ML18113A744

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Forwards Proposed Tech Spec Change 71,amending License to Reflect Reorganization of Production Operations Dept Onsite & Offsite, & to Incorporate Several Minor Corrections
ML18113A744
Person / Time
Site: Surry  
Issue date: 12/28/1978
From: Stallings C
VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To: Harold Denton
Office of Nuclear Reactor Regulation
References
NUDOCS 7901020156
Download: ML18113A744 (20)
v • d • e


Text

VIRGINIA ELECTRIC AND POWER COMPANY RIOHMOND, V1RGINI.A. 23261 December 28, 1978 Mre Harold R *. Denton, Director Office of Nuclear Reactor Regulation Attn:

Mre Albert Schwericer, Chief.

Operating Reactors.Branch No. r Divisiin of Reactor Litensing

  • U.*S. Nuclear Regulatory Commissio*n Wash i.ngton, DC *20555

Dear Mr. Denton:

Serial No. 737 LQA/RMN:esh Docket Nos.

50,-280

. 50-281 License Nos. DPR-32 DPR-37

  • AMENDMENT *.ro: OPE RAT I NG *LICENSE*
  • *SURRY' POWER. STAT l ONUN IT* NOS~ 1 'AND 2
  • PROPOSED 'TECHNICAL* SPECIF I CAT I ON: CHANGE* NO~* 71 *
  • ADDITIONAL' INFORMATION**

Pursuant to 10 CFR 50.90, the Virginia Electric and Power Company requested, by letter dated August 31, 1978, Serial No. 500, an amendment in the form of changes to the Technical Specifications to Operating License Nos *.

DPR-32 and DPR-37 for Surry Power Station, Unit Nos. 1 and 2m

  • The proposed changes, designated as Change No. 71 were to reflect a reorganization of the Production o*perations Department onsite and offsite and to* incorporate several minor changes and corrections.

Enclosed is additional information on this change.

The technical specifications concerning the power station onsite o.rga-nization are revised to reflect a reorganization. This reorganization is sche-duled to take effect January 1, 1979.* The new organization* is designed to increase efficiency and therefore enh~nce safeti.

Figure 6.1~2, the offsite organization chart that was sub~itted with our August 31, 1978 letter, has been *improved.

The improved chart is 1 imited to th.ose positions that directly support Surry Power Station's nuclear opera-tions.

Technical specification 6e3-1, 11Action to be Taken if a Safety Limit is Exceeded," is revised to incorporate consistency in the reporttng require-ments.

Exceeding a safety 1 imit is, by the definition in section 2.0 of the technical specffication~, a reportable occ~rrence requiring prompt notification_

with written follow-upe The proposed change would be compatible with this definition.

Technical Specificatiori:6~1~8.5 is revised to increase the fire brigade training and drill interval to 92 days.

This* is consistent with the NRC position as s.ho.wn lntne standard techriicc1l specifications.

Section 1.0 'is revised to delete reference to F.igure 3.12-:-7.

This figure was deleted in amendment 47.

7901020\\~

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e VIRGINIA ELECTRIC AND POWER COMPANY TO Mr. Harold R. Denton 2

Several typographical errors are corrected in sections 3.3, 3.6, 3.8, and 4.17.

These corrections are very minor and do not cha.nge the interpretation of the specification.

This proposed change has. been reviewed and approved by the Station Nuclear Safety and Operatin~ Committee, and the System Nuclear Safety and Op-erating Committee.

It hai been determined that this request does not involve an unreviewed safety question, as defined in 10 CFR 50,.59.

Attachment:

Proposed Technical Specification cc:

Mr. James P. 0 1Reilly, Director Office of Inspections and Enforcement Region I I

e e

VIRGINIA ELECTRIC AND PowER Co1':i:PANY RICHMOND, VIRGINIA 23261 December 28, 1978 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation Attn:

Mr. Albert Schwencer, Chief

  • Operating Reactors Branch No. T Divisi6n of Reactor Licensing
  • U. s. Nuclear Regulatory Commissio*n Wash i_ngton, DC *20555

Dear Mr. Denton:

.. AMENDMENT.* TO: OPERATING ti CENSE

  • _

Serial No. 737 LQA/RMN:esh Docket Nos.

50-280 50-281 License Nos. DPR~32 DPR-37.

. 'SURRY:POWER'STATIQN'.UNIT:Nos~*1 :AND 2

.. PROPOSED 'TECHNICAL. SPECIF I CAT I ON.' CHANGE* NO~: 71

  • . ADD IT I ONAL. INFORMATION Pursuant to 10 CFR 50.90, the Virginia Electric and Power Company requested, by letter dated August 31, 1978, Serial No. 500, an amendment in the form of changes to the Technical Specifications to Operating License Nos.

