ML20197J256

Proposed Tech Specs Changing Pages 2.3-6,2.3-7,3.1-11, 3.1-14,3.1-16,3.4-8,3.8-2,3.8-3,4.3-1,4.5-13 & 6-1

ML20197J256
Person / Time
Site:	Oyster Creek
Issue date:	12/10/1997
From:	GENERAL PUBLIC UTILITIES CORP.
To:
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ML20197J248	List: ML20197J243 ML20197J256
References
NUDOCS 9801020122
Download: ML20197J256 (12)

Text

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t OYSTER CREEK NUCLEAR GENERATING STATION OPERATING LICENSE ,

NO. DPR 16 l

TECilNICAL SPECIFICATION CilANGE REQUEST NO.171  !

DOCKET NO. 50-219 l Applicant subinits by this Technical Specification Change Request Nc.171 to the Oyster Creek Nuclear Generating Station Technical Specl0 cations, modined pages: 2.3-6,2.37,3.111, 3.1-14,3.1 16,3.4-8,3.8 2,3.8 3,4.3-1,4.513, arxl 6-1.

By: 9 h[

Michael B. Roche Vice President and Director Oyster Creek Sworn to and Subscribed before me this (*N day of b"~ #'"- 1997.

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A Notary Public of New Jersey OEggAltMNE E.tAVIN

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NUCLEAR Fork 6d River, NJ N73103N Tel 609 9714000 i

December 10, 1997  !

6730-97 2222 i

~. Mr. Kent Tosc ,h Director  ;

Bureau of Nuclear Engineering -

Department of Environmental Protection -

CN 415

- Trenton, NJ 08628

Subject:

Oyster Creek Nuclear Generating Station  :

Operating License No. DPR 16 Technical Specification Change Request No.171 Enclosed is one copy of the Technical Specification Change Request No.171 for the Oyster Creek i-Nuclear Generating Station Operating License. ,

This document was filed with the U.S. Nuclear Regulatory Commission en December 10. 1997.

Very truly yours,

%IlJ/%(r Michael D. Roche Vice President and Director Oyster Creek MBR/JJR

- Enclosure i

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION i

1 IN Tile MA1TER OF j GPU NUCLEAR DOCKET NO. 50-219  ;

CORPORATION -

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CERTIFICATE OF SERVICE This is to certify that a copy of Technical Specification Change Request No.171 for the Oyster Creek Nuclear Generating Station Technical Specifications, filed with the U.S. Nuclear Regulatory Commission on December 10,199has this day of December 10. 1997, been served on the Mayor of lxcy Township, Ocean County, New Jersey by deposit in the U.S. mail, addressed as

- follows:

The lionorable L. Nick Mayor of Lacey Township 818 West Lacey Road Forked River, NJ 08731 By: )L Nb.-

Michael B. Roche Vice President & Director Oyster Creek 7

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GPU Nucleat,Inc.

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December 10,1997 6730-97-2222 l

The !!onorable L. Nick Mayor of Lacey Township l 818 West lxey Road

~ Forked River, NJ 08731

Subject:

Oyster Creek Nuclear Generating Station Operating License No. DPR 16  ;

Technical Specification Change Request No.171 l

Enclosed is one copy of the Technical Specification Change Request No.171 for the Oyster Creek  !

Nuclear Generating Station Operatlag License.

This document was filed with the U.S. Nuclear Regulatory Commission on December 10, 1997.

Very truly yours, ,

he '0 ~

Michael B. Roche Vice President and Director Oyster Creek MBR/JJR Enclosure -

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, ENCLOSURE OYSTER CREEK NUCLEAR GENERATING STATION OPERATING LICENSE NO. DPR 16 DOCKET NO. 50-219 TECilNICAL SPECIFICATION Cil ANGE REQUEST NO.171 Applicant hereby requests the Commission to change Appendix A of the above captioned license as follows:

1. Sections to bef.hangei A. Administrative Changel

1. Specification 2.3 bases: The bases have been expanded to more clearly describe the justifications for reactor safety functions.

