable. With the present TSs, surveillance tests of SRMS and IRMs could not be perfonned because the tests require movement of the detectors. Making the exception a part of the definition will correct this type of operational problem. Where particular conditions are required for nomal movement of detectors, these conditions are retained in the applicable TSs.

For example, the requirement for SRMs to be operable when replacing LPRMs is retained in Specification 3/4.3.1, "Reactor Protection System Instrumentation."

i

,. The NRC staff has reviewed the proposed changes to the GGNS-1 TSs related to core alterations. The detectors in the SRM, IRM, LPRM, TIP and the special movable detectors are sealed unit fission detectors and their reactivity worth is insignificant with respect to reactivity excursion events. There' ore, allowing the nomal movement of these detectors will not significantly increase the probability or consequences of an accident previously analyzed in the Final Safety Analysis Report.

The proposed change would only permit nomal movement of the incore detectors.

Nomal movement of these detectors includes insertion anti withdrawal using detector drives, replacement of detectors, and movement of special movable detectors in the core region. The addition, removal or relocation of SRMs, IRMs LPRMs and TIPS would still be prohibited.

The staff concludes that the proposed enanges to the definition of core alteration i

and the deletion of footnotes in the TSs would not significantly reduce the level of safety and would tend to enhance safety by making the TSs more readable. Accordingly, the proposed changes are acceptable.

(2) Snubber Sample Plan To verify the operability of safety-related snubbers, Surveillance Pequirement 4.7.4.e in the TSs requires functional testing to be perfomed on a periodic basis. The TSs pemit the use of any one of three specified sampling plans.

Essentially, all three plans require the testing of an initial sample of snubbers from the total population.

For every inoperable snubber identified during testing of an initial sample of snubbers, an additional or subsequential sample is required to be tested.

For Sample Plan 1, the size of the initial and the subsecuential samples is 10% and 10%, respectively. The initial sample size of 10% for Sample Plan 1 was selected on the basis that every snubber in the plant will be tested at least once every 15 years when the associated functional testing period is 18 months. The subsequential sample size of 10% was selected as a conservative value.

For Sample Plans 2 and 3, initial and subsequential sample sizes are both detemined by statistical considerations, and the subsequential samples are half that of the initial samples. All three sample plans should yield the same resul ts. Yet for a population that would produce the same initial sarple size fnr Sample Plans 1 and 2 or 1 and 3, the subsequential sample sizes will differ i

by twice as much. To make all three plans have an equal basis, the conservatively detemined subsequential size of 10% for Sample Plan 1 should be reduced to 5%.

The American Society of Mechanical Engineers Operation and Maintenance Working Group 4 Standard (0&M 4 Standard), "Examination and Perfomance Testing of Nuclear Power Plant Dynamic Restraints (Snubbers)." has taken this into consideration and changed the recomended subsequential sample size from 10%

to 5% for Sample Plan 1.

The standard was approved by the NRC staff and will be adopted by ASME Roller & Pressure Yessel Code Section XI for plant surveillance guidance,

e P In conclusion, the proposed change to Sample Plan I would make it consistent with the other two sample plans in the TSs, is in accordance with the requirements recommended by the OAM 4 Standard, and is therefore acceptable, j

ENVIRONMENTAL CONSIDERATION This amendment involves a change to a requirement with respect to the inste11ation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes to the surycillance f

requirements. The staff has determined that the anendment involves no t

significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite and that there is no signifi-cant increase in individual or cumulative occupational radiation exposure.

The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding. Accordingly, this amendment meets the eligibility l

criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b), no envircnmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.

CONCLUSION The staff has concluded, based on the considerations discussed above, that:

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and the security nor to the health and safety of the public.

Dated: September 10, 1987 Principal Contributors:

W. Brooks, Reactor Systems Branch, DEST H. Shaw, Mechanical Engineering Branch, DEST I

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