ML17297B534

Forwards Amend 3 to 820406 TMI-2 Lessons Learned Implementation Rept,Addressing NUREG-0737 Items

ML17297B534
Person / Time
Site:	Palo Verde
Issue date:	05/24/1982
From:	Van Brunt E, Vanbrunt E, Zan Brunt E ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
To:	Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-1.A.1.1, TASK-1.A.1.3, TASK-1.B.1.2, TASK-1.C.5, TASK-2.F.1, TASK-3.D.1.1, TASK-3.D.3.3, TASK-TM ANPP-20853-WFQ, NUDOCS 8205260258
Download: ML17297B534 (60)

Text

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REOULATCQ INFORMATION DI'STRISUTIONSTEM

('RIDS)

AOCBSSION NBR:8205260258 DOC.DATE: '82/05/24

'NOTARIZED: YFS FAOIL:STN"50"528 Palo Verde Nuclear 'Stat)one Unitt ii Arizona 'Publi

'STN 50~529 'Palo Verde Nuclear Station~

Unitt '2r Ar)zona 'Pugli

'STN 50~530 'Palo Verde Nuclear Station~

Uni~t >3r Arizona Publi

'AUTH BYNAME AUTHOR AFF ILIATION.

VANBRUNTiE.E.

Ar.,izona Public 'Service Co ~

iRHCIP ~ NAME RECIPIENT AFFIL'IATION Office of Nuclear Reactor Regulations Director

SUBJECT:

For wards Amend 3 (to 820406 TMI 2 lessonslearned implementation reptiaddressing NUREG 0737 *items.

DI'GTRIBUTION 'CODE:

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P.o. BOX 21666 PHOENIX, ARIZONA86036 Mav 24 1982 ANPP-25853 WFg/KWG Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C.

20555

Subject:

Palo Verde Nuclear Generating Station Units 1, 2 and 3

Docket Nos. STN-50-528/529/530 File:

82-056-026

Dear Sir:

Arizona Public Service Company (APS),

as Project Manager and Operating Agent for Palo Verde Nuclear Generating Station (PVNGS) Units 1,

2 6 3, is submitting herewith forty (40) copies of Amendment 3

to the PVNGS Lessons Learned Implementation Report tendered April 6, 1981.

This amendment provides an update to the Lessons Learned Implementation Report.

Sincerely, ECUa ~~@

APS Vice President Nuclear Projects ANPP Project Director EEVB/KWG/wp Attachment cc:

P. L. Hourihan R. Greenfield E. Licitra A. C. Gehr

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STATE OF ARIZONA

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) st COUNTY OF MARICOPA)

I> A 'arte'r Rogers, represent that I am Nuclear Engineering Manager of Arizona Public Service Company, that the foregoing doc-ument has been signed by me for Edwin E. Van Brunt, Jr., Vice Presi-dent Nuclear Projects, on behalf of Arizona Public Service Company with full authority so to do, that I have read such document and know its contenets, and that to the best of my knowledge and belief, the statements made therein are true.

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. Swoin to before me this i

day of A. Carter Roger 1982 Notary Public My Commission expires:

INy Commission Expires Dec. 22, 1985

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PVNGS LLIR CHANGE PAGE LIST FOR AMENDMENT 3 Remove vii/viii I.A.1.1-1/-2 I.A.1.3-1/-2 I.A.1.3-3/-4 I.B.1.2-1/-2 I.C.1-7/blank I.C.5-1/-2 II.B.4-3/-4 II.F.l-l/-2 through II.F.1-4A/-4B III.A.1.1-5/-6

'II.D.l.l-l/-2through III.D.1.1-7/blank III.D.3.3-1/-2 Insert vii/viii I.A.1.1-1/>>2 I.A.1.3-1/-2 I.A.1.3-3/-4 Deleted I.B.1.2-1/-2 I.B.1.2-3/-4 I.B.1.2-5/blank I.C.1-7/blank I.C.5-1/-2 I.C.5-3/blank II.B.4-3/-4 II.F.1-1/-2 through II.F.1-4C/-4D III.A.1.1-5/-6 III.D.l.l-l/-2through III.D.1.1-13/blank III.D.3.3-1/-2 IV-1/-2 IV-3/-4 May 1982 Amendment 3

PVNGS LLIR II.K.3.3 Reporting Safety Valve and Relief Valve Failures and Challenges II.K.3.5 Automatic Trip of Reactor Coolant Pumps During Loss-of-Coolant Accident II.K.3.17 Report on Outages of Emergency Core-Cooling Systems Licensee Report.

and Proposed Technical Specification Changes II.K.3.25 Effect of Loss of. Alternating-Current Power on Pump Seals II.K.3.30 Revised Small-Break I.'oss-, of-.

Coolant-Accident Methods to Show Compliance with 10 CFR Part 50, Appendix K II.K.3.31 Plant-Specific Calculations to Show Compliance with 10 CFR Part 50.46 III.