2. Table 3.1.1.G.1 Correct a typographical error. Change a " < " to a " s; ".

3. Table 3.1.1.M.2 Move a relay number from the 10 second timer list to the 120 r.econd timer list.

4. Section 4.3.C Change a reference to the correct section of 10 CFR.

5. Section 61.1 Correct the title of the Director - Operations and Maintenance,

11. Itchnicalfbanges

1. Table 3.1.1, note b Correct an inappropriately revised setroint.

2. Section 3.4 Ilases Remove an outdated section.

3. Section 3.8 ilases Remove an indication referenu in the bases.

4. Section 4.5 Ilases Replace a brand name with a ? nical description.

2. Change Requestei Replace old pages: 2.3-6, 2.3-7, 3.1-11, 3.1-14, 3.1-16, 3.4-8, 3.8-2, 3.8-3, 4.3 1, 4 5-13, and 6-1 with new pages : 2.3-6, 2.3-7, 3.1-11, 3.1-14, 3.1-16, 3.4-8, 3.8 2, 3.8-3. 4.3-1, 4.5-13, and 6-1

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6730 97 2222 Enclosure

. Page 2

3. No Sienificant flazards Determiration Pursuant to 10 CFR 50.91, this Technical Speel0 cation Change Request has been determined to contain No Signincant liarards. These evaluations are specified in 10 CFR 50.92.

A. Administrative Chances

1. Speci0 cation 2.3 bases Plant modi 0 cations and new computer analysis have caused the existing bases to become ambiguous and unclear. He proposed new wording more clearly dennes the justincation for the existing limits.

2 Table 3.1.1.G.1 The liigh Drywell Pressure setpoint was increased to s 3.5 psig in Amendment 112. Amendment 171 contained a typographical error which inadvertently changed the value to < 3.5 psig. This request corrects that error.

3 Table 3.1.1.ht.2 The service water pumps have two different time delays. SKI A and SK2A are 120 second timers, while SK7A and SK8A are 10 second timers. Technical Specincation Amendment 160 contained a typographical error which placed SK2A in the 10 second timer column. This request corrects that error.

4 Section 4.3.C 4

The inservice Test Program requirements are regulated pursuant to 10 CFR 50.55a (0(6)(i). Amendment 82 contahied a typographical error which referred requirements to 10 CFR 50.55a(g)(6)(0. This request corrects that error.

5 Section 6.1.1 Section 6.1.1 of the Technical Speci0 cations allows for the duties and responsibilities of the Vice President and Director of Oyster Creek to be assumed by the Deputy Director. The title of the Deputy Director position has been changed to e Director - Operations and hiaintenance. No change in the duties or responsibilities of this position is proposed, This request changes only the title from the Deputy Director to Director - Operations and hiaintenance.

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, I 6730-97 2222 Enclosure

, Page 3 The preceding Ove change requests are typical of e . ample 1.c.2.e.1 in SlFR7744. hefore, they do not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or

2. Create the possibility of a new or different kind of accident from any accident previously evaluated; or

3. Involve a significant reduction in the margin of safetyt in that they are purely administrative changes to achieve consistency or correct an error in the Technical Speci0 cations.

11. Irdtuical Changen

1. Table 3.1.1, note b.

The low condenser vacuum and the main steam isolation valve (htSIV) closure anticipatory scrams provide reactor protection against pressure and flux transients which occur during turbhie stop valve and htSIV closure from power. While in the startup mode and with reactor pressure below a predetermined setpoint value, bypassing these anticipatory scrams is permitted to allow for the establishment of turbine seals and condenser vacuum.

Prior to restart from refueling outage 10R, the lower bound of the hiinimum Critical Power Ratio (hiCPR) safety limit pressure range was raised from 600 psig to 800 psia. As part of the Cycle 10 reload submittal, the Technical Speci0 cation scram bypass reactor pressure setpoint in Table 3.1.1, note b, was changed from < 600 psig to

< 800 psia. The Technical Specification bases specifying < 600 psig were not changed.

This proposed change would revise the Table 3.1.1, note b, setpoint back to its original value of 600 psig.

Although the evaluation to raise the h1CPR lower bound was completed, the actual evaluation to raise the Technical Specification was not performed. Due to the discrepancy between the Technical Specifications and the bases, the new value was never implemented at the Oyster Creek plant. Plant procedures and standing orders remained at the

<600 psig value. Therefore, plant operation was never impacted by the < 800 psia value.

6730-97 2222 Enclosure

. Page 4 This request has been determined to involve No Significant llazards in that it does not:

1. Involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated; (or)

The proposed chaage would restore the original value of < 600 psig. This lower value would not increase the probability of any accident as it provides a more conservative level below which p;otection can be bypassed.

2. Create the possibility of a new or different kind of accident from any accident previously evaluated; (or)

The proposed change would restore the original value of < 600 psig. The setpoint of a bypass cannot create a different kind of accident, it can only affect the severity.

3. Involve a significant reduction hi a margin of safety; As the requested change lowers the bypass setpoint, the margin of safety will be increased.

