ML20236T481

Proposed Tech Specs Re Changes to Administrative Controls

ML20236T481
Person / Time
Site:	Oyster Creek
Issue date:	07/21/1998
From:	GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20236T463	List: ML20236T461 ML20236T475 ML20236T481
References
NUDOCS 9807280209
Download: ML20236T481 (15)

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REVISED TECIINICAL SPECIFICATION PAGES l l

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9807280209 990721 PDR ADOCK 05000219 P PDR ,,

ADMINISTRATIVE CONTROLS 6.1 .

RESPONSIBILI'D' 1 4

l 6.1.1 The Vice President & Directot Oyster Creek shall be responsible for overall fudity operation. l Those responsibilities delegated to the Vice President & Director as stated in the Oyster Creek j Technical Specifications may also be fulfilled by the Director - Operations and Maintenance. l The Vice President & Director shall delegate in writing the succession to this responsibility I during his and/or the Director - Operations and Maintenance absence.

l 6.2 ORGANIZATION 6.2.1 Corporate 6.2.1.1 An onsite and offsite organization shall be established for unit operation and corporate management. The onsite and olTsite organization shall include the positions for activitics affecting the safety of the nuclear power plant.

6.2.1.2 Lines of authority, responsibility and communication shall be established and defined from the highest management levels through intermediate levels to and including operating organization positions. These relatio iships 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 - GPU Nuclear shall have corporate responsibility for overall plant nuclear safety and shall take measures needed to ensure acceptable performance of the ,

staffin operating, maintaining, and providing technical support in the plant so that l 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 activitics necessan 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 sta1Tmg:

. One (1) group shift supenisor

. Tw (2) control room operators

. Three (3) equipment operators - one may be a Radwaste Operator

. One (1) Shift Technical Advisor (sce h. below)

Except for the group shift supervisor, shift crew composition may be one less than the minimum requirements, for a period of time not to exceed two hours, in order to accommodate unexpceted absence of on-duty shift crew members.

Immediate action must be taken to restore the shift crew composition to within requirements gisen above. This provision does not permit any shill crew position to be unmanned upon shift change due to an incoming shill crew member being ime or absent.

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

OYSTER CREEK 6-1 Amendment No.: 54,59,65,69,102,134,194,195 i

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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 reactor operator at the control (.

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d. At least two licensed reactor operators shall be in the control room during all reactor startups, shutdowns, and other periods imching planned control rod i manipulations 1

e. All CORE ALTERATIONS shall be directly supervt.ed by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.

f. An individual qualified in radiation protection measures shall be on site when fuelis in the reactor,

g. (deleted)

h. Each on duty shift shall include a Shift Technical Advisor except that the Shift .

Technical Advisors position need not be filled if the reactor is in the refuel or shutdown mode and the reactor is less than 212 F.

i. Administrative procedures shall be dewloped and implemented to limit the working hours of unit staffwho perfonn safety related functions.

In the event that unforeseen problems require substantial amounts ofovertime to be used or during extended periods of shutdown for refueling, major maintenance or major plant modifications, on a temporary basis, the following guidelines shallbe followed:

a. An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> straight, excluding shift tumover time.

b. An individual should not be permitted to w>rk more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 24-hour period, nor more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48-hour period, j nor more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any seven-day period, all excluding shift i tumover time.  !

c. A break of at least eight hours rhould be allowed between work penod,imhdy shift turnover time. 1

d. In a, b, and c above, the t:me required to complete shift tumover is to i a

be counted as break time and is not to be counted as work time. i

e. Except during extended shutdown periods, the use ofovertime should l be considered on an individual basis arul not for the entire staffon a

' shift.  !

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OYSTER CREEK- 6-2 Amendment No.: 92,102,134,16!

4 l7 ' Any deviation from the above guidelines shall be authorized by the

!- Department Managers, or higher levels of management, in accordance with established procedures and with documentation of the basis for granting the desiatim.

j. The Plant Operations Director and the Group Shut Supenisor l

require Senior Reactor Operators licenses The Lantrol Room j Operators require a Reactor Operators license.

I 6.2.2.3 Individuals who train the operatmg staff and those who carry out the health physics and quality assurance function shall have sufficient orgaruzational freedom to be -

independent of operational pressures, however, they may report to the appropriate manager on site.