EMERGENCY PREPAREDNESS AND RADIATION EFFECTS II.K.3.3-1 II.K.3.5-1 II.K.3.17-1 II.K.3.25-1 II.K.3.30-1 II.K.3.31-1 III.A III.A.1.1 NRC AND LICENSEE PREPAREDNESS I

Upgrade Emergency Preparedness III.A.1.1-1 III.A.1.1-1 III.A.1.2 Upgrade Emergency Support Facilities III.A.1.2-1 III.A.2 Improving Licensee Emergency Preparedness--Long-Term III.A.2-1

PVNGS LLIR III.D RADIATION,PROTECTION III.D.1.1-1 III.D.l.l Integrity of Systems Outside Con-tainment Likely to Contain Radio-active Material for Pressurized-Water Reactors and Boiling-Water III.D.3.3 III.D.3.4 I

Reactors Improved Inplant Iodine Instrumenta-tion Under Accident Conditions Control-Room Habitability Reguirements III.D.1.1-1 III.D.3.3-1 III.D.3.4-l IV.

RESPONSE

TO NRC UESTIONS Question IV.1 (NRC I&E Question 20) I.B.1.2 IV-1 Question IV.2 -(NRC I&E Question 21) I.A.1.3 IV-1 Question IV.3 (NRC ICE Question 24) I.A.l.l,,

~

IV-2 Question IV.4 (NRC ISLE Question 25) I.B.1.2 Question IV.5 (NRC F.J. Miraglia letter dated January 8,

1982) III.D.1.1 IV-2 IV-3 Amendment 3

May 1982

PVNGS LLIR I.

OPERATIONAL SAFETY I.A OPERATING PERSONNEL I.A.1.1 SHIFT TECHNICAL ADVISOR Position Each licensee shall provide an on-shift technical advisor to the shift supervisor.

The shift technical advisor (STA) may serve more than one unit at a multiunit site if qualified to perform the advisor function for the various units.

The STA shall have a bachelor's degree or equivalent in a scientific or engineering discipline and have received specific training in the response and analysis of the plant for transients and accidents.

The STA shall also receive training in plant design 'and layout, including the capabilities of instrumenta-tion and controls in the 'control room.

The licensee shall assign normal duties to the STAs that pertain to the engineer-ing aspects of assuring safe operations of the plant, including the review and evaluation of operating experience.

PVNGS Evaluation A shift technical advisor (STA) will be provided onsite in addi-tion to the shift supervisor for PVNGS for each shift.

The STA May 1982 I.A.1.1-1 Amendment 3

PVNGS LLIR will serve all three PVNGS units.

The duties of the STA will include:

~

Diagnose accidents and off-normal events for their significance to reactor safety and advise the shift supervisor.

~

Incorporation into the onsite Independent Safety Engineering Group (see section I.B.1.2).

Organizationally, the STA will report through the supervising engineer of the Independent Safety Engineering Group to the Operations Engineering Supervisor who is independent of opera-tions.

STA training, qualifications, and selection criteria are dis-cussed in FSAR Section 13.2.1.3.2.

STA requalification training will be conducted as described in amended FSAR Section 13.2.2.2.3.

Facility. Technical Specification 6.2.4 and Table 6.2 will be

proposed, further describing the station and duties of the STA.

Amendment 3

I.A.1,.1-2 May 1982

PVNGS LLIR I.A.1.3 SHIFT MANNING Position (1)

Limit Overtime Administrative'procedures shall be established to limit maximum work hours of all personnel performing a safety-related function.

(2)

Minimum Shift Crew The minimum shift crew for a unit shall include three operators, plus an additional three operators when the unit. is operating.

Shift staffing may be adjusted at-multi-unit stations to allow credit for operators holding licenses on more than one unit.

N In each control room, including common control rooms for multiple units, there shall be at all times a licensed reactor operator for each reactor loaded with fuel and a

senior reactor operator licensed for each reactor that. is operating. 'here shall also be onsite at all times, an additional relief operator licensed for each reactor, a

licensed senior reactor operator who is designated as shift supervisor, and any other licensed senior reactor operators rec{uired so that, their total number is at least.

one more than the number of control rooms from which a reactor is being operated.

November 1981 I.A.1.3-1 Amendment 2

PVNGS LLIR PVNGS Evaluation 1.

Limit Overtime PVNGS administrative procedures

shall, by fuel load, pro-vide provisions limiting maximum hours worked by personnel performing a safety related function to the guidelines of NRC Generic Letter No. 82-02:

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 turnover time).

b.

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 /> 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, 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 turnover time).

c.

A break of at least eight hours should be allowed between work periods (including shift turnover time).

d.

The use of overtime should be considered on an individ-ual basis and not for the entire staff on a shift.

Recognizing that very unusual circumstances may arise requiring deviation from the above guidelines, such devia-tion shall be authorized by the Manager of Nuclear Operations or his designee, or higher, levels of management.

