Text

Proposed Changes to STS Re Reactivity Control Sys
	ML20050B689
Person / Time
Site:	Grand Gulf
Issue date:	12/18/1981
From:	; MISSISSIPPI POWER & LIGHT CO.
To:	;
Shared Package
ML20050B688	List: ML20050B687; ML20050B689;
References
	NUDOCS 8204070214
	Download: ML20050B689 (30)
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REACTIVITY CONTROL SYSTEMS

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"d LIMITING CONDITION FOR OPERATION (Continued)

ACTION (Continued)

-2.-If the-inoperable control rod (s)-is-inserted;-within-cr,e-heur-

-disarm-the-associated-directional control-valvesaa-either:

4 Electrically, er bb-Hy&=1-ical4y uy vmis m.

uri e etwi and ed,= t t: m l

-iselatioit-valves.

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

l With more than 8 control rods inoperable, be in at least L T SHUTDOWN within c.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The scram discharge volume drain and vent valves shall be

.ITDr demonstrated OPERABLE by:

At least once per 31 days verifying each valve to be open,* and a.

b.

At least once per 92 days cycling each valve through.at least one complete cycle of full travel.

4.1.3.1.2 When above the low power setpoint of the RPCS, all withdrawn control rods not required to have their directional control valves disarmed electrically or hydraulically shall be demonstrated OPERABLE by moving each control rod at least one notch:

a.

At least once per 7 days, and b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when any control rod is immovabl,e.as_ #

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result of excessive friction or mechanical interfetencgfM a Ltv!!y sm eated,

4.1.3.1.3 All control rods shall be demonstrated OPERABLETyTrformfrife of Surveillance Requirements 4.1.3.2, 4.1.3.3, 4.1.3.4 and 4.1.3.5.

These valves may be closed intermittently for testing under administrative

controls,

.IMay-be-rearmed-intermittent-ly,-under-edministrative-coatroly-to permit-

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_tx H an-...or4.* " d '-restoring-the-control-rod-to-OPERABLE-status.

8204070214 820405 PDR ADOCK 05000416 A

PDR EEC,18 1981 GRAND GULF-UNIT 1 3/4 1-4

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3/4.1.3 CONTROL RODS CONTROL ROD OPERABILITY LIMITING CONDITION FOR OPERATION 3.1.3.1 All control rods shall be OPERABLE.

APPLICABILITY:

OPERATOSLCONDITIONS1and4.

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ACTION:

poi fu/ly mseated ad With one control rod inoperable due to being immovable, as a result of a.

excessive friction or mechanical interference, or known to be untrippable:

1.

Within one hour:

a)

Verify that the inoperable control rod, if withdrawn, is separated from all other inoperable control rods by at least two control cells in all directions.

b)

Disarm the associated directional control valves ** either:

1), Electrically, or 2)

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

c)

Comply with Surveillance Requirement 4.1.1.c.

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

l 2.

Restore the inoperable control rod to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months

l or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

With one or more control rods trippable but inoperable for causes other than addressed in ACTION a, above:

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If the inoperable control rod (

h _ithdr=.h, ithin one houg %egfd z"

verify:

g a)

That the inoperable withdrawn control ro s is separated tro & a all other inoperable control rods by at least two control cells in all directions,-and-

-b)---The-insert-ion 4apabil4ty-of-the-inoperable-withdenn centtel-l

-rod (+)-by-inserting-the-control-rod (+)-at-least ene retch-by.

.dr4ve-water-pressure-within-the normaLoperating-range *-

Otherwise, insert the inopecable withdrawn control od(s) an disarm l

the associated directional control valves ** either:

7g,

a)

Electrically, or b)

Hydraulically by closing the drive water and exhaust water isolatio.1 valves.

x ;-.2 Mhe-inoperable-control--rod-may,then-be withdrawn i.o a posit-ion ~ne further-.

w.-

withdrawn-than-its-position when 'fcund to be -inop'erable--

- May be rearmed intermittently under administrative control to permit testing associated with restoring the control rod to'0PERABLE s,tatus. _

GRAND GULF-UNIT 1 3/4 1-3

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. to AECM-82/115

. Grand Gulf Unit 1 Technical ~ Specifications Page 3/4 4-1, dated

, December 18, 1981 1)

Specification 3.4.1.1:

The action. statement for the Recirculation System limiting condition for operation does not allow for continued operation after recovery of an inoperative recirculation loop. This change is consistent with Hatch 2 approved technical

-specifications. The change to 3.4.1.1.b has been'added for clarity.