2. Section 3.4 Bases The bases for Section 3.4 presently contain a paragraph with describes the ability of the Control Rod Drive pump to provide high pressure injection capabilities for Small Break LOCAs below the size of .002 square feet. While this statement is true and does reflect an original design basis, it had been superceded by more recent submittals (specifically in reference to 10 CFR 50 Appendix K) which take no credit for CRD system. Additionally, the new small break LOCA is now calculated down to 0.05 square feet, far oeyond the capabilities of the CRD system. The paragraph (and its reference) are being removed from the Bases to minimlic confusion about the function of the CRD system.

This request has been determined to involve No Significant llazards in that it does not:

1. Involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated; (or)

The proposed change to t'.c isases removes a possible area of confusion from die Technical Specificationr,, ar41 updates the Bases to reficct the results of newer, approved methodologies. Therefore, no change to any probability calculation occurs.

6730-97 2222 Enclosure

, Page 5

2. Create the possibility of a new or different kind of accident from any accident previously evaluated; (or)

The proposed change addresses an existing accident (Small Break LOCA) and removes outdated and possibly confusing information. Therefore, no new or different kind of accident is created.

3. Involve a signincant reduction in a margin of safety; The propose / Sange does not change the way that the plant is operated or the way design bases are maintained, it only removes an outdated and possibly confusing paragraph from the Bases, therefore, no margin of safety is affected.

3. Section 3.8 bases The bases for Section 3.8 describe the features and capabilities of the Isolation Condenser system. The description of the vent lines from the condensers discusses a radiation monitor installed to provide the operators with information about potential leaks. The Isolation Condenser system was recently r.odified to remove the radiation monitors. This request will revise the Technical Specincations to accurately describe the Oyster Creek plant.

This request has been determired to involve No Signincant llazards in that it does not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated; (or)

The Isolation Condenser Radiatica u onitors had no impact of the operation of any plant system. Additionally, the monitors were not relied upon for any post accident evaluations. They were removed from the plant using the 10 CFR 50.59 process. As this request updates the Technical Speci0 cation Bases to reDect the plant as currently coMgured, no impact on the probability or consequences of any previously evaluated accident is possible.

2. Create the possibility of a new or different kind of accident from any accident previously evaluated; (or)

The Isolation Condenser Radiation monitors had no impact of the operation of any plant system. Additionally, the monitors were not relied upon for any post accident evaluations, They were removed from the plant using the 10 CFR 50.59 process. As .

this request updates the Technical Specification Dases to reflect the plant as currently con 0gured, no new or different kind of accident is created.

i 6730-97 2222 T Enclosure  !

.- Page 6 i 4

3. Involve a significant reduction in the margin of safety; The isolation Condenser Radiation monitors had no impact of the operation of any [

plant system. Additionally, the monitors were not relied upon for any post accident  ;

evaluations. They were removed from the plant using the 10 CFR 50.59 process. As  ;

this request updates the Technical Specification Bases to reflect the plant as currently configured, no reduction in any margin of safety can occur.  !

4. Section 4.5 Bases  !

r The bases being changed currently specify a brand name product. The requested change would replace the brand name with a tecnnical description of the product  :

allowirig for alternate replacements.

This request has been determined to involve No Significant liazards in that it does not:

1. Involve a significant increase in the probability or consequence of an accident previously evaluated: (or) ,

No change to any procedure, nor any modification to any system is requested. 'the same surveillarce will be performed at the same frequency. Only the brand of the chemical used to perform the surveillance will be affected. As an equivalent chemical will be selected, no increase in the probability or consequences of an accident

• previously evaluated can be created.

2. Create the possibility of a few or different kind of accident from any accident  ;

previously evaluated: (or)

No chang: to any procedure, nor any modification to any system is requested. The ,

same surveillance will be performed at the same frequency. Only the brand of the chemical used to periorm the surveillance will be affected. As an equivalent chemical Till be selected, no new or different kind of accident previously evaluated can he created.-

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2. Involve a significant reduction in the margin of safety; No change to any procedure, nor any modification to any system is requested. The  !

same surveillance will be perfornal at the same frequency. Only the brand of the l chemical ur.ed to perform the surveillance will be affected. As an equivalent chemical . I will be selected, no margin of safety can be affected. l r

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TECHNICAL SPECIFICATION  !