6.3 Emility StaffOualifications l

6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of ANSI /ANS 3.1 of 1978 for comparable positions unless otherwise noted in the Technical Specifications.

Licensed operators shall meet the supplemental requirements specified in Sections A and C of Enclosure 1 of the March 28,1980 NRC letter to all licensees. Technicians and maintenance personnel who do not meet ANSI /ANS 3.1 of 1978, Section 4.5, are pennitted to perform work for which qualificatim has been demonstrated l

[ 6.3.2 The management position responsible for radiological controls shall meet or exceed the

[ qualifications of Regulatory Guide 1.8 (Rev.1-R, 9/75). Each other member of the radiation l~ pmtection organization for which there is a comparable position described in ANSI N 18.1-1971 ll shall meet or exceed the muumum palifications specified therein, or in the case ofradiation j protection technicians, they shall have at least one 3 car's contmuous experience in applied L radiation protection work in a nuclear facility dealing with radiological problems similar to l those encountered in nuclear power stations and shall have been certified by the management L . position responsible for radiological controls as qualified to irrfonn assigned functions. This certification must be based on an NRC approved, documented program consisting ofclassroom training with appropriate eununations and documented positive fmdings by responsible supenision that the individual has demonstrated his ability to perform each specified procedure and assigned function with an understandmg ofits basis and purpose.

l l 6.3.3 The Shift Technical Advisors shall have a bachelor's degree or equivalent in a scientific or l crgirwring discipline with specific training in plant design, response and analysis of the plant for transients and accidents.

OYSTER CREEK 6 2a Amendment No.: 134

6.4 TRAINING 6.4.1 A retraining program for operators shall be maintamed under the direction of the Manager l responsible for plant tranung and shall meet the requirements and recommendation of 10 CFR Part 55. Replacement training programs, the content of which shall meet the requirements of 10 CFR Part 55, shall be conducted under the direction of the Manager responsible for plant training for licensed operators and Senior Reactor Operators.

6.5 REVIEW AND AUDIT 6.5.1 TECHNICAL REVIEW AND CONTROL The Corporate Officers of GPU Nuclear, Inc. shall be responsible for ensuring the preparation, l resiew, and approval ofdocuments required by the activities described in 6.5.1.1 through 6.5.1.5 within his functional area of responsibility as assigned in the GPU Nuclear Resiew and i Approval Matrix. Implementing approvals shall be pedormed at the cognizant manager level or abow.

ACTIVITIES 6.5.1.1 Each procedure required by Technical Specification 6.8 and other procedures which affect nuclear safety, and substantive changes thereto, shall be prepared by a  ;

designatal individual (s)/ group knowledgeable in the area affected by the procedure. I Each su:h procedure, and substantive change thereto, shall be resiewed for adequacy .

by an individual (s)/ group other than the preparer, but who may be from the same  !

division as the individual who prepared the procedure or change.

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OYSTER CREEK 6-3 Amendment No.:69,78,125,134,161,194 1

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.i RECORDS 6.5.1.13 Written records of activities performed under specifications 6.5.1.1 through 6.5.1.11 1 shallbe mamtamed

' QUALIFICATIONS 6.5.1.14 Responsible Technical Reviewers shall meet or exceed the qualifications of ANSI /ANS 3.1-1978 Section 4.6 or 4.4 for applicable disciplines or have 7 years of appropriate experience in the field ofhis specialty. Credit towards experience will be given for .

advanced degrees on a one-for-one basis up to a maximum of two > tars. These Reviewers shall be designated in writing.

'6.5.2 INDEPENDENT SAFETY REVIEW FUNCTION 6.5.2.1 The Corporate Officers of GPU Nuclear, Inc. shall be responsible for ensuring the l penodic independent safety review of the subjects described in 6.5.2.5 within his assigned area of safety review responsibility, as assigned in the GPUN Review and Approval Matrix.

6.5.2.2 Independent safety review shall be completed by an individual / group not having direct responsibility for the performance of the actisities under resiew, but who may be from the same functionally cognizant organization a.s the individual / group performing the original work.