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..., J.... '. i.: , s u i' i 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation.

APPLICABILITY:

OPERATIONAL CONDITIONS 1* and 2*

opga(mu pray cowb^wej kshe ACTION:

hopf loofs b yesa% 4W '*

hems on a.

With one reactor coolant system recirculation oop not in operation, be in at least, HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

With no reactor coolant system recirculation loops in operation,Yplace the reactor mode switch in the Shutdown position.

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4.4.1.1 Each reactor coolant system recirculation loop flow control valve shall be demonstrated OPERABLE at least once per 18 months by.

Verifying that the control valve fails "as it"~ on loss of hydraulic a.

pressure at the hydraulic unit, and l

b.

Verifying that the average rate of control valve movement is:

1.

Less than or equal to 11% of stroke per second opening, and l

2.

Less than or equal to 11% of stroke per second closing."

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"See Special Test Exception 3.10.4.

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i DEC,181981 GRAND GULF-UNIT 1 3/4 4-1

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. to AECM-82/115 Grand Gulf Unit 1 Technical Specifications Page 3/4 8-3,. dated-March 17, 1982 l

1)

Specification 4.8.1.1.2.a.5:

This item is changed.to allow

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ELECTRICAL POWER SYSTEMS SUf

'LLANCE REQUIREMENTS SbkW 3 We /RM-82l 4.8.1.1.1 Each of the above required independent circ 0fts between the offsite transmission network and the onsite Class,1E distribution system shall be:

a.

Deterr'ned OPERABLE at least once per 7 days by verifying correct breakr alignments and indicated power availability, and b.

Demonstrated OPERABLE at least once per 18 months during shutdown by transferring, manually and. automatically, unit power supply from the normal circuit to the alternate circuit.

4.8.1.1.2 Each of the above required dies.el generators shall be demonstrated OPERABLE:

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a.

In accordance with the frequency specified in Tabic (.8.1.1.2-1 on a STAGGERED TEST BASIS by:

1.

Verifying.the fuel level in the day tar..,..

2.-

Verifying the fuel level in the fuel storage tank.

3..

Verifying the fuel transfer pump starts and transfers fuel from the storage system to the day tank.

4.

Verifying the diesel starts from ambient condition and accelerates to at least 441 rpm for diesel generators 11 and 12 and 882 rpm for diesel generator 13 in less than or equal to 10 seconds.. The

~ generator voltage and frequency shall be 4160 i 416 volts a..)

60 1 1.2 Hz within 13 seconds after the start signal.

The diesel generator shall be started for this test by using one of the following signals:

a)

Manual, l

b)

Simulated loss of offsite power by itself.

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c)

Simulated loss of offsite power in conjunction with'an ESF actuation test signal.

d)

An ESF actuation test signa itself.

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5.. Verifying the diesel generator is chronize 1

ded to greater than or equal to 3500 kW for die e generators 1 and 12 and 1650 kW'for diesel. generator 13 in less than or equal to 60 seconds, ago erates with these" loads for at least'60 ' minutes.

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6.

fying the diesel generator is aligned to provide standby power to the associated emergency busses.

7.

Verifying' the pressure in all diesel generator air start-receivers to be greater than or equal to:

a) 160 psig for diesel generator 11 and 12, and b) 175 psig for diesel generator 13.

b.

At least once per 31 days and after each operation of the diesel where the period of operation was greater than or equal to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months by checking for and removing accumulated water from the day fuel tanks; GRAND GULF-UNIT 1 3/4 8-3

- to AECM-82/115 Grand Gulf Unit'l Technical Specifications Page 3/4 8-4,. dated March 17, 1982 1)

Specification 4.8.1.1.2.c:

This insolubles test is not practical and MP&L has obtained vendor release from the requirements of an insolubles test; additionally,_MP&L FSAR commitments concerning diesel oil are contained in responses to Questions 40.44 and 40.45.

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THIS PAGE OPE ^! PEi!DEU RECEIPT OF

'INFORMATLi. ROM Tile A??MAMT P Pi D D F B H K 1 C O P Y ELEC"DT N POWER SYSTIMS._

//M [ c I e M [ o MMJ SURVEILLANCE REQUIREMENTS (Continued)

At least once per 92 days and from new fuel. oil prior to addition to c.

the storage tanks by verifying that a sample obtained in accordance with ASTM-0270-1975 has a water and sediment coatent of less than or equal to.05 volume percent and a kinematic viscosity @ 40*C of greater than or equal to 1.9 but less than or equal to 4.1 when tested in accordance with ASTM-D3 E-77,- ad :.. i...gity levci of less the,. 2 m;;. -

sf in&bM ;;;r 100...l.

he. te;ted ir, ::::rd:rc: with ^.ST"-02271-70.-

d.