) CHANGE REQUEST d l No.171  ;

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The reactor coolant system safety valves offer yet another protection feature for the reactor coolant system pressure safety limit since these valves are sized assuming no credit for other pressure relieving devices, in compliance with Section I of the ASME Boiler and Pressure Vessel Code, the safety valve must be set to open at a pressure no higher than 103% of design pressure, and they must limit the reactor pressure to no more than 110% of design pressure. The safety valves are sired according to the Code for a condition of main steam isolation valve closure while operating at 1930 MWt, followed by (1) a reactor scram on high neutron flux, (2) failure of the recirculation pump trip on high pressure,

('.) failure of the turbine bypass valves to open, and (4) failure of the isolation condensers and relief valves to operate. Under these conditions, a total of 9 safety valves are required to turn the pressure tramient. The ASME D&PV Code allows a 1 1 % of working pressure (1250 psig) variation in the lift point of the valves. This variation is tecognized in Specification 4.3.

The low pressure isolation of the main steam line at 825 psig was provided to give protection against fast reactor depressur! ration and the resulting rapid cool <iown of the vessel. The low-pressure isolation protection is enabled with entry into IRM range 10 or the RUN mode. In addition, a scram on 10% main steam isolation valve (MSIV) closure anticipates the pressure and flux transients which occur during normal or inadvertent isolation valve closure. Hypass of the MSIV closure scram function below 600 psig is permitted to provide scaling steam and allow the establishment of condenser vacuum. Advantage is taken of the MSIV scram feature to provide protection for the low-pressure portion of the fuel cladding integrity safety limit. To continue operation beyond 12% of rated power, the IRM's must be transferred into range 10. Reactor pressure must be above 825 psig to successfully transfer the IRM's into range 10. Entry into range 10 at less than 825 psig will result in main steam line isolation valve closure and MSIV closure scram. This provides automatic scram protection for the fuel cladding integrity safety limit which allows a maximum power of 25% of rated at pressures below 800 psia. Below 6M) psig, when the MSIV closure scram is bypassed, scram protection is provided by the IRMs.

Operation of the reactor at pressure lower than 825 psig requires that the mode switch be in the STARTUP position and the IRMs be in range 9 or lower. The protection for the fuel clad integrity safety limit is provided by the IRM high neutron flux scram in each IRM range. The IRM range 9 high flux scram setting at 12% of rated power provides adequate thermal margin to the safety limit of 25% of rated power. There are few possible significant sources of rapid reactivity input to the system through IRM range 9: effects of increasing pressure at zero and Ic void content are minor; reactivity excursions from colder makeup water, will cause an IRM high flux trip; and the control rod sequences are constrained by operating procedures backed up by the rod worth minimizer, in the unlikely event of a rapid or uncontrolled increase in reactivity, the IRM system would be more than adequate to ensu.c a scram before power could exceed the safety limit. Furthermore, a mechanical stop on the IRM range switch requires an operator to pull up on the switch handle to pass through the stop and enter range 10. This provides protection against an inadvertent entry into range 10 at low pressures.

The IRM scram remains active until the mode switch is placed in the RUN position at which time the trip becomes a coincident IRM upscale, APRM downscale scram.

OYSTER CREEK 2.3-6 Amendment No.150,

The low level water level trip setting of 11'5' above the ky of the active fuel has been established to assurb that the reactor is not operated at a water level below that for which the fuel cladding integrity safety limit is applicable. With the !, cram set at this point, the generation of steam, and thus the loss of inventory is stopped. For example, for a loss of feedwater flow a reactor scram at the value indicated and isolation valve closure at the low low water level set point results in more than 4 feet of water remaining above the core after isolation (6).

During periods when the reactor is shut down, decay heat is present and adequate water level must be maintained to provide core cooling. Thus, the low low level trip point of 7'2" above the core is provided to actuate the core spray system (when the core spray system is required as identified in Section 3.4) to provide cooling water should the level drop to this point.*

The turbine stop valve (s) scram is provided to anticipate the pressure, neutron flux, and heat Oux increase caused by the rapid closure of the turbine sky valve (s) and failure of the turbine bypass system.

The generator load rejection scram is provided to anticipate the rapid increase in pressure and neutron flux resulting from fast closure of the turbine control valves to a load rejection and failure of the turbine bypass system. 'this scram is initiated by the loss of turbine acceleration relay oil pressure.

The timing for this scram is almost identical to the turbite trip.

The undervoltage protection system is a 2 out of 3 coincident logic relay system designated to shift emergency buses C and D to on-site power should normal power be lost or degraded to an unacceptable level. 'Ihe trip points and time delay settings have been selected to assure an adequate power source to emergency safeguards systems in the event of a total loss of normal power or degraded conditions which would adversely affect the functioning of engineered safety features connected to the plant emergency power distribution system.