6.5.2.3 GPU Nuclean. ;w sbil collectively have or have access to the experience and competence required to independently resiew subjects in the following areas:

a. Nuclear power plant operations

b. Nuclear engmeenng

c. Chemistry and radiochemistry

d. Metallurgy

e. Nondestructive testmg i

. f. Instrumentation and control - {

g. Radiological safety

h. Mechanical engmmring

i. Electrical ergirwing

j. Admmistrative controls and quality assurance practices

k. Emergency plans and related organi7ation, procedures and equipment

1. . Other appropriate fields associated with tie unique characteristics of Oyster Creek i

i 6.5.2.4 Cm= hts may be utilized as determined by the cogruzant Corporate Officer to l provide expert ad. ice.

l OYSTER CREEK 6-5 Amendment No.: 49,134,181,194 i l

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RESPONSIBILITIES 6.5.2.5 The following subjects shall be independently raiemxi by the functionally assigned divisions

a. Written safety evaluations of changes in the facility as described in the Safety Analysis Report, of changes in p' rocedures as described in the Safety Analysis Report, and of tests or expenments not described in the Safety Analysis l

Report, which are completed without prior NRC approval under the provisions of 10 CFR 50.59(a)(1). This review is to verify that such changes, tests or experiments did not involve a change in the Technical Specifications or

! an unresiewed safety question as defmed in 10 CFR 50.59(a)(2). Such L resiews need not be performed prior to implementation. j

b. Proposed changes in procodures, proposed changs in the facility, or proposed tests or experiments, any of which invohrs a change in the Technical Specifications or an untmiewed safety question as defined in 10 CFR i 50.59(c). Matters of this kind shall be resiewed prior to submittal to the NRC.

L c. Proposed changes to Technical Specifications or license amendments related

! to nuclear safety shall be reviewed prior to submittal to the NRC for approval.  ;

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d. Violations, desiations, and reportable events which require reporting to the NRC in writing. Such reviews are performed after the fact. Rmiew of events

(' covered under this subsection shall include results of any investigations made j ,

and the recommendations resulting from such imestigations to prevent or j i

reduce the probability of recurrence of the event. l 1

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e. Written summaries of audit reports in the areas specified in section 6.5.3 and invohing safety related functions. l l

i l f. Any other matters invohing safe operatiore of the nuclear }ower plant which j a raiewer deems appropriate for consideration, or which is referred to the L mdependent raiewers.

QLIALIFICATIONS l 6.5.2.6 The indq=L nt raiewer(s) shall either have a Bachelor's Degree in Engineering or the Physical Sciences and tive (5) years of professional level exph.ience in the area being reviewed or have 9 years of appropriate experience in the field ofhis spe d alty. An individual performing reviews may possess competence in more than one .pecialty L

cxa. Credit toward experience will be given for advanced degmes on a one-for-one l basis up to a maximum of two years.

RECORDS 6.5.2.7 Reports of reviews encompassed in Section 6.5.2.5 shall be prepared, maintaimxi and tr=cminul to the cogmzant Corporate Officer.

l OYSTER CPIEK 6-6 Amendment No.: 69,134

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l 6.5.3.3 Audit reports encompassed by sections 6.5.3.1 and 6.5.3.2 shall be forwardal for l action to the management positions responsible for the areas audited within 30 days l after completion of the audit. Upper managemmt shall be informed per the Operation Quality Assurance Plan. ,

l 6.5.4 ' INDEPENDENT ONSITE SAFE'IY REVIEW GROUP QQSR_Q) l I

STRUCT1EE 6.5.4.1. 'Ihe IOSRG shall be a full-time group ofengineers experienced in nuclear power plant ergiruing, operation and/or technology, independent of the facility staff, and located onsite.-

i ORGANIZATION 6.5.4.2 a. ~ The IOSRG shall consist of a Manager responsible for Nuclear Safdy l Assessment and staff members who meet the qualifications of 6.5.4.5. Group )

expertise shall be multidisciplined.  !

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b. The IOSRG shall report to the Director responsible for Nuclear Safi:ty ~ l Assessment. j FUNCTION 6.5.4.3 The periodic review functions of the IOSRG shall include the following on a selective ,

and overview basis:

1) Evaluation for technical adequacy and clarity of procedures important to the safe operation of the facility.

2) Evaluation of facility operations from a safety perspective.  !

3) Assessment of facility nuclear safety programs

4) Assessment of the facility perfonnance regardmg conformance to requirements related to safi:ty.

5) Any other matter invohing safe operation of the nuclear power plant that the manager deems appropriate for consideration.