At least once per 18 months, during shutdown, by:

1.

Subjecting the diesel to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for this class of standby service.

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2.

Ver.ifying the diesel generator capability to reject a load of greater than or equal to 1735 kW for diesel generator 11, greater than or equal 'to 890 kW for diesel generator 12, and greater than or equal to 2780 kW for diesel generator 13 while maintaining (voltage at 4160 t 416 volts and frequency at 60 i 1.2 Hz) (less than or equal to 75% of the difference between nominal speed and the overspeed 4. rip setpoint, or 15% above nominal, whichever is less.)

3.

Verifying the diesel generator capability to reject a load of 7000 kW for diesel generators 11 and 12 and 3300 kW for diesel

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generator 13 without tripping.

The generator voltage shall not exceed 4576 volts during and following the load rejection.

4.

Simulating a loss of offsite power by itself, and:

a)

For Divisions 1 and 2:

1)

Verifying deenergization of the emergency busses and load shedding from the emergency busses.

2)

Verifying the diesel generator starts on the auto-start signal, energizes the emergency busses with permanently l

connected loads within 13 seconds, energizes the auto-l connected shutdown loads through the load sequencer i

and operates for greater than or equal to 5 minutes while its generator is loaded with the shutdown loads.

After energization, the steady state voltage and frequency of the emergency busses shall be maintained at 4160 1 416 volts and 60 1.2 Hz during this test.

b)

For Division 3:

1)

Verifying de-energization of the emergency bus.

2)

Verifying the diesel generator starts on the auto-start signal, energizes the emergency bus with the loads within 13 seconds *and operates for greater.than or equal to 5 minutes while its generator is loaded with the shutdown loads.

After energization, the steady state voltage and frequency of the emergency bus shall be maintained at 4160 1 416 volts and 60 1 1.2 Hz l

during this test.

MAR 17 1982 GRAND GULF-UNIT 1 3/4 8-4

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Attach:ent 5 to AECM-82/115 Grand Gulf Unit 1 Technical Specifications Page'3/4 1-19, dated February 1, 1982 1)

Specification 4.1.5.c.4:

It is suggested that this item be deleted since the Standby Liquid Control System is entirely located within the Mark III containment. Additionally, Specification 4.1.5.a.3 discusses operability of the heat tracing; and, with concerns for ALARA and handling of toxic material within confined spaces this demonstration is not necessary, i

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I REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b.

At least once per 31 days by; 1.

Starting both pumps and recirculating demineralized water to the test tank.

2.

Verifying the continuity of the explosive charge.

3.

Determining that the available weight of sodium pentaborate is greater than or equal to 5500 lbs and the concentration of boron in solution is within the limits of Figure 3.1.5-1 by chemical analysis.*

4.

Verifying that each valve, manual, power operateu or automatic, in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

At least once per 18 month's during shutdown by; c.

1.

Initiating one of the standby liquid control system loops,

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including an explosive valve, and verifying that a flow path from the pumps to the reactor pressure vessel is available by pumping demineralized water into the reactor vessel.

The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch which has been certified by having one of that batch success-fully fired.

Both 'njection loops shall be tested i.n 36 months.

2.

Demonstrating that, when tested pursuant to Specification 4.0.5, the minimum flow requirement of 41.2 gpm at a pressure of greater than or equal to 1220 psig is met.

3.

Demonstrating tha't the pump relief valve setpoint is' less l

than or equal to 1386 psig and verifying that the relief valve does not actuate during recirculation to the test tank.

4.

-**De enstrating that ell hcot traccd piping bett!cer th: terege tank and the reacLur ve5sc1 is unblecked by pumping fic, the stetage tank tO the. test tank and ther draining end fic5hing.

the-. piping wi th uemineroi s tcd water.

5.

Demonstrating that the storage tank heaters are OPERABLE by verifying the expected temperature rise for the sodium pentaborate solution in the storage tank after the heaters are energized.

"This test shall also be performed anytime water or boron is added to the solution or when the solution temperature drops below the limit of Figure 3.1.5-1.