References (1) FDSAR, Volume 1, Section Vll-4.2.4.2 (2) FDSAR, Amendment 28, item lil.A-12 (3) FDSAR, Amendment 32, Question 13 (4) Letters, Peter A. Morris, Director, Division of Reactor, Licensing, USAEC, to John E.

logan, Vice President, Jersey Central Power and Light Company (5) FDSAR, Amendment 65, Section B.XI (6) FDSAR, Amendment 65, Section B.lX OYSTER CREEK 2-3.7 Amendment No.175

• Correction 11/30/87

TABLE 3.1.1 PROTECTIVE INSTRUMENTATION REQUIREMENTS Reactor Modes Min. No. of hiin. No. of

, in which Function OPERABLE Instrutnent

- Must Be OPERADLE or Channels Per .

OPERATING Trip [ tripped] OPERABl.B Action Trin System EMIXh9B StitiDS Shu!dein Rtfus] StatBID RUD Trio Systents Ecquitsp D. C91t.SPIE ,

i low Low " X(t) X(t) X(t) X 2 2(pp) Consider the Reactor Water respective Level cose spray loop inoper.

able and comply nth 2 liigh Drywell 53.5 X(t) X(t) X(t) X 2(k) 2(k)(pp) Spec. 3.4 Pressure psig 3 law Reactor 2 285 psig X(t) X(t) X(t) X 2 2(pp)

Pressure (vaht -

permissive)

E. ContaituntutSPIE Comply with Technical Specification 3.4

~

F. PrunatLC9ntailuncath91atinD ,

I liigh Drywell 53.5 X(u) X(u) X(u) X 2(k) 2(k)(oo) Isolate Pressure psig containment or PLACE IN COLD 2 Low Low 27'2" X(u) X(u) X(u) X 2 2(oo) SIL'TDOWN Reactor Witer above CONDITION Level TOPOF ACTIVE FUEL G. AutoJDal!G Depigssurization I liigh Drywell s3.5 X(v) X(v) X(v) X 2(k) 2(k) .Sec note h Pic mre psig l

2 Low Low Low 24'8" X(v) X(v) X(v) X 2 2 See note h Reactor Water above Level TOP OF ACTIVE FUEL 3 Core Spray >21.2 X(v) X(v) X(v) X Note i Note i See note i.

Booster Pump psid

&p Permissive OYSTER CREEK 3.1 11 Amendment No.: 190, Change 4; Correction 5/11/84

r TABLE 3.1.1 PROTECTIVE INSTRUMENI ATION REQUIREMENTS  !

Reactor Modes Min. No. of Min. No. of in which Function OPERABLE Instrument i Must Be OPERABLE or Chat.nels Per OPERATING l Trip [ tripped) OPERABLE Action  !

Function Setting Slaadens Erfusi Startup Run Trip Systems Trio System Bagdad* i Time dday I M. Danni Generator Load Seguagg aAer Ignars energization of l relay 1 CRD pump 60 mi15% X X X X 2(m) 1(n)(kk) Consider the pump inoperable 7 and comply

., with Spec.

3.4.D (see '

Note q) 2 Service Water 120 sec.i 15% - X X X X 2(o) 2(p)(kk) Consider Pump (aa) (SKIA) the pump (SK2A) inoperable i 10 sec.115% and comply r (SK7A) within 7 (SK8A) days (See l Note q) 3 Reactor 166 sec.115% X X X X 2(m) 1(n)(kk) Consider ,

Building the pump Closed inoperable Cooling Water and comply Pump (bb) within 7 days (See  !

Note q)

N. Loss of Power a 4.16KV "  !

X(ff) X(ff) X(ff) . X(ff) 2 1(kk)

Enugency i Bus Undervoltage (Loss of Voltage) b 4.16 KV " X(ff) X(if) X(ff) X(ff) See note ec 2 3(kk)

Enugency Bus -

- Undervoltage (Degraded

, Voltage)

OYSTER CREEK 3.1 14 Amendment No. 15,44,60, 80,160,171, l

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l TABLE 3.1.1 (CONT'D)

. Action required when minimam conditions for operation are not satisfied. Also permissible to trip inoperable trip system. A channel may be placed in an inoperable i status for up to six hours for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE instrument channel in the same trip system is monitoring that parameter. l See Specification 2.3 for Liruiting Safety System Settings.  !

NOTES:

a. Permissible to bypass, with control rod block, for reactor protection system reset in REFUEL MODE.

b. Permissible to bypass below 600 psig in REFUEL and STARTUP MODES,

c. One (1) APRM in each OPERABLE trip system may be bypassed or inoperable provided

, the requirements of Specification 3.1.C and 3.10.C an satisfied. Two APRMs in the same quadrant shall not be concurrently bypassed except as noted below or permitted by note.