DPR-32 and DPR-*37 for Surry Power Station, Unit Nos. 1 and 2a

  • The proposed

.changes, designated as Change No. 71 were to reflect a reorganization of the Pr6duction dperations Dep~rtment onsite and offsite ~nd t6 incorporate several minor changes and corrections.

Enclosed is additional info.rmation on this cha_nge.

The technical specifications concerning the power station onsite orga-.

nization are revised to reflect a reorganizati6n*.

This reorganization is sche-du 1 ed* to take effect January 1, 1979; The new organization* is des_i gned to increase efficiency and therefore enhance safety.

Figure 6.1~2~ the offsite organization chart that was submitted with our August 31, 1978 letter, has been *improved.

The improved chart is limited to th.ose positions that directly support Surry Power Station's nuclear opera-tions.

Technical specification 6.3-1, 11Action to be Taken if-a Safety Limit is Exceeded," is revised to incorporate consistency.in the reporting require-ments.

Exceeding a safety limit is, by the definition in section i.o of the technical speci'fication~, a reportable occurrence requiring prompt notification with written follow-up.

The proposed change would be compatible with this definition.

Technical Specification 6~1~s.5 ls revised to increase the fire brigade training and drill interval to 92 days.

This is consistent with the NRC pos.iti*on as sho.wn in the standard technical specifications.

Section*l.O.is revised t6 delete reference to Figure 3.12~7.

This f_igure was deleted in amendment 47.

I.

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e VI ROI NIA ELECTRIC. AXD POWER COMPA1'-Y. TO Mr. Harold R. Denton 2

Several typographical errors are corrected in sections*3.3, 3.6, 308, and 4.17.

These corrections are very minor and do not cha.nge the interpretation of the specification.

This proposed change has been reviewed and approved by the Station*

Nuclear Safety and Operating. Committee, and the System Nuclear Safety a!Jd Op-erating Committee.

It has* been determined that this reques.t does not involve*

an un.rev i ewed safety quest ion, as defined in 10 CFR 500 59.

Attachment:

Proposed Technical Specification cc:

Mr. James*P. 01Reilly, Director Office of Inspections and Enforcement R_eg ion 11

e e

TS 1.0-1 1.0 DEFINITIONS The following frequently used terms are defined for the uniform interpretation of the specifications.

A.

Rated Power A steady state reactor core heat output of 2441 MWt.

B.

Thermal Power The total core heat transferred from the fuel to the coolant.

c.

Reactor Operation,

1.

Refueling Shutdown Condition When the reactor is subcritical by at least 10% !J.k/k and T

  • is avg

<1400F and fuel is scheduled to be moved to or from the reactor core.

2.

Cold Shutdown Condition When the reactor is subcritical by at least 1% flk/kand T is <200°F.

avg

3.

Intermediate Shutdown Condition Wlien the reactor is subcritical by an amount greater than or.equal to 1.77% !J.k/k and 200°F <T

<547°F.

avg*

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D.

e TS 1.0-2

4.

Hot Shutdown Condition When the reactor is subcritical by an amount greater than or equal to 1. 77% t:.k/k. and T is >5470F.

avg

5.

Reactor Critical

6.

Whe~ the neutron chain reaction is self-sustaining and keff = 1.0.

  • Power Operation When the reactor is critical and the neutron flux power range instrumentation indicates greater than 2% of rated power.

Refueling Operation Any operation involving movement of core components when the vessel head is unbolted or removed.

Operable A system or component is operable when it is capable of performing its in-tended function within the required range.

The system or component shall be con*sidered to have this capability when: (1) it satisfies the limiting conditions for operation defined in Section 3, and (2) it has been tested periodically in accordance with Section 4 and meets its performance requirements.

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TS 3.3-6 Basis The normal procedure for starting the reactor is, first, to heat the reactor coolant to near operating temperature by running the reactor coolant pumps.

The reactor is then made critical by withdrawing control rods and/or diluting boron in the coolant.

With this mode of startup the Safety Injection System is required to be operable as specified.

During low power physics tests there is a negligible amount of energy stored in the system; therefore an accident comparable in severity to the Design Basis Accident is not possible, and the full capacity of the Safety Injection System is not required.

The operable status of the various systems and components is to be demonstrated by periodic tests, detailed in TS Section 4.1.

A large :fraction: of these tests are performed while the reactor is operating in the power range.

If a com-ponent is found to be inoperable, it will be possible in most cases to effect repairs and restore the system to full operability within a relatively short time.