The para-mount consideration in such authorization shall be that significant reductions in the effectiveness of operating personnel woul'd be highly unlikely.

Amendment 3

I.A.1.3-2 May 1982

PVNGS LLIR In addition, procedures encourage licensed operators at the controls to be periodically relieved and assigned to other duties away from the control board during their tour of duty The personnel effected by this requirement will be senior reactor operators, reactor operators, radiation protection

'echnicians, auxiliary operators, I 8 C technicians and key maintenance personnel.

2.

Minimum Shift Crew The minimum shift crew for a unit is discussed in FSAR Section 13.1.2.3 and FSAR Table 13.1-2 and meets the above requirements.

PVNGS administrative procedures will by fuel load provide provisions governing required shift staffing.

May 1982 I.A.1.3-3/-4 Deleted Amendment 3

A 1

I k

t

PVNGS LLIR I.B OVERALL ORGANIZATION I.B.1.2 INDEPENDENT SAFETY ENGINEERING GROUP Position Each applicant for an operating license shall establish an onsite independent, safety engineering group (ISEG) to perform inde-pendent reviews of plant operations.

The principal function of the ISEG is to examine plant operating characteristics, NRC issuances, Licensing Information Service advisories, and other appropriate sources of plant design and operating experience information that may indicate areas for improving plant safety..

The ISEG is to perform independent review and audits of plant activities, including maintenance, modifications, operational

problems, and operational

analysis, and aid in the establishment of programmatic requirements for plant activities.

Where useful improvements can be achieved, it is expected that this group will develop and present detailed recommendations to corporate management for such things as revised procedures or equipment modifications.

Another function of the ISEG is to maintain surveillance of plant operations and maintenance activities to provide inde-

.pendent verification that these activities, are performed cor-rectly and that human errors are reduced as far as practicable.

ISEG will then be in a position to advise utility manag'ement on the overall quality and safety -of operations.

ISEG need not perform detailed audits of plant operations and shall not be I.B.1.2-1

PVNGS LLIR responsible for sign-off functions such that it becomes involved in the operating organization.

PVNGS Evaluation In response to the recommendations of NUREG-0737, PVNGS has made changes to the plant staff organization which provide additional assurance that PVNGS is operated in a safe manner.

These organizational changes involve the Shift Technical Advisors (STA), as well as the Independent Safety Engineering Group (ISEG) and feedback of operating experience function.

The following is a description of the the organizations and staffing changes.

Inde endent Safet.

En ineerin Grou Or anization As described. in NUREG-0737, recently proposed revisions to Regulatory Guide 1.33, and INPO's recommendations on this subject, the independent safety review, operating experience evaluation, and Shift Technical Advisor accident assessment functions are directly related and overlap in many areas.

Accordingly, we have combined the STA responsibilities for all three units and the ISEG functions into a single onsite organization.

This group has a

close relationship with the Safety Audit Committee, our staff organization, which performs company independent safety reviews.

In addition to consolidating the closely related functions, this arrangement has the following advantages:

a.

An awareness of industry operating experience will be an important aid in STA accident.

assessment.

Amendment 3

I.B.1.2-2 May 1982

PVNGS LLIR b.

The ISEG duty "surveillance of plant activities to provide independent verification that these activities are performed correctly and human errors are reduced as much as possible" gains the STA familiarity with the equipment and a feeling for what types of equipment failures and human errors are most likely to occur.

c ~

Preparation of responses to IE circulars and informa-tion notices and other reports on plant problems gives the STA's exposure to management philosophy, an appreciation for the rigorous approach required, and also gives management a chance to evaluate STA performance prior to an accident. environment.

d.

STA's on duty need more work to do than accident, assessment.

ISEG duties enhance the STA job and pro-vide a well organized set of work assignments.

e.

STA training and responsibility form a good basis for ISEG decisions.

The Independent Safety Engineering Group will not replace either the Safety Audit Committee or the Plant Onsite Review Committee.

Its members support and may receive technical direction from the Safety Audit Committee.

Administratively, they report to the Engineering and Technical Services Manager who is independent from the responsibilities for day-to-day operations.

While on duty as STA, they advise the shift supervisor and report functionally through the station organization.

May 1982 I.B.1.2-3 Amendment, 3

PVNGS LLIR APS has issued a policy document, recpxiring that:

(1) reports of the ISEG dealing with safety issues be sent to the Chairman of the Safety Audit Committee (SAC),

and (2) safety issues raised by the ISEG are to be reviewed and resolved by the Chairman of the SAC.

The policy provides an appeal path for resolution of potential differences between the Chairman of SAC and the Vice President of Electric Operations.

Work Schedule and Function Our plan is to have the STA/ISEG personnel stand watch on a 24-hour duty day basis.

Thus, they will be asleep at times while on duty, but will be available in the Control Room on short notice.