Any one APRM may be removed from service for up to six hours for test or calibration without inserting trips in its trip system only if the remaining OPERABLE APRMs meet the requirements of Specification 3.1.B.1 and no control rods are moved outward during the calibration or test. During this short period, the requirements of Specifications 3.1.B.2, 3.1.C and 3.10.C need not be met.

d. The IRMs shall be inserted and OPERABLE until the APRMs are OPERABLE and reading at least 2/150 full scale,

e. Offgas system isolation trip set at .12,000 mrem /hr. Air ejector isolation valve closure time delay shall not exceed 15 minutes,

f. Unless SRM chambers are fully inserted,

g. Not applicable when IRM on lowest range,

h. With one or more instrument channel (s) inoperable in one ADS trip system, place the relay contact (s) for the inoperable initiation signal in the tripped condition within 4 days, or declare ADS inoperable and take the action required by Specification 3.4.B.3.

With one or more instrument channel (s) inoperable in both ADS trip systems, restore ADS initiation capability in at least one trip system within I hour, or declare ADS inoperable and take the action required by Specification 3.4.B.3.

Relief valve controllers shall not be bypassed for more than 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> (total time for all controllers) in any 30-day period and only one relief valve controller may be bypassed at a time, OYSTER CREEK 3.1-16 Amendment No.:75,108,110,171 184,190,191

~d The containment spray system is provided to remove heat energy from the containment in the event i of a loss of coolant accident. Actuation of the containment spray system in accordance with plant  ;

emergency operating procedures ensures that containment and torus pressure and temperature  ;

conditions are within the design basis for containment integrity, EQ, and core. spray NPSH l requirements. De flow from one pump in either loop is more than ample to provide the required l heat removal capability (2). The emergency service water system provides cooling to the l' containment spray heat exchangers and, therefore, is required to provide the ultimate heat sink for

the energy release in the event of a loss-of-coolant accident. The emergency service water pumping l

- requirements are those which correspond to containment cooling heat exchanger performance  !

I Implicit in the containment cooling description. Since the loss-of-coolant accident while in the cold shutdown condition would not require containment spray, the system may be deactivated to permit  !

integrated leak rate testing of the primary containment while the reactor is in the cold shutdown j condition. -

[

The core spray main pump compartments and containment spray pump compartments were provided j

. with water tight doors (4). Specification 3.4.E ensures that the doors are in place to perform their  ;

intended function.-

Similarly, since a loss of coolant accident when primary containment integrity is not required would  !

not result in pressure build up in the drywell or torus, the containment spray system may be made ..

Inoperable under these conditions.  ;

Reference 2

1. NEDC-31462P, ' Oyster Creek Nuclear Generathig Station SAFER /CORECOOlJGESTR-LOCA less-of-Coolant Accident Analysis," August 1987. >

2. Licensing Application, Amendment 32, Question 3 4

3. (Deleted) ,

1 4. Licensing Application, Amendment 18, Question 4 ,

5. GPUN Topical Report 053, " Thermal Limits with One Core Spray Sparger" December 1988.  !

6. NEDE-30010A, ' Performance Evaluation of the Oyster Creek Core Spray Sparger*, January 19M.

t

7. Letter and enclosed Safety Evaluation, Walter A. Paulson (NRC) to P. B. Fiedler (GPUN),

July 20,- 1984.

8. APED-5736, ' Guidelines for Determining Safe Test Intervals and Repair Times for Engineered Safeguards *, April 1%9.

. OYSTER CREEK 3.4-8 Amendment No.: 165 i

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Each condenser containing a minimum total water volume of 22,730 gallons prosides 11,060 gallons above the condensing tubes. Based on scram from a reactor power level of 1950 htWt (the design basis power level for the isolation conder.sers) the condenser system can accommodate the reactor decay heat G 5)(corrected for U 239 and NP-239) for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 40 minutes without need for makeup water. One condenser with a minimum water volume fo 22,730 gallons can accommodate the ret.ctor decay heat for 45 minutes aller scram from 1950 hiWt before makeup water is required. In order to accommodate a scram from 1950 htWt and cooldown, a total of 107,500 gallons of makeup water would be required either from the condensate storage tank or from tFe fire protection system. Since the rated reactor power is 1930 hiWt, the above calculations ,epresent conservative estimates of the isc,lation condenser system capability.

The vent lines from each of the isolation condenser loops to the main steam lines downstream of the main steam lines isolation valves are provided with isolation valves which close automatically on isolation condenser actuation or on signals which close the main steam isolation valves.11igh temperature sensors in the isolation condenser and pipe areas cause alarm in the control room to alert the operator of a piping leak in these areas.