- Ihis te51 5 hall Gl;c be pe"fe""ed 'SCnever 'ulh huot tracing Ci"cnite havn _

u baa found te be inoperable and m;y bc perivo. icd by any series vf seqccatial, oV.Erlapping er 10t61 fluw pdi.h sleps such T. hat Lne ent,1re f low pai.h Is-daciuded:-

l GRAND GULF-UNIT 1 3/4 l-19

. to AECM-82/ll5 Grand Gulf Unit 1 Technical Specifications Pages 6-22 & 6-23, dated January 5, 1982.

1)

Specification 6.12.1 & 6.12.2:

Paragraph 6.12.1 contains a statement that requires a radiation work permit.when the intensity of the radiation is greater than 100 mrem /hr. An exemption to health physics personnel and personnel who are escorted by health physics personnel is given in.the footnote to this paragraph. Due to the operational surveillance activity in high radiation areas it is necessary for certain personnel to enter these areas without health physics escort. Normally to cover the administrative requirements, a blanket radiation work permit is issued to these personnel for periods up to 31 days. This procedure is self defeating in our opinion.

Let us look closer at Radiation Work Permit philosophy.

The Radiation Work Permit provides management with appropriate controls to assure safe radiological work practices in hazardous environments.

A common problem in a nuclear power plant is one of personnel becoming too casual about radiological environments. Management must have a mechanism by which the hazards of working environments are pricritized. Radiation Work Permits can play an essential role in emphasizing to the worker that a higher normal level of hazard exists.

Two types of abuses can. detract from the usefulness of the RWP in prioritizing radiological hazards:

a)

Overuse due to excessive use in low environments or excessive use due to the presence of too many hazards (i.e., sloppy housekeeping and poor contamination control).

b)

The use of " blanket" or " indefinitely extended" RWPs such~that a worker always feels he is on an RWP and therefore exempt from hazard.

A solution to this problem is to (1) administrative 1y establish a limit to the life of an RWP to one working shift with the ability to extend to a maximum of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months if needed (i.e., no extended RWPs), (2) to train certain individuals as pre-approved (based on need) by the RPM to a higher level of radiological expertise (i.e... Rad Worker III or "self monitor"). lie may then enter higher radiation areas (greater than 100mR per hour but Icss than 1000 mR per hour) and/or contamination areas (10,000 dpm to 50,000 dpm) for surveillances or operational activities without an RWP.

A Rad Worker III may not enter "Very High Radiation Areas", " Airborne Areas", or " Potential Airborne Areas" (i.e., greater than 50,000 dpm) without an RWP.

These measures will enable us to maintain worker awareness of the unusually hazardous environment in which RWP work is being done.

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It is therefore proposed that we amend the tech specs to allow an additional level of radiation worker training and control over high radiation area entry. The wording of the proposed amendment may take the form of an additional paragraph d:

i d.

For each high radiation area in which the intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr, specific exemption from a radiation work permit may be made by the Unit Health Physicist, provided that the individual receives specific training for the purpose of entering high radiation areas without a Health Physics escort.

The scope and content of such training shall be under the administrative control of the Unit Health Physicist.

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ADMINISTRATIVE CONTROLS

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RECORD RETENTION (Continued)

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..a d.

Records of gaseous and liquid radioactive mat.erial released to the environs.

. yf.

Records of transient or operational cycles for those unit components e.

identified in Table 5.7.1-1.

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f.

Records of eactor tests and experiments.

g.

Records of training and qualification for current members of the unit' staff.

h.

Records of in-service inspections performed pursuant to these Technical Specifications.

i.

Records of Quality Assurance activities required by the Operational Quality Assurance Manual.

j.

Records of reviews performed for' changes made to procedures or equipment or reviews of, tests and experiments pursuant to 10 CFR 50.59.

k.

Records of meetings of the PSRC and the SRC.

1.

Records'of the service lives of all hydraulic and mechanical snubbers listed on Tables 3.7.5-1 and 3.7.5-2 including the date at which the service life commences and associated installation and maintenance records.

Records of analyses required by the radiological environmental m.

monitoring program.

6.11 RADIMION PROTECTION PROGRAM 6.11.1 Procedures for perspnnel radiation protection shafl be prepared' consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.

6.12 HIGH RADIATION AREA

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6.12.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each hicjh radiation area in which the intensity of radiation.is greater than 100 mrem /hr but le'ss than 1000 mrem /hr shall be barri-caded and conspicuously posted as a high radiatio'n area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP).*

Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more, of the following:

A radiation monitoring device which continuously indicates the a.

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radiation dose rate in the area.