As we gain more experience with this arrangement, we will continually reassess the work schedule and make modifi-cations needed to provide the most effective arrangement.

As STA, the primary responsibility is to provide technical assistance to the shift. supervisor during an off-normal event.

When on duty, but not assisting the shift supervisor, and when off duty, these personnel will perform the functions of the ISEG as listed below:

'a ~

See item I.C.5 for a description of the operating experience information evaluation program.

b.

Independent Evaluation and Surveillance of Plant Activities A wide range of plant activities, including operations, maintenance, and modifications are monitored during Amendment 3

I.B.1.2-4 May 1982

PVNGS LLIR the actual performance of the work to evaluate the technical adequacy. of the methods

used, recommend equipment

changes, and aid in the establishment of programmatic requirements for plant activities.

This surveillance provides independent verification that these activities are performed correctly and reduces the potential for human error as far as possible.

c.

Other duties involving the safe operation of the plant. as directed by the Safety Audit Committee.

The combined STA/ISEG Organization will be staffed with persons who meet the qualifications of FSAR Section 13.2.1.3.2 as

STA, and the organization will have at least five members with at least three years average level of nuclear power plant experience.

Facility Technical Specification 6.2.3 will be proposed to pro-vide ISEG function, composition, responsibilities, and authority.

May 1982 I.B.1.2-5 Amendment 3

(

PVNGS LLIR The C-E Owners Group held a meeting with the NRC Division of Systems Integration and Human Factors Safety on January 30,

1981, to discuss the process being used for revision of the emergency

'I procedure guidelines.

The revised emergency procedure guide-lines were submitted to the staff on June 30, 1981 as CEN-152 in addition to the submittal of CEN-156 titled Emergency Pro-cedure Guideline Development.

The NRC provided comments on these documents in a July 24, 1981 meeting and in a letter to C-E Owners Group dated September 15, 1981.

The C-E Owners Group is in the process of reviewing these

.guidelines in light of the NRC comments.

PVNGS intends to submit a Procedures Generation

Package, in accordance with Section 6.0 of Draft NUREG-0899, to the NRC for staff review following NRC approval of C-E Owner's Group emer-gency procedure guidelines.

- A target submission date of August 1, 1982 has been established for this package.

Emergency operating procedures will be developed and implemented in accord-ance with this package and will be ready for NRC onsite review 60 days prior to fuel load.

May 1982 I.C.1-7 Amendment 3

PVNGS LLIR I.C.5 PROCEDURES FOR FEEDBACK OF OPERATING EXPERIENCE TO PLANT STAFF Position In 'accordance with Task Action Plan I.C.S, Procedures for Feed-back of Operating Experience to Plant Staff (NUREG-0660),

each applicant. for an operating license shall prepare procedures to assure that operating information pertinent, to plant safety originating both within and outside the utility organization is continually supplied to operators and other personnel and is incorporated into training and retraining programs.

These procedures shall:

(1)

Clearly identify organizational responsibilities for review of operating experience, the feedback of pertinent information to operators and other person-

nel, and the incorporation of such information into training and retraining programs; (2)

Identify the administrative and technical review steps necessary in translating recommendations by the operating experience assessment group into plant actions (e.g.,

changes to procedures; operating orders);

(3)

Identify the recipients of va'rious categories of information from operating experience (i.e., supervi-sory personnel, shift technical advisors, operators, maintenance personnel, health physics technicians)

I.C.5-1

PVNGS LLIR or otherwise provide means through which such informa-tion can be readily related to the job functions of the recipients; (4)

Provide means to assure that affected personnel become aware of and understand information of sufficient importance that should not wait,for emphasis through routine training and retraining programs; (5)

Assure that plant personnel do not routinely receive extraneous and unimportant information on operating experience in such volume that it would obscure priority information or otherwise detract from over-all job performance and proficiency; (6)

Provide suitable checks to assure that conflicting or contradictory information is not conveyed to operators and other personnel until resolution is reached;

and, (7)

Provide periodic internal audit to assure that the feedback program functions effectively at all levels.

PVNGS Evaluation An operating experience review program is being developed for PVNGS which will establish the responsibilities and methodologies for reviewing the operating experience of PVNGS and other nuclear plants.

PVNGS will participate in the Institute of Nuclear Power Operations (INPO) Significant Event Evaluation and Infor-mation Network (SEE-IN) as discussed in NRC Generic Letter Amendment 3

I.C.5-2 May 1982

PVNGS LLIR No. 82-04.

The program and implementing procedures will be developed in accordance with this reguirement and will.be in effect prior to Unit 1 fuel load.

May 1982 I.C.5-3 Amendment 3

PVNGS LLIR E.

Radiation Monitorin

Response

of Process and Area Monitors to severe damages; behavior of detectors when saturated; method for detecting radiation readings by direct measure-ment at detector output (overranged amplifier);

expected accuracy of detectors at different loca-tions; use of detectors to determine extent of core damage.

2.