AlJfHORITY 6.5.4.4 The 10SRG shall have access to the facility and facility records as necessary to l perform its evaluations and assessments Based on its reviews, the IOSRG shall  !'

i proside reumoucalations to the management positions responsible for the areas reviewed.

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OYSTER CREEK 6-8 Amendment No.:69,108,134

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l. QUALIFICATIONS V

6.5.4.5 IOSRG engineers shall have either (1) a Bachelor's Degree in Engineering or i appropriate Physical Science and three years ofprofessional level experience in the nuclear power field which may include technical supponing fimetions or (2) eight years ofappropriate experience in nuclear power plant operations and/or tecimology. Credit l toward experience will be given for advance degrees on a one-tome basis up to a

! maximum of two years.

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6.5.4.6 Reports ofevaluations and assessments encompassed in Section 6.5.4.3 shall be

! prepared, approved, and transmitted to the Director and the Corporate Officer l l

! ' responsible for nuclear safety assessment, Vice President & Director Oyster Creek, and

, the nunagement positions responsible for the areas reviewed.

' '66 REPORTABLE EVENT ACTION f

L 6.6.1 The following actions shall be taken for REPORTABLE EVENTS:

. a. The Commission shall be notified and a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50;and

b. Each REPORTABLE EVENT shall be reponed to the cognizant manager and the l cogmzant division Vice President and the Vice President & Director Oyster Creek.

l The functionally meninnt division staffshall prepare a Licensee Emnt Report (LER) I in accordance with the guidance outlined in 10 CFR 50.73(b). Copics of all such

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l reports shall be submitted to the functionally cognizant Corporate Officer and the Vice q p President & Director Oyster Creek. J I

6.7 SAFETY LIMITVIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is siolated:

a. If any Safety Limit is exceeded, the reactor shall be shut down immediately until the Commission authorizes the resumption of operation.

! b. The Safety Limit violation shall be reported to the Commission and the Vice President  ;

I. and Director Oyster Creek. I i

c. A Safety Limit Violation Report shall be prepared The report shall be submitted to i the Vice President and Director Oyster Creek.' This report shall describe (1) applicable circumstances precedmg the violation, (2) effects of the violation upon facility components systems or structur- (3) corrective action taken to prevent recurrence.

d. The Safety Limit Violation Report shall be submitted to the Comnussion within ten days ofthe violation. I OYSTER CREEK 6-9 Amendment No.: 69,78,84,117,134

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REVISED TECHNICAL SPECIFICATION BASES PAGES l l

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4 The low level water level trip setting of I l'5" above the top of the active fuel has been established to assure that the reactor is not operated at a water level below that for which the fuel cladding integrity safety limit is applicable. With the scram set at this point, the generation of steam, and  !

thus the los's ofinventory 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 ofwater remaining above the core after isolation (6). The TAF definition of 353.3 inches from vessel zero is based on a fuel length of 144 inches and it is applicable to the current fuel length of 145.24 inches. l 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 i 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 flux increase caused by the rapid closure of the turbine stop 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 turbine trip.

The undervoltage protection system is a 2 out of 3 coincident logic relay system designated to l shift emergency buses C and D to on-site power should normal power be lost or degraded to an unacceptable level. The 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 functioung of engineered safety features connected to the plant emergency power distribution system.

References (1) FDSAR, Volume 1, Section VII-4.2.4.2 (2) FDSAR, Amendment 28, item Ill.A-12 (3) FDSAR, Amendment 32, Question 13 (4) Letters, Peter A. Morris, Director. Divh;icr. Of Reaction Licensing, USAEC, to John E.

Logan, Vice President, Jersey Cemral Power and Light Company (5) FDSAR, Amendment 65, Section B.XI (6) FDSAR, Amendment 65, Section B.1X t

OYSTER CREEK. 2.3-7 Amendment No. 4-757 495

• Correction 11/30/87 l

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Transformation temperature. The minimum temperature for pressuriz2 tion s.t any time in life has to i account for the toughness properties in the most limiting regions of the reactor vessel, as well as the effects of fast neutron embrittlement.