" Health Physics personnel 'or persanngl"estatt'ed by Healtfi $ysic's personn'e1.,

shall be exempt from t'he RWP issuance iequirement during the performance of-T their assignedWidiergretht-fer. duties, 'pr~ovided they are otherwise

.following plant radiation protection rocedur s for ent into high radiation areas.

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9 Attechment'7 to AECM-82/113 Grand Gulf Unit-1 Technical Specifications-Page 6-8, dated January 5, 1982 1)

Specification 6.5.1.6.a:

In accordance with the clarification provided by Bob Benedict of. Licensee Qualification Branch, MP&L is taking the position that the PSRC shall not have to review all procedures and programs required _by Specification 6.8; but, rather that the PSRC will be responsible to see that the procedures and

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programs required by Specification 6.8 are reviewed.

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8 ADMINISTRATIVE CONTROLS

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RESPONSIBILITIES 6.5.1.6 The PSRC shall be responsible for:

Review of (1) all procedures required by Specification 6.8 and changes a.

thereto, (2) all programs required by Specification 6.8 and changes thereto, and (3) any other proposed procedures or changes thereto as determined by the Plant Manager to affect nuclear safety.

b.

Review of all proposed -tests and experiments that affect nuclear safety.

Review of all proposed changes to Appendix "A" Technical Specifications.

c.

d.

Review of all proposed changes or modifications to unit systems or equipment that affect nuclear safety.

Investigation of all violations of the Technical Specifications e.

including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Senior

. Vice President - N6 clear and to the Safety Review committee (SRC).

f.

Review of' events requiring 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months written notification.to the Commission.

g.

Review cf unit operations to detect potential nuclear safety hazards.

h.

Performance of special reviews, investigations or analyses and reports thereon as requested by the Plant Manager or the SRC.

t i.

Review of the Security' Plan and implementing procedures and shall submit recommended changes to the SRC.

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j.

Review of the Emergency Plan and implementing procedures and shall i

submit recommended changes to the SRC.

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k.

Review of every unplanned onsite release of radioactive material to the environs including the preparation and forwarding of reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence tc the Plant Manager and to the SRC.

l.

Review of changes to the PROCESS CONTPOL PROGRAM, DFFSITE DOSE l

CALCULATION MANUAL, and radwaste systems.

AUTHORITY 6.5.1.7 The PSRC shall:

a.

Recommend in writing tu the Plant Manager approval or disapproval of items considered under 6.5.1.6(a)'through (d) above.

. b.

Render determinations in writing with regard to whether or not each item considered.under 6.5.1.6(a) through (e) above constitutes an l

l unreviewed safety questjon.

.c e Providewritt'ennotificaticnwithin,'2_..hourstotheSRbof 4

c.

3 disagreement between the'PSRC a'nd the Plant Manager; however, the Plant Manager shall have responsibility for resolution of such d

l disagreements pursuant to 6.1.1 above.

GRAND GULF-UNIT 1 6-8,

'JAN 5 1982 l

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ADMINISTRATIVE CONTROLS

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6.8 PROCEDURES AND PROGRAMS 6.8.1 Written proce'dures shall be established, implemented and maintained covering the activities referenced below:

a.

The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2 February 1978.

b.

Refueling operations.

s c..

Sure:illance and test activities of safety related equipment.

I d.-

Security Plan implementationi e.

Emergency Plan implementation.

f.

Fire Protection Program implementation.

g.

PROCESS CONTROL PROGRAM implementation.

h.

.OFFSITE DOSE CALCULATION' MANUAL implementation.

.' Quality Assurance Program for effluent and environmental monitoring, (jev[em,

using the guidance in Regulatory Guide 4.15, February 197

.U Each procedure of 6.8.1 above, and changes thereto, shall b ;- c W,,y d[

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-by the PSRC and approved by. the Plant Manager prior to implementation and

-shall.be reviewed periodically as set forth in administrative procedures.

6.8.3 Temporary changes to procedures of 6.8.1 above may be made provided:

a.

The intent of the original procedure is not altered.

b.

'The change is approved by two members of the unit management staff, at least one of whom holds a Senior Reactor Operator's Licens.e on l

the unit affected.

c.

The change is documented, reviewed by the PSRC and approved by the Plant Manager within 14 days of implementation.

6.8.4 The following programs shall be established, implemented, and maintained:

a.

Primary Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside l

containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels.

The systems include the:

l 1.

'RCIC system outside containment containing steam or water (,

l except the drain line to the main condenser).

2.