Methods of determining dose rate inside containment from measurements taken outside containment.

F.

Gas Generation Methods of H2 generation during an accident; other sources of gas (Xe, Kr); techniques for venting or disposal of non-condensables.

2.

H flammability and explosive limit; sources of 0 in containment or Reactor Coolant System.

(2)

Complete Training The course shall be developed and these personnel shall participate prior to fuel loading of unit one.

PVNGS'valuation (1)

Develop Training Program A course will be developed to train shift technical advisors and operating personnel through the operations II.B.4-3

PVNGS LLIR chain to the licensed operators in the use of installed plant systems to control or mitigate an accident in which the core is severely damaged.

The training program will include the topics suggested in the H.R. Denton letter of March 28, 1980.

This training will consist of approxi-mately 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of formal classroom presentations by either a private consultant or the engineering staff.

Operators will acquire the theoretical basis for these actions in the academic programs and the practical application during the simulator training course.

The total training time will exceed 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />.

Managers and technicians in the Instrumentation and Control (I 6 C), radiation protection and chemistry sections will receive training commensurate with their responsibilities that meets the requirements of the H.R. Denton letter of March 28, 1980.,

(2 )

Complete Training The above training will be completed prior to fuel load operation.

Amendment 3

II.B.4-4 May 1982

PVNGS LLIR II.F INSTRUMENTATION AND CONTROLS II.F.l ADDITIONALACCIDENT-MONITORING INSTRUMENTATION A human factor analysis will be performed to ensure that 'the displays and controls added for additional-accident accomplish this monitoring,. do not increase the potential for operator error (see section I.D.l).

Installation will be completed prior to fuel load.

II.F.1.3.

ATTACHMENT l, NOBLE GAS EFFLUENT MONITOR Position Noble gas effluent monitors shall be installed with an extended range designed to function during accident conditions as well as during normal operating conditions.

Multiple monitors are considered necessary to cover the ranges of interest.

(1)

Noble gas effluent monitors with an upper range capacity of 10 pCi/cc (Xe-133) are considered to be practical and should be installed in all operating plants.

(2)

Noble gas effluent monitoring shall be provided for the total range of concentration extending from normal condi-tion (.as low as reasonably achievable (ALARA)) concentra-tions to a maximum of 10 pCi/cc (Xe-133).

Multiple monitors are considered to be necessary to cover the ranges of interest.

The range capacity of individual monitors should overlap by a factor of ten.

November 1981 II.F.1-1 Amendment 2

PVNGS LLIR PVNGS Evaluation FSAR section.11.5 provides detailed descriptions of the effluent monitors installed at Palo Verde Units 1, 2 and 3.

This.includes the additional monitors that have been added specifically to address NUREG-0737 and Reg Guide 1.97, Rev 2 requirements for radiation monitoring.

Installation and calibration will be completed by fuel load.

A description of the calibration

sources, frequency of calibration, and technique is pro-vided in FSAR Table 11.5.1 and Sections 11.5.2.1.6.2 and 11.5.2.1.6 respectively.

The instrumentation is described in detail in FSAR Table 11.5-1.

The outputs of the effluent monitor and flow meter will be in pCi/cc and cc/h respectively and when obtained need only be multiplied to obtain pCi/h.

Sampling of effluents will meet the criteria of ANSI N13.1 1969 as discussed in FSAR Sections 11.5.2.1.1.7.2.2 and 11.5.2.2.1.

Monitors are designed to meet a

90% efficient level for partic-ulates and 90% efficiency for iodine as required by NUREG-0737 Table II.F.1-2.

They are also designed to conform with design basis shielding envelopes for sampling media as discussed in FSAR Section 12.1.2.4 and item II.B.2.

Monitors are designed to allow personnel to remove;

replace, and transport sampling media without exceeding the criteria of GDC19 of 5 rem whole-body and 75 rem to the extremities.

Each process or effluent channel includes a sampling assembly which consists of a sampler and the-associated piping, fittings, and other components as required to.transport the sample Amendment 3

II.F.1-2 May 1982

PVNGS LLIR through the system.

The sampling assembly is a closed, sealed system and includes a sampling pump, valves, interconnecting piping, filters, fittings, flow and pressure transducers, and other local" control and instrumentation elements as required.

Samplers, with the exception of the condenser vacuum pump/gland seal exhaust particulate-iodine

sampler, house radiation detec-tion equipment and check source(s).

Sampler piping and connections are welded except where maintenance considerations make flanged or Swagelok joints necessary.

Sampler outlet piping connections are located to minimize cleaning requirements and background buildup due to the adherence of radioactive particles to the sampler walls.

For liquid samplers, welding of pressure-containing components is performed in accordance with ANSI B31.1.

For ESF monitors, welding of pressure-

/

containing components is performed in accordance with AWS Dl.1-1972 (with 1973 revisions).

Welding of other equipment is performed in accordance with industry standards.