Curves A, B and C on Figures 3.3.1,3.3.2 and 3.3.3 are derived from an evaluation of the fracture toughness properties performed on 'M specimens contained in Reactor Vessel Materials Surveillance Program Capsule No. 2 (Reference 14j. The results of dosimeter wire analyses (Reference .t4) indicated that the neutron fluence (E>l.0 MeV) at the end of 32 effective full power years of operation is 2.36 x 1018 n/cm2at the 1/4T(T= vessel wall thickness) location. This vabe was used in i

the calculation of the adjusted reference nil-ductility temperature which, in turn, was used to generete l the pressure-temperature curves A, B and C on Figures 3.3.1,3.3.2 and 3.3.3 (Reference 15). The )

1 250 F maximum pressure test temperature provides ample margin against violation of the minimum  !

( required temperature. Secondary containment is notjeopardized by a steam leak during pressure testing, and the Standby Gas Treatment system is adequate to prevent unfiltered release to the stack.

Stud tensionmg is considered significant from the standpoint of brittle fracture only when the preload exceed approximately 1/3 of the final design value. No vessel or closure stud minimum temperature L requirements are considered necessary for preload values below 1/3 of the design preload with the l vessel depressurized since preloads below 1/3 of the design preload result in vessel closure and l

average bolt stresses which are less than 20% of the yield strengths of the vessel and bolting materials. Extensive service experience with these materials has confirmed that the probability of brittle fracture is extremely remote at these low stress levels, irrespective of the metal temperature.

l The reactor vessel head flange and the vessel flange in combination ivith the double "O" ring type j seal are designed to provide a leak tight seal when bolted together. When the vessel head is placed on l the reactor vessel, only tnat portion of the head flange near the inside of the vessel rests on the vessel 1 flange. As the head bolts are replaced and tensioned, the vessel head is flexed slightly to bring together the entire contact surface adjacent to the "O" rings of the head ano vessel flange. The original Code requirement was that boltup be done at qualification temperat ures (T3OL) plus 60* F.

l Current Code requirements state (Ref.16) that for application of full bolt preload and reactor pressure l up to 20% of hydrostatic test pressure, the RPV metal temperature must be at RTNDT or greater. The boltup temperature of 85* F was derived by determining the highest value of(T3OL + 60) and the highe>t value of RTNDT, and by choosing the more conservative value of the two. Calculated values of(T3OL + 60) and RTNDT of the RPV metal temperature were 85 F and 36* F, respectively (Ref. l

! 15). Therefore, selecting the boltup temperature to be 85* F provides 49 F margin over the current Code requirement based on RTNDT-Detailed stress analyses (4) were made on the reactor vessel for both steady state and transient conditions with respect to material fatigue. The results of these analyses are presented and compared to allowable stress limits in Reference (4). The specific conditions analyzed cirrently include 240 cycles (17) of noitnd startup and shutdown with a heating and cooling rate of 100* F per hour applied continuously over a temperature range of 100* F to 546* F and for 10 cycles of emergency cooldown at a rate of 300 F per hour applied over the same range. A review of the original analysis shows that the components with the highest fatigue usage factor are the reactor vessel studs and reactor vessel basin seal skirt. These components have the potential to exceed the allowable fatigue usage factor if the number of thermal cycles (i.e., heatup/cooldown) exceed design assumptions. The number of heatup and cooldown cycles was reanalyzed, as documented by Reference (17), for a higher number of cycles (240) than expected in the original analysis (120). The reanalysis confirmed that the original fatigue usage factor limit of 0.8 is maintained. All other components have relatively low usage factors'and are not expected to exceed the fatigue usage factor limit of 0.8 for the design life of 40 years. Thermal stresses from this analysis combined with the primary load OYSTER CREEK 3.3-5 Amendment No: 15,12,12^,!S!,188

References:

(1) FDSAR, Volume I, Seci . IV-2

• (2) Letter to NRC dated May 19,1979, " Transient of May 2,1979" (3) General Electric Co. Letter G-EN-9-55, " Revised Natural Circulation Flow Calculation", dated May 29,1979 (4) Licensing Application Amendment 16, Design Requirements Section

( (5) (Deleted)

(6) FDSAR, Volume I, Section IV-2.3.3 and Volume II, Appendix 11 l_ .(7) FDSAR, Volume 1, Table IV-2-1 l -(8) Licensing Application Amendment 34, Question 14 j (9) Licensing Application Amendment 28, Item III-B-2 (10) Licensing Application Amendment 32, Question 15 (II) (Deleted)

(12) (Deleted)

(13) Licensing Application Amendment 16, Page 1 (14) GPUN TDR 725 Rev. 3: Testing and Evaluation ofIrradiated Reactor Vessel Materials Surveillance Program Specimens (15) - GENE-B13-01769 (GE Nuclear Energy): Pressure-Temperature Curves Per Regulatory Guide 1.99, Revision 2 for Oyster Creek Nuclear Generating Station.