Specification 3.8.H allows reduction in redundancy ofisolution capability for isolation condenser i inlet (steam side) isolation valves. Reaconable assurance ofisolation capability is provided by testing the operability of the redundant valve. Specification 3.8.F allows short term inoperability of the AC motor operated isolation condenser outlet (condensate return) valve. It is not necessary to test the redundant DC motor operated "alve as this valve is normslly in the closed position. These specifications permit troubleshooting and repair as well as :outine maintenance, such as valve stem packing addition or replacement, to be performed during reactor operation without reducing the redundancy of the isolation condenser heat sink function. The out of senice time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is consistent with that permitted for primyy containment isolation valves W Either of the two isolation condensers can accomplish the purpose of the system. If one condenser is found to be inoperable, there is no immediate threat to the heat removal capability for the reactor and reactor operation may continue while repairs are being made. Therefore, the time out of service for one of the condensers is based on considerations for a one of two systemW The test interval for operability of the valves required to place the isolation condenser in operat:en is once/ month (Specification 4.8). An acceptable out of senice time, T, is then

letermined to be 10 days. Ilowever, if at the time the failure is discovered and the repair time is longer than 7 days, the reactor will be placed in the cold shutdown condition. If the repair time is not more than 7 days the reactor may continue in operation, but as an added factor of conservatism, the motor operated isolation condenser and condensate makeup valves on the operable isolation condenser are tested daily. Expiration of the 7 day period or inability to meet the other specifications requires that the reactor be placed in the cold shutdown condition which is normally expected to take no more than 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />. The out of service allowance when the system is required is limited to the run mode in order to require system availability, including

• redundancy, at startup.

OYSTER CREEK 3.8-2 Change No. 7 Amendment No.: 72

Esfatences:

l. FDSAR, Volume I, Section IW3

' 2. K. Shure and D. J. Dudziak, " Calculating Energy Release by Fission Products," U.S. AEC  !

Report, % APD T-1309, March 1961.  :

3. K. Shure, " Fission Product Decay Heat," in U.S. AEC Report, WAPD-HT-24, December 1961.

4. Specification 3.2, Bases.  ;

5. Specification 3.5.3.a.l. i i

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OYSTER CREEK 3.8 3 Amendment No.: 72

O 4.3 REACTOR COOLANT Apnlicability: Applies to the surveillance requirements for the reactor coolant system.

Obiective: To determine the condition of the reactor coolant system and the operation of the safety devices related to it.

Specification: A. Materials surveillance specirnens and neutron flux monitors shall be installed in the reactor vessel adjacent to the wall at the midplane of the active core.

Specimens and monitors shall be periodically removed, tested, and evaluatcJ to determine the effects of neutron fluence on the fracture toughness of the vessel shell materials. The results of these evaluations shall be used to assess the adequacy of the P-T curves A, B, and C in Figure 3.3.1,3.3.2 and 4 3.3.3. New curves shall be generated as required.

B. Inservice inspection of ASME Code Class 1. Class 2 and Class 3 systems and components shall be performed in accordance with Section XI of the

- ASME Boiler and Pressuie Vessel Code and applicable Addenda as required by 10 CFR, Section 50.55a(g), except where specific written relief has been granted by the NRC pursuant to 10 CFR, Section 50.55a(g)(6)(i).

C. Inservice testing of ASME Code Class 1, Class 2 and Class 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR, Section 50.55a(f), except where specific written relief has been granted by the NRC pursuant to 10 CFR, Section 50.55a(f)(6)(i).

D. A visual examination for leks shall be made with the reactor coolant system at pressure during each scheculed refueling outage or after major repairs have been made to the reactor coolant system in accordance with Article

'5000,Section XI. The requirements of specification 3.3.A shall be met during the test.

E. Each replacement safety valve or valve that has been repaired shall be tested in accordance with subsection IWV-3510 of Section XI of the ASME Boiler and Pressure Vessel Code. Setpoints shall be as follows:

Ngmber of Valves Set Points (psip 4 1212112 5 1221112 F. A sample of reactor coolant shall be analyzed at least every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for the purpose of determining the content of chloride ion and to check the conductivity.