For liquid and process

channels, the sampler is a lead-shielded steel chamber.

For particulate and iodine channels, the sampler is a lead-shielded filter assembly.

Four n shielding is furnished for all process and effluent detectors.

Airborne particulate and iodine monitors and samplers, with the exception of the containment, building atmosphere

monitor, sample isokinetically in accordance with the principles and methods of ANSI N13.1-1969, Guide to Sampling Airborne May 1982 II.F.1-3 Amendment 3

PVNGS ILIR Radioactive Materials in Nuclear Facilities.

The particulate and iodine sample flow is maintained constant over the normal expected range of filter paper and/or charcoal cartridge differential pressure by an automatic control system.

Local flow indication and high-and low-flow alarm signals are pro-vided.

These signals actuate local alarms and the channel failure alarms.

Particulate samplers (except the main con-denser air ejector low range) are moving paper filter type and incorporate microcomputer-controlled step advance and feed failure channel failure alarm.

Sampling assembly fit-tings are provided which allow grab sampling of the monitored airstreams.

A flow-integrating elapsed sample volume indicator is provided downstream of each particulate and/or iodine channel.

It has a local digital readout and is resettable to zero.

A human factor analysis was performed as discussed in item I.D.l.

A.

Wide-Range Effluent Monitor In order to cover the dynamic range required, a normal range monitor is used with a high range monitor.

One decade is used for overlap when switching between monitors.

Both monitors sample using controlled isokinetic flows.

Their combined range is 10 pCi/cc to 10 yCi/cc.

These monitors are located on the plant vent, main condenser/gland seal exhaust, and the fuel building vent.

In addition to noble gas monitoring capability, these instruments have separate particulate and iodine sample chambers for both the low and high range, and use charcoal or Silver-Zeolite cartridges in conjunction with a portable Amendment 3

II.F.1-4 May 1982

PVNGS LLIR pump.

All cartridges will be removed to the counting laboratory for gamma spectrum analysis.

High range monitor iodine samples are provided with a lead shield.

Procedures will be developed to define ALARA concepts for removal, transport and analysis.

These monitors have complete digital readout and control from the Health Physics Office and the main control room.

The high range monitors automatically switch to a new particulate/iodine cartridge pair when the current cartridge reaches a preset radiation level.

Filter materials used minimize absorption of noble gases.

Samples are preconditioned as necessary to assure accurate results without damaging the sample assemblies.

Each monitor is controlled by a remote microprocessor.

This microprocessor is linked by a "daisy chain" to a minicomputer which provides multiple informational displays on request by the operator.

A dedicated alarm status line is maintained on the CRT display.

This status line does not move with each change of CRT displays.

Thus alarms are provided regardless of the status of the displays in the Health Physics Office and Main Control Room.

Monitors are provided with an open structural construction that provides for easy maintenance and good heat dissipation.

Backup battery power is provided to assure continued micro-processor memory during a loss of external power sources.

Multiple detectors are used to achieve the dynamic range required.

Hard copy readouts are available from dedicated printers in the Health Physics office and the control room.

May 1982 II.F.1-4A Amendment 3

PVNGS LLIR B.

Main Steam Line Monitor One area monitor with a collimating lead shield is mounted adjacent to each main steam line in the Main Steam Support Structure approximately one foot upstream of the atmospheric dump valves.

Refer To FSAR Figure 12.3-2.

These monitors measure direct dose rates from the main steam line to'dentify effluent from the atmospheric

dump, main steam relief valves, and auxiliary feedwater pump discharge.

An extra 2 inches of shielding is placed on the containment side of the detector shield.

There are a total of 4 detectors with one remote microprocessor for each 2 detectors.

The ion chamber covers a range from lmr/h to 10 mr/h.

The microprocessor is linked by a "daisy chain" to a minicomputer which provides multiple informational displays at the request of the operator.

Alarms are provided regardless of the status of the displays in the Health Physics Office and the Control Room.

Hardcopy readouts are provided from dedicated printers in the Health Physics Office and the Control Room.

Backup battery power is provided to assure continued microprocessor memory during a loss of external power sources.

The detector is designed to operate in a post.-accident environmental condition with a background of 10 R/h.

C.

Meeting the NRC Requirement These monitors operate in conjunction with the other monitors as discussed in FSAR section 11.5 and fulfillthe requirements as Amendment 3

II.F.1-4B May 1982

PVNGS LLIR outlined in NUREG-0737 and Regulatory Guide 1.97, Rev 2.

Installation will be completed prior to fuel load.

II.F.1.2 ATTACHMENT 2, SAMPLING AND ANALYSIS OF PLANT EFFLUENTS Position Because iodine gaseous effluent monitors for the accident con-dition are not considered to be practical at this time, capability for effluent monitoring of radioiodines for the accident condition shall be provided with sampling conducted by adsorption on charcoal or other media, followed by onsite laboratory analysis.