RHR system outside containment containing steam or water, except l

the line to the LRW system and headers that are isolated by manual valves.

3.

HPCS system.

4.

LPCS system.

5.

Hydrogen analyzers of the combustible gas control system.

I GRANDGULF-UNIi1 6-13 l '

e.

' to AECM-82/115 Grand Gulf Unit 1 Technical Specifications Pages 3/4 6-12 & 3/4 6-13,-

dated March 3, 1982 1

1)

. Specifications 3.6.1.9, 4.6.1.9.1, and 4.6.1.9.2:.The shown chaages are provided as a followup to AECM-82/28 which offered information regarding the design and use of containment purge.

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LIMITING CONDITION FOR OPERATION l

shall be OPERABLE and either the 20 inch or the 6 inch pU, isolation 3.6.1.9 The containment purge system supply and exhaust rge line supply and exhaust isolation valves may be open for pi:rge system operation.with nd 7#E-operation d to hours per 305 days.

[g g,Q

[,

7 YF

""# "##^

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2 and 3, K va.fves y -

ACTION:

With a containment purge system supply and/or exhaust isolation a.

, valve (s) inoperable,'close the inoperable valve (s) or.otherwise isolate the penetration (s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN with e

/m,e 40/"o/jiavy

/4 followin'g 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , 00g e en-b.

With a containment purge system in operation with upply and/or exhaust isolation valve (s) open for more than.

hours per 365 days, close the open valve (s) or otherwise isolate the penetratian(s) within four hours or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the folloking 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

With a containment purge' supply and/or exhaust isolation valve with l

resilient material seals having a measured leakage rate exceeding the limit of Surveillance Requirement 4.6.1.9.2, restore the l

inoperable valve (s) to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT SHUTOOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTOOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . P 5

a GRAND GULF-UNIT 1 3/4 6-12 Mf,R 3 ' 19 6

.., N MNTS hANNWW

-l r

i ggf 3 ;gif.jj,,r.e i,I

.'. '... ~ ~ ' '

' ~ "

CONTAINHENT ' SYSTEMS' 'Cf[l lisc. :t.'(Livinii r r. : r e - -..--..

~,.

g,.g g

11) v ani,i s av..

I I

8 SURVEILLANCE REQUIREMENTS m

the.to lyh yonye like) 4.6.1.9.1 Thei cumulative time that the containment purge system has been in operation withTsupply and exhaust isolation valves open during the past 365 days shall be determined at least once per 7 days.

h

/$ MONN$

44.6.1.9.2 At least once per 02 1:y; each containment p0fge supply and exhaust isolation valve with resilient material seals shall be demonstrated OPERABLE by ver'fying that the measured -leakage rate is. less than or equal to (0.01) L when pressurized to P,. P a

l l

l GRAND GULF-UNIT 1 3/4 6-13

AR 3 1982

Attschment 9 to.AECM-82/115 Grand Gulf Unit 1 Technical Specifications. Page (proposed) 3/4 7-45, dated March 30, 1982 1)

Specification 3/4.7.10 (proposed): This proposed specification is

'in response to a SER open item identified in Section 2.5.5 of Supplement 1.

AE2D10

Attechment 9 to AECM-82/115

-PLANT SYSTEMS 3/4.7.10 ACCESS ROAD AND DRAINAGE BASIN SLOPES LIMITING CONDITION FOR OPERATION 3.7.10~

The slopes of the access roa'd (at Culvert No. 1) and the drainage basin shall be OPERABLE.

APPLICABILITY: 'At all times.

ACTION:

~

Withanyo[theslopesoftheaccessroad(atCulvertNo.1)orthe drainage basin showing irregularities in alignment and variances from originally constructed slopes:

a.

. Repair the slopes within 120 days or be in Hot Shutdown in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and Cold Shutdown within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.7.10 Inspection of the access road slopes (at Culvert No. 1) and the slopes of the drainage shall be conducted once a year and as needed af ter a storm.

AE2D11 i

' 0 to AECM-82/115 Grand Gulf Unit 1 Technical Specifications Page 3/4'3-2, Table 3.3.1-1 1)

Specification 3.3.1: This Specification references Table 3.3.~1-1 that requires APRM's during Operational Condition 5. -The basic concern that arises from requiring APRM operability during mode 5 is that'it will increase doses to personnel. This occurs because instrument cables underneath the reactor associated with the APRM's cannot be removed. Not only must workers involved in CRD maintenance proceed at a slower pace with work to avoid damaging cables, but some cables inevitably will have their connectors broken and have to be repaired. This process involves re-soldering Connectors.