A single component being inoperable does not negate the ability of the system to perform its function, but it reduces-the redundancy provided in the reactor design and thereby limits the ability to tolerate additional equipment failures.

To provide maximum assurance that the redundant component(s) will operate if required to do so, the redundant component-Cs) are to be tested prior to initiating repair of the inoperable component and, in some cases are to be retested at intervals during the repair period.

In some cases, i.e. charging pumps, additional components are installed to allow a component to be inoperable without affecting system redundancy.

For those cases which are not so designed, if.it develops that (a) the inoperable componeI).t is not repaired within the

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. TS 3.6-1 3.6.

TURBINE CYCLE Applicability Applies to the operating status of the Main Steam and Auxiliary F~d*

Systems Objective To define the conditions required in the Main Steam System and Auxiliary Feed System for protection of the steam generator and to assure the capability to remove residual heat from the core during a loss of station power.

Specification A.

A unitws Reactor Coolant System temperature or pressure shall not

  • exceed 350°F or 450 psig, respectively, or the reactor shall not be critical unless the five main steam line code safety valves associated I

with each steam generator in unisolated reactor coolant loops, are operable.

B.

To assure residual heat removal capabilities, the following conditions

L Two of the three auxiliary feedwater pumps shall be operable.

C

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TS 3.6-3 450 psig, respectively, residual heat removal requirements are normally satis~

fied by steam bypass to the condenser. If the condenser is unavailable, steam can be released to the atmosphere through the safety valves, power operated relief valves, or the 4 inch decay heat release line.

The capability to supply feedwater to the generators is normally provided by the operation of the Condensate and Feedwater Systems. _In the event of complete loss of electrical power to the station, residual heat removal would continue to be assured by the availability of either the steam driven auxiliary feedwater.

pump or one. of the motor driven auxiliary _feedwater pumps and the 100,000 gallon condensate storage tank.

A minimum of 92,000 gallons of water in the 110,000 gallon condensate tank is sufficient for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of residual heat removal following a* reactor trip and loss of all off-site electrical power. If the protected condensate storage tank level is reduced to 60,000 gallons, the immediately available replenishment water in the 300,000 gallon condensate tank can be gravity-feed to the pro-tected tank if required for residual heat removal. An alternate supply of feed-water to the auxiliary feedwater pump suctions i~ also available from the Fire Protection System Main in the auxiliary feedwater pump cubicle.

The five main steam code safety valves associated with each steam generator have a total combined capacity of 3,725,575 pounds per hour at their tndividual set pressure; the total combined capa~ity *of all fifteen main steam code safety valves is 11,176,725 pounds per hour. The ultimate power rating steam flow is 11,167,923 pounds per hour. The combined capacity of the safety valves required by Specification 3.6 always exceeds the total steam flow corresponding to the maximum steady-state power t;han can be obtained during one, two or*.:t°Q.ree reactor

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C TS 3. 8-2

4.

The containment integrity shall not be violated when the reactor vessel head is unbolted unless a shutdown margin greater than 10 per*cent A k/k is maintained.

5.

Positive reactivity changes shall not be made by rod drive motion_

or boron dilution unless the containment integrity is intact.

B.

Internal Pressure

1.

If the internal air partial pressure rises to a point 0.25 psi

  • .above the preset value of :the* *air partial pressure (TS Figure.- 3~8-1), I
  • the reactor shall be brought to the hot shutdown condition.
2.

If the leakage condition cannot be corrected without violating the containment integrity or if the internal partial pressure continues to rise, the reactor shall be brought to the cold shutdown condition utilizing normal operating procedures.

3.

If the internal pressure falls below 8.25 psia the reactor shall be placed in the cold shutdown condition.

Basis The Reactor Coolant System temperature and *pressure being below 350°F and 450 psig, respectively, ensures that no significant amount of flashing steam will be formed and hence that there would be no significant pressure buildup in the containment if there is a loss-of-coolant accident.

TS 3.13-2 heat exchangers shall be operable.

2.

For two unit operation, three component cooling water pumps and heat exchangers shall be operable.

3~

The Component Cooling Water Subsystem shall be operable for immediate supply of cooling water to the following components, if required:

a.

Two operable residual heat removal heat exchangers.

  • . -... *.1*1',
4.

During power operation, Specification A-1, A-2, or A-3 above may.

be modified to allow one of the required components to be inoperable provided immediate attention is directed to making repairs.

If the system is not restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> _to the requirements of Specification A-1, A-2, or A-3, an operating reactor shall be placed in the hot shutdown condition.

If the repairs are not completed within an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the affected reactor shall be placed in the cold shutdown condition.

5.