PVNGS Evaluation PVNGS response to this item is included in the evaluation of section II.F'.l.l requirements.

May 1982 II.F.1-4C Amendment 3

PVNGS LLIR 3l This page intentionally blank 3)

Amendment 3

IX.F. 1-4D May 1982

PVNGS LLIR Data acquisition of the two independent meteorological system signals may be accomplished for projected dose calculations, CRT displays and remote data transmission.

The same capability is provided for the nuclear data link.

As independent, redundant and validated si'gnals are used, the system can achieve a non-availability of 0.01.

Below is a summary table of the sensors used in each inde-pendent meteorological system to monitor the environmental

'I parameters, and the common data processing available.

Sensor Location 200'ind Speed and Wind Direction Sensor 35'ind Speed and Wind Direction Sensor.

200'c 35'5'spirated Aspirated Temperature Dewpoint Sensor Sensor Rainfall Monitor Located at Ground Level Tower System A

Provided Provided Provided Provided Provided Tower System B

Provided Provided Provided Not

.Provided Not Provided Digital 6c Analog Process-ing.

Provided 35'mbient

'ddt 200'-35'T Provided Provided The environmental parameters monitored by each independent tower system permits highly accurate and reliable meteor<<

ological data necessary to cover all data for the Pasquill stability classes

.and transport protections needed for the 1

PVNGS site.

November 1981 III.A.1.1-5 Amendment 2

PVNGS LLIR The auxiliary analog information for each tower system is provided on analog recorders located at the meteorological station in the respective tower system equipment trailers.

Analog information for each tower system is converted to digital data and transmitted by two serial links to the plant site where the data is reduced to fifteen-minute and hourly average meteorological parameters, and where all effluent and environmental parameters are recorded and available for time history displays in the control room, emergency response facilities, and at external locations.

Serial data is also provided to the radiation exposure management system to meet the reporting requirements of Regulatory Guide 1.21.

In addition, the signal is avail-able to the multi-channel processing system for offsite and projected dose calculation, technical support center and the emergency operations facility CRT displays.

The models used for providing the estimates of offsite exposure are:

Plume exposure - Class A model will be used with 15 minute average meteorology data for initial transport and diffusion estimates within 15 minutes following the classification of the incident.

Ingestion Zone A. modified Class A model will be used for relative concentration of the plume emer-gency planning zone (EPZ) and ingestion EPZ.

Amendment 3

III.A.1.1-6 May 1982

PVNGS LLIR III.D RADIATION PROTECTION

.III.D.1.1 INTEGRITY OF SYSTEMS OUTSIDE CONTAINMENT LIKELY TO CONTAIN RADIOACTIVE MATERIAL FOR PRESSURIZED-WATER REACTORS AND BOILING-WATER REACTORS Position Applicants shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as-low-as'practical levels.

This program shall include the following:

(1)

Immediate leak reduction (a)

Implement all practical leak reduction measures for all systems that could carry radioactive fluid outside of containment.

(b)

Measure actual leakage rates with system in operation and report them to the NRC.

(2)

Continuing Leak Reduction -- Establish and implement a program of preventive maintenance to reduce leak-age to as-low-as-practical levels.

This program shall include periodic integrated leak tests at intervals not to exceed each refueling cycle.

PVNGS LLIR PVNGS Evaluation 1.

Design Review Refer to CESSAR Appendix B, Item III.D.1.1.

In addition, a

PVNGS design review was performed on the system below to assure that potential radioactive release paths following a serious transient or accident is reduced to as-low-as-reasonably achievable (ALARA) levels.

A.'hutdown Coolin S stem SCS The existing design incorporates all-welded piping.

Vent and drain lines throughout the system are capped when not in use.

Relief valves on the system relieve to the equipment drain tank (a tank designed to. accept radioactive fluids).

The leakage from the LPSI pump seals and system valve stems is ALARA.

Potential leakage from the SCS into the essential cooling water system (through the shutdown heat exchanger) can be detected during normal operation by installed radiation monitoring.

B.

Containment S ra Recirculation S stem CS The existing design incorporates all-welded piping.

Vent and drain lines throughout the system are capped when not in use.

Relief valves on the system (external to the containment) relieve to the equip-ment drain tank.

The leakage from the CS pump seals and system valve stems is ALARA. Potential leakage Amendment 3

III.D.1.1-2 May 1982

PVNGS LLIR during normal operation from the CS into the essential cooling water system (through the shutdown heat exchanger) can be detected by installed radiation monitoring.

C.

CVCS Char in and Letdown S stem The existing design incorporates all-welded piping.

The letdown system is isolated upon CIAS and SIAS.

Relief valves on the system relieve to the equipment drain tank.

The leakage from the CVCS charging pumps (positive displacement pumps) and other system equipment is ALARA as they are hard-piped to drains.

The nuclear cooling water system is monitored for potential leakage from the CVCS through the letdown heat exchanger during normal operation.

D.