(16) Paragraph G-2222(C), Appendix G,Section XI, ASME Boiler and Pressure Vessel Code,1989 Edition with 1989 Addenda, " Fracture Toughness Criteria for Protection

. Against Failure."

(17) GPUN Safety Evaluation, SE-000221-004, " Reactor Vessel Thermal Cycles".

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l OYSTER CREEK ' 3.3-8a Amendment No: 135, 90, !S!, I88

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- As indicated in Am:ndm:nt 18 to ths Licensing Application, th:re are numtrous sourc:s of diesel fuel which c"n be obtained within 6 to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and the heating boiler fuel in a 75,000 gallon tank on the site could also be used. As indicated in Amendment 32 of the

. Licensing Application and including the Security System loads, the load requirement for the loss ofoffsite power would require 12,410 gallons for a three day supply. For the case ofloss of offsite

, power plus loss-of-coolant plus bus failure 9790 gallons would be required for a three day supply.

In the case ofloss of offsite power plus loss-of-coolant with both

- diesel generators starting the load requirements (all equipment operating) shown there would not be three days' supply. However, not all of this load is required for three days and, after evaluation of the conditions, loads not required on the diesel will be curtailed. It is reasonable to expect that within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />

- conditions can be evaluated and the following loads curtailed:

1. One Core Spray Pump

2. One Core Spray Booster Pump

3. One Control Rod Drbe Pump

4. One Containment Spray Pump i

5. 'One Emergency Service Water Pump 1

- With these pieces of equipment taken off at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the incident it would require a total consumption of 12,840 gallons for a three day supply. Therefore, a minimum technical specification ,

requirement of 14,000 gallons of diesel fuel in the str.adby diesel generator fuel tank will exceed the engineered safety features operational requirement after an accident by approximately 9%.

During plant cold shutdown or refueling, it may be necessary to inspect, repair and replace the 15,000 gallon standby diesel generator fuel storage tank. This would require tank partial or full drain down. An alternate fuel supply configuration may be established which consists of temporary tanker trucks capable of containing 14,000 gallons. This configuration is capable of supporting continuous operation of both diesels for at least 3 days.

The temporary configuration is acceptable since a minimal power load would be required during and t'owing a desiga basis condition of a

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loss of offsite power while the @t is in cold shutdown or

• refueling. Analysis shows that in the event of a tornado or sek ic event which may cause a loss of offsite power and a temporary loss of the temporary EDG fuel oil supply, power can be restored before the consequences of pw >lously analyzed conditions are exceeded.

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References:

1 (1) . Let'er, Ivan R. Finfrock, Jr. to the Director of Nuclear Reactor Regulation dated Aprii 4,1978.

l OYSTER CREEK - 3.7-4 Amendment No.: 148 l

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~ Each condenser containing a minimum total water volume of 22,730 gallons provides 11,060 gallons above the condensing tubes. Based on scram from a reactor power level of 1950 MWt (the design basis power level for the isolation condensers) the condenser system can accommodate the reactor decay heat (2,3) (corrected for U-239 and NP-239) i 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 of 22,730 gallons can accommodate the reactor decay heat for 45 l minutes after scram from 1950 MWt before makeup water is required. In order to accommodate a scram from 1950 MWt and cooldown, a total of 107,500 gallons of makeup water would be required either from the condensate storage tank or from the fire protection system. . Since the rated reactor power is 1930 MWt, the above calculations t represent conservative estimates of the isolation condenser system capability.

l l The vent lines from each of the isolation condenser loops to the main steam lines l: downstream of the main steam lines isolation valves are provided with isolation valves l which close automatically on isolation condenser actuation or on signals which close the main steam isolation valves. High 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.E allows reduction in redundancy ofisolation capability for isolation condenser 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 valve as this valve is normally in the closed position. These specifications permit troubleshooting and repair as well as routine 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 service 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 primary containment isolation valves.(5)

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 l a one out of two system.(4) The test interval for operability of the valves required to place the isolation condenser in opccation is once/ month (Specification 4.8). An scceptable out of service time, T, is then determincd to be 10 days. However, 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 verified operable 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.: 72,195