Whenever the component cooling water radiation monitor is inoperable, the surge tank vent valve shal 1 remain closed.

B.

For each unit whose Reactor Coolant System exceeds a temperature of 350°F and a pressure of 450 psig, or when a unit's reactor is critical, I

TS 4.17-6 e

e LEGEND ACCESSIBILITY CATEGORY A= Accessible I= Inaccessible RADIATION CATEGORY H = High radiation area only during periods of reactor operation.

In acceptable radiation work area during periods of reactor shutdown.

N = Acceptable radiation work area during period of both reactor operation and shutdown.

REMOVAL CATEGORY D = Difficult to remove, i.e., large line size, large component, physical location (overhead), lines under load, jacks necessary R = Can be removed

TS 6.1-1 6.0 ADMINISTRATIVE CONTROLS 6.1 ORGANIZATION, SAFETY, AND OPERATION REVIEW Specification A.

The Station Manager shall be responsible for the safe operation of the facility.

The Station Manager shall report to the Director-Nuclear Operations.

The relationship between this Director and other levels of company management is shown in TS Figure 6.1-1 and 6.1-2.

B.

The Station organization shall conform to the chart as shown in TS Figure 6~1-3.

1.

Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N.18.1-1971 for comparable positions, except for the Supervisor-Health Physics who shall meet or exceed the quali-fications of Regulatory Guide 1.8, September 1975.

e TS 6.l-2b

5.

A training program for _the fire brigade and fire teams shall be main-tained under the directions cif a Fire Marshall and shall meet or exceed-the requirements of the NFPA Code Section 27 (1975), except that training a~ssions and drills shall be held at least once per 92 days, C.

Organization units to provide a continuing review of the.operational and*

safety aspects of the nuclear facility shall be constituted and have the authority and responsibilities outlined below:

1.

Station Nuclear s*afety and Operation Committee

a.

Membership

.(1)

Chairman - Station Manager (2)

  • Vice Chairman - Superintendent Operations.
  • 1 l

TS. 6.1..

3.

Member - Superintendent-Maintenance

4.

Member - Superintendent-Technical Services

b.

Qualifications:

The qualifications of the regular members.

of the Station Nuclear Safety and Operating Committee with regard to the combined experience and technical specialties of the individual members shall be maintained at a level at least equal to those described in Section 6.1.B.l of the Specifications.

c.

Meeting frequency:

As called by the Chairman but not less than monthly.

d.

Quorum:

Chairman or Vice Chairman, and two others to provide a quorum of three members.

The Chairman or Vice Chairman may appoint a similarly qualified designee to represent a member other than the Chairman or Vice Chairman on a temporary basis.

No more than two alternates shall participate as voting members in SNSOC activities at any one time.

e.

Responsibilities

1.

Periodically review all proposed normal, abnormal, and emergency operating procedures and all proposed maintenance procedures.

Review proposed changes to those procedures,.

and any other proposed procedures or changes thereto as determined by the Station Manager which affect nuclear safety.

I DIRECTOR PRODUCTION TRAINING*

& SAFETY I

NUCLEAR TRAINING s_uprnv I SOR".' SURP_V

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VICE PRESIDENT*

POWER SUPPLY &

PRODUCTION OP~RATIONS I

MANAGER PRODUCTION OPERATIONS I,

DfRECTOR PRODUCTION OPER.

& MAINTENANCE SUPPORT r

(SUPERINTENDENT~

  • \\

PRODUCTION II OPERATIONS

~

I COORDINATOR PROOUCTION

.._g(:.,URITY SUPERVISOR NUCLEAR OPERATIONS I

I SUPER I NT EN DENT

. 'MATERlflLS MANAGD'.ENT

'SUPERINTENDENT!

MAINTENANCE SERVIC'ES

DIRECTOR NUCLEAR OPERATIONS I

I 1 STATION

, MANAGER DIRECTOR PRODUCTION TECHNICAL SUPPORT 1.----/--'*----

SUPER I NTENDENT TECHNICAL SERVICES SUPER I NTEtJDENT ENG I NEER 11:G SERVICES I

.SYSTEM HtAL Tl!