OYSTER CREEK 4.3-1 Amendment No.: 82,90,120,150, 151,164,188,

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ae j of valve operability is intended to assure that valve operability and position indication system -

. , = performance does not degrade between refueling inspections. When a vacuum breaker valve is'

' exercised through an opening- closing cycle, the position indicating lights are designed to -

c

function as fellows:

Full Closed -- 2 Green On (Closed to 0.10" open): . 2 Red - OfL Open 0.10" 2 Green - Off

. (010" open to full open) 2 Red - oft During each refueling ouy Mur suppres 'on chamber-drywell vacuum breakers will be

' inspected to assure compw.c have not deteriorated. Since valve internals are designed for a

~ 40 year lifetime,' an inspecaon prograa which cycles through all valves in about 1/10th of the design lifetime is extremely conservative. The alarm systems for the vacuum breakers will be calibrated during each refueling outage. This frequency is based on experience and engineeringjudgernnt.-

Initiating reactor building isolation and ol eration i of the standby gas treatment system to maintain a 1/4 inch of water vacuum, tests the operation of the reactor building isolation

-valves, leakage tightnen of the reactor building and performance of the standby gas treatment i systtm. Checking the initiating sensors and associated trip channels demonstrates the capability for automatic actuation. Performing the reactor building in leakage test prior to refueling demonstrates secondary containment capability prior to extensive fuel handling operations associated with the outage. Verifying the efficiency and operation of charcoal filters once per 18 months givec sufficient confidence of standby gas treatment system performance capability. A charcoal filter efficiency of 99% for halogen removal is adequate.

The in-place testing of charcoal filters is performed using halogenated hydrocarbon refrigerant I which is injected into the system upstream of the charcoal filters. Measurement of the refrigerant concentration upstream and downstream of the charcoal filters is made using a gas l chromatographc The ratio of the inlet and outlet concentrations gives an overall indication of -

the leak tightness of the syvem. Although this is basically a leak test, since the filters have charcoal of kno'vn efficiency and holding capacity for elemental iodine and/or methyl iodide,

the test also gives an indication of the relative efficiency of the installed system. The test procedure is an adaptation of test procedures developed at the Savannah River Laboratory i Lwhich were described in the Ninth AEC Cleaning Conference.*

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- High efficiency particulate filters are installed before and after the charcoal filters to minimize

- potential releases of particulates to the environment and to prevent clogging of the iodine filters. - An efficiency of 99% is adequate to retain particulates that may be released to the reactor building following an accident. This will be demonstrated by testing with DOP at testing medium.

[ D.R. Muhabier, "In Place Nondestructive Leak Ast for lodine Adsorbers," Proceedings of -

l the Ninth AEC Air Cleaning Conference, USAEC Report CONF-660904,1966 OYSTER CRREK '4.5 13 Amendment No.: 52,186

ADMINISTRATIVE CONTROLS 6.1

• RESPQNSlWilD' 6.1.1 The Vice President & Director Oyster Creek shall be responsible for overall facility operation. Those responsibilities delegated to the Vice President & Director as stated in the Oyster Creek Technical Specifications may also be futulled by the Directs.. - Operations and Maintenance. The Vice President &

L Director shall delegate in writing the succession to this responsibility during his and/or the Director -

Operations and Maintenance absence.

6.2 010ANIZAT10N 6.2.1 Comorats 6.2.1.1 An onsite and ofTsite organization shall be established for unit operation and corporate management.

The onsite and ofTsite organiiation shall include the positions for activities affecting the safety of the nuclear power plant.

6.2.1.2 Lines of authority, respo.tsibility and communication shall be established and defmed from the highest management levels through intermediate levels to and including operating organization positions.

These relationships shall be documented and updated as appropriate, in the form of organizational charts. These organizational charts will be documented in the Updated FSAR and updated in accordance with 10 CFR 50.71e.

6.2.1.3 The President - GPUN shall have corporate responsibility for overall plant nuclear safety and shall take measu:es needed to ensure acceptable performance of the stafTin operating, maintaining, and providing technical support in the plant so that continued nuclear safety is assured.

6.2.2 FACILITY STAFF 6.2.2.1 The Vice President & Director Oyster Creek shall be responsible for overall unit safe operation and shall have control over those onsite activities neces ary for safe operation and maintenance of the plant.

6.2.2.2 The facility organization shall meet the following:

a. Each on duty shift shall include at least the following shift staffing:

o One (1) group shift supervisor o Two (2) control room operators o Three (3) equipment operators - one may be a Radwaste Operator

b. At all times when there is fuel in the vessel, at least one licensed senior reactor operator shall be on site and one licensed teactor operator should be at the controls. 1

c. At all times when there is fuel in the vessel, except when the reactor is in COLD SHUTDOWN or REFUEL modes, two licensed senior reactor operators and two licensed reactor operators shall be  ;

on site, with at least one licensed senior reactor operator in the control room and one licensed I reactor operator at the controls, l

l l OYSTER CREEK 6-1 Amendmert No.: 134 i i

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