The following is the process used for a CRD removal. First, since the area under the vessel is a high-radiation area, workmen and supervisor's will be required to suit up.

This would normally be full Anti-C's and might mean a " bubble suit" which would mean dragging an air hose behind. The area underneath the core resembles a mass of brittle organized spaghetti, with the TIP machines, SRM's, IRM's, and of course the APRM cables.

Removal of the " shoot-out" steel is quite involved and requires massive socket wrenches with handles in excess of three feet in length. Even with the socket attached the handle cuts-an are less than a foot below the " shoot-out" steel. As the' cables are hanging beneath the steel with curls drooping a foot or more below the steel it is very difficult to turn the socket without hooking the-handle of the wrench in one of the curls. Also the wiring is hanging beneath the steel so that when the nuts (which are' attached to approximately 2" diameter bolts) are removed the cables are still looped around the steel.

Assuming the steel can be removed without snapping the brittle connectors, removal of the drive is more difficult because cables-are hanging down around the bottom.of it.

If the APRM's are not required the cables can be disconnected, the ends can be covered, then bagged with dessicate, and coiled and tied and thus moved out of the way. After removal of the steel the coils can be tied above the bottom of the CRD.

When the seal on the drives is broken the water in the under-piston area drains out.

If the APRM cables are removed and tied out of the way they do not risk getting wet.

The connectors are very brittle and will not stand being broken.

Repair to a broken connector requires soldering.

In conclusion, operability of the APRM's during mode 5 presents a problem in_ performing maintenance on CRD's (normally involving removal). The base concern is that with a requirement for the APRM's being operable,. disconnecting APRM cables from instruments is not allowed. By not allowing removal of the APRM cables personnel doses are increased both by the increased work time that AE2D12

will be required to remove the CRD's and the extra time that will be required to repair the brittle connectors which will inevitably be broken in greater numbers. Replacing connectors is not a simple repair.since it does involve soldering connections.

It should be pointed out that APRM operability is not required since the SRM's and IRM's provide sufficient indication of core activity.

Succintly stated, if APRM's are required operable in mode 5, down time in outages will be; increased and man-run doses on CRD maintenance and replacement of broken' connectors will be substantially increased.

1 AE2D13

~.

i_...._.....-.

_' ',{ ^ [1', *]

TABLE 3.3.1-l' 5

REACTOR PROTECTION SYSTEM INSTRUMENTATION 5

cn APPLICABLE MINIMUM E

OPERATIONAL OPERABLE CllANNELS FUNCTIONAL UNIT CONDITIONS PER TRIP SYSTEM (a)

ACTION c.

1.

Intermediate Range Monitors:

T.

a.

Neutron Flux - High 2

3 1

3, 4 2

2 ID)

S 3

3 b.

Inoperative 2

3 1

3, 4 2

2 5

3 3

2.

Average Power Mange Monitor (c);

a.

Neutron Flux - High, Setdown 2

3 1

l

~'

g

,,{

c t

Y b.'

Flow Biased Simulated Thermal Power - High 1

3 4'

c.

Neutron Flux.- liigh 1

3 4

d.

Inoperative 1, 2

'3 '

1 4

Y 3.

Reactor Vessel Steam Dome E

Pressure - liigh 12, 2(d) 2 1

0 4.

Reactor Vessel Water Level - Low, Level 3 1, 2 2

1 5'

5.

Reactor Vessel Water Level-High, 4

Level 8 1(*)

2 4

e 6.

Main Steam Line Isolation Valve -

Closure 1(*)

4 4

hN 7.

Main Steam Line Radiation - High 1,2(d) 2 5

8.

Drywell Pressure - liigh 1, 2(I)'

2 1

s

~

Attachment'11 to AECM-82/115=

Grand Gulf Unit 1 Technical Specifications Page 3/4 1-5

~

I 1)

Specification 4.1.3.1.4:

We have reviewed Specification 4~.1.3.1.4; and, it is our understanding that the aim of this specification is to prevent recurrence of a' failure of control rods to scram as s

i.

occurred in July, 1980, r Browns Ferry 3.