PIIYSJCIST I

I SYSTEM CHEMISTRY SUP ERV I SOR Offsite Organization for Faci"lity Management and Technical Support

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I N

DIRECTOR PRODUCTION TRAINING

& SAFETY_

TRAINlNG SUl'~:RV J.SOR SUPERINTENDENT

~I.!_~~PERATIONS SL

-SL OPF.RATING SUPEIW I SOil SlllFT sun:RVISOR ASS1S1'ANT SHIFT J SUPllRVISOR SL OL CONTROi. ROOM OPERATOR ASSISTANT CO!l1'ROL ROOM OPERATOR

.OL AUXILAR'i OPERATOR STATION NUCLEAR SAFETY & OPERATING COMMITTEE" DIRECTOR NUCLEAR OPERATION STATION MANAGER SUPERINTENDENT

-~-~_!~B~~E __

MECHANICAL SUl'llRVISOR MAIN'l'ENANCll COORDINATOR ELl'.CTRICAL SUPERVISOR SURRY POWER STATION ORGANIZATION CHART SUPERINTENDENT TECHNICAL SERVICES SUPERVISOR UEALTII PHYSICS SUPERVISOR

-~IEMISTR'.f INSTRUHENT SUPERVISOR ENGINEERillG SUPERV_l_SOR_

LEGEND SUPERVISOR QA OPERATIONS &

MAINTENANCE_

RESIDENT QUALITY CO:-ITJ:OL ENGINf.llR SUPERVISOR ADMINISTRATIVE SERVICES FIRE MAIISIIALL STATION SECURITY SUPERVISOR SL-Senior License OL-Operator's License

Communications e

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e e TS 6.3-1 6.3 ACTION TO BE TAKEN IF A SAFETY LIMIT IS EXCEEDED Specification A.

Should a safety limit (see Section 2.0 of the Technical Specifications) be exceeded, the reactor shall be shutdown a_nd reactor *operation shall only be resumed in accordance with the authorization within 10 CFR 50.36 (c)(l)(i),

B.

An immediate report of the incident shall be made to the Station Manager, Director~ Nuclear Operations and the Chairman of the System

c.

Nuclear Safety and Operating Committee.

The Station Manager shall promptly report the circumstances to the NRG as specified in Section 6.6 of these Specifications.

D.

A complete_ analysis of the incident together with recommendations to prevent recurrence shall be prepared by the Shift Supervisor and the Operating Supervisor.

A preliminary written report shall be reviewed and approved by the Station Nuclear Safety and Ope~ating Connnittee.

The final report shall be forwarded to the Director - Nuclear Operations, the Manager - Production Operations and the Vice Presioent ~*

Power Supply and Production Operations.

Appropriate analyses or reports will be submitted to the NRC as specified in Section 6.6 of these specifications.

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e TS 6.4-7a D.

All procedures described in A and b above shall be followed.

E.

Temporary changes to procedures described in A and B above which do F.

not change the intent of the original procedure may be made, provided such changes are approved prior to implementation by the person designa-ted below based on the type of procedure to be changed:

.1.

Administrative

2.

Abnormal

3.

Annunciator

4.

Health Physics

5.

Emergency

6.

Electrical Maintenance

7.

Mechanical Maintenance

8.

Operating

9.

Periodic Test

10.

Start-up Test lL Special Test

12.

Quality Assurance

13.

Chemistry Station Manager Shift Supervisor Shift Supervisor

~Health Physicists Shift Supervisor

  • Electrical Foreman

~Mechanical Foreman Shift Supervisor 1.Cogµ_:izan.t: supervisor

  • Engineering-Supervisqr.

~gineering Supervisor..

Resident Quality Control Engineer

  • Chemist
  • These procedures must have the approval of a llcensed Senior-Reactor Operator.

Such changes will be documented and subsequently reviewed by the Station Nuclear Safety and Operating Committee and approved by the Station Manager within seven days.

Temporary changes to procedures described in A and B above which change the intent of the original procedure may be made, provided such changes are approved prior to implementation by the person designated below based on the type of procedure to be changed.

,\\,

TS 6.4-7b e

1.

Administrative Station Manager

2.

Abnormal Operating Supervisor or Superintendent - Operations

3.

Annunciator Operating Supervisor or Superintendent-Operations

4.

Health Physics Supervisor-Health I'hysics

5.

Emergency Operating Supervisor or Superintendent-Operations

6.

Maintenance Mechanical Supervisor Electrical Supervisor Instrument Supervisor

7.

Operating Operating Supervisor Superintendent - Operations

8.

Periodic Test Engineering Supervisor

9.

Start-up Test Engineering Supervisor

10.

Special Test Engineering Supervisor

11.

Quality Assurance Resident-Quality Control Engineer

12.

Chemistry Supervisor-Chemistry Such changes will be documented and subsequently reviewed by the Station Nuclear Safety and Operating Committee and approved by the Station Mananger.

G.

In cases of emergency, operations personnel shall be authorized to depart from approved procedures where necessary to prevent injury to personnel or damage to the facility.

Such changes shall be documented and reviewed by the Station Nuclear Safety and Operating Committee and approved by the Station Manager.

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