1 First note that operability of the scram discharge volume (SDV) vent and drain valves is verified once per~31 days and each SDV vent and drain valve is-cycled at,least once per 92 days.. Logic checks and channel checks are also provided periodically for-2 associated instrumentation.

f

-Thus, the~only concern' apparent is verification that the SDV and scram discharge instrument volume (SDIV) remain relatively water-free.

r We do not believe the scram from at least 50% rod density should be reauired, first because it imposes a unnecessary strain on the drives. The only information provided is the verification of the 4

operability of the mechanical-equipment required to scram and that sufficient volume was provided in the SDV'to hold exhaust water from the CRD over piston area. The operability 1of the actual equipment for accomplishing the scram is already provided under various other surveillance requirements.

As for the capability of the SDV to accept exhaust water, we 'o not d

believe this is a valid concern based on the differences between Grand Gulf and Brown's Ferry.

Brown Ferry 3 is a BWR 4.

The water exhausted frem the over piston area of the CRD piston is exhausted to one'of two scram discharge volumes (see Figure 1). The two SDV's each correspond to one half of the core and are called the east bank and the west bank SDV's.

This is similar to Grand Gulf as far as terminology is concerned.

There are, however, significant differences in the installed 4

hardware.

The Brown Ferry 3 SDV's are actually 6" headers which provide a volume to drain water exhausted from the over piston area of the CRD during a scram. At Brown's Ferry the two banks are connected-to the instrument volume via 2" lines (see Figure 2). The instrument volume contains level instruments. To assure that sufficient volume is maintained to accomodate water exhausted from all CRD's in a full core scram, instrumentation is provided in the SDIV which give a " scram discharge volume not drained" alarm at 3 gallons, a rod block at 25 gallons and a scram at.50 gallonsaof water accumulation in the SDIV.

At Brown's Ferry 3, 76 of 185 control rods failed to scram on a 1

manual scram. Of these 75 of the rods which failed' to insert ~ were in the East-Bank and one was a West Bank rod. The West-Bank rod inserted to notch 02*.

It took three more successive manual scrams before all control rods were inserted.

~* Full in is Notch 00.

AE2D14 l'

- - -. ~

~

The failure of the East Bank rods to scram was concluded to be'due-to water occupying a large part of the associated scram discharge volume. Of the three hypotheses as to why this occurred all conclude the failure was due to an obstruction or blockage of some type which prevented water accumulations in the SDV from draining to the instrument volume.

The key point is that connection between the scram discharge volume and the instrument volume is via a 2" drain line.

At Grand Gulf drainage from the CRD's is to a series of 8" and 10" diameter pipe, called headers, which drain to the scram discharge volume. The SDV is formed of 12" diameter pipe mounted in a position similar to an inverted "L" as shown in Figure 3.

The instrument volume is physically the low part of the "L".

There is a instrument volume associated with scram discharge volume.

r.

Instrumentation is provided to assure that-sufficient volume

remains to accommodate the water exhausted from the control rod drives during a full core scram. The instrument cause a " Scram Discharge Volume Not Drained" alarm at 3 gallons, a rod block 'at 26 gallons, and a scram at 54 gallons. There are two separatc S0V's

~

at Grand Gulf, each with a instrument volume and associated instrumentation.

~

It should also be noted that level instrumentation at Brown's Ferry 3 consists of level switches. These switches consist of a float chamber with a moving float rod that provides level information.~ - '"

In the past these type level switches have been found inoperable at, Hatch and Brunswick.

s At Grand Gulf instrumentation is by means of analog level transmitters which have no floats (see Figure'4.). Two separate level transmitters provide input to two channels each that 'pr' ovide one out of two taken twice~ logic.

)

~

Based on the above discussion, the requirement that GGNS demonstrate SDV operability by a scram from greater th n or equal a

to 50% rod density is excessive.

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lines SDIV at 25 gal.

tiCUs ilCUs ilCUs liCUs o) 1 level switch (alarm) at 3 gal.

Other drains SDV Floor elevation u f1 1

fl y

'M '

'b 4" drain header Other drains and vents licactor building equipment drain sump m

~

Figure 2 Sketch of Browns Ferry SDV Equipment Layout (elevation)

O O

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/Wadey // 6> AEUst-r2 fir-SCRAM DISCHARGE VOLUME 10" pipe f

Cl i

CL ECU's Ccnnection-Tvo. 196 l

Ch i

~45 ft.

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/ 1" vent

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connec dc.is

~20 ft.

12,, p:.pe

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VALVE 5 SHOWN IN 75 s For inWeamen Is.

,o O SD V /5crV NCRMAL AT POWER PCSITICI 9

do no+ show gLai con p,gu,-duon Co.c. do nol necenersly sharc cornman poin4s. )

l Od 5 v.'=RE :: Ct.

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ML20050B689

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