Sam lin S stem The existing design of the normal sampling system incorporates "Swagelok" connections,

however, the design will be upgraded to all-welded piping for sections which would come into contact with highly radioactive fluids.

The system is isolated upon CIAS and SIAS.

Relief valves relieve to the equipment drain tank.

Leakage from the system is also mini-mized by the small size of the lines.

May 1982 III.D.1.1-3 Amendment 3

PVNGS LLIR The post-accident.

sampling system will also be constructed of all-welded piping, except within cabinets.

E.

Hi h-Pressure In'ection Recirculation HPSI The design incorporates all-welded piping.

Relief valves on the system (external to the containment) relieve to the equipment drain tank.

The vent and drain lines throughout the system are capped when not in use.

Leakage from the HPSI pump seals and system valve stems is as-low-as reasonably achievable.

P I

Miniflow connections to the refueling water tank (RWT) are isolated upon Recirculation Actuation Signal (RAS).

Manual cross over valves to the CVCS are normally locked shut.

F.

Waste Gas S stem The waste gas system is isolated from the contain-ment upon CIAS.

(The normal vent path from the reactor drain tank (RDT) and the reactor head vent system is isolated.)

By design, the introduction of highly radioactive fluids to the system is precluded.

t As part of the system testing program, each of the above systems is hydrostatically tested to 150% normal operating pressure per the reguirements of ANSI B31.1, Summer 1976 Addendum for ANSI B31.1 piping systems, and to 125% normal Amendment 3

III.D.1.1-4 May 1982

PVNGS LLIR 2.

operating pressure per the requirements of,ASNE Boiler

& Pressure Vessel

Code,Section III, 1977 Edition, for AS'iping systems.

Leakage Reduction Program PVNGS will institute a program to maintain leakage rates of systems outside containment, to as low as practical which consists of the following:

A.

Systems Included in the Pxogram 1.

High pressure safety injection system (recircu-lation portion only).

2.

Low pressure safety injection system (shutdown cooling portion only).

3.

Reactor coolant sampling system (post-accident sampling piping only).

4.

Containment spray system.

5.

6.

Radioactive waste gas system (post-accident sampling return piping only).

Liquid radwaste system (post-accident sampling return piping only).

7.

Containment combustible gas and atmospheric sampling system (hydrogen, monitoring sub-system and post-accident sample piping associated with this function).

May 1982 III.D.1.1-5 Amendment 3

PVNGS LLIR B.

Systems excluded from the program:

(They will not preclude any option of cooling the reactor core nor will they prevent the use of, needed safety systems).

Radioactive liquid waste

system, except as discussed above.

2.

3.

Radioactive waste gas

system, except as discussed above.

(The system is not required for use post-accident.)

Reactor coolant letdown system, except for portions required for, post-accident sampling described in FSAR section 9.3.2.2.2.

(The system is not required to function post-accident.

The plant can be brought to a cold shutdown condition without the letdown system.

'\\

The letdown system is isolated on SIAS and CIAS.)

Reactor coolant pump"seal bleed-off system.

(The system is not required to function outside containment post-accident.

The seal bleed-off system is isolated outside containment on CIAS.

The system remains isolated post-accident.

If seal bleed-off is required post-.accident, pressure in the seal bleed-off header will increase and the header relief valve willliftproviding a flow path to the reactor drain tank.)

Amendment 3

III.D.1.1-6 May 1982

PVNGS ILIR 5.

Charging System.

(The charging system under post-accident conditions does not contain radioactive fluid since the letdown system is isolated as dis-cussed in item 3 above.

The charging system takes suction from the refueling water tank.)

6.

Fuel Pool Cooling system (FPC).

(The FPC is normally isolated from potentially highly con-taminated systems by double, locked shut isolation valves.)

C.

Program Features Immediate leak reduction measures:

The program will consist of periodic monitoring of the systems during operation and inservice leak testing.

Leaks will be identified and corrective maintenance performed.

1.

Vent and drain lines will be capped to prevent 2.

release due to seal leakage.

The packing of valves (except Kerotest which is a packless, stainless steel diaphragm valve) in the scoped liquid systems will be inspected for leakage or evidence of leakage such as boric acid accumulation.

Maintenance will be performed on the packing of liquid system valves identified as requiring work.

3.

The seals and packing on pumps in the scoped liquid systems will be, inspected for leakage or signs of leakage.

May 1982 III-.D.1.1-7 Amendment 3

PVNGS LLIR 5.

Valves, fittings, and compressor seals in the scoped gaseous systems will be checked for leakage.

Maintenance will be performed on gas system valves and instrument fittings identified during leak tests as requiring work.'ystems and subsystems identified in 2A will be leak tested prior to exceeding 5% power and on an interval not to exceed the period between refuel-ing outages.

Test records including measured leak rates will be maintained at PVNGS for NRC review.

A report including the measured leak rates will be submitted for NRC staff review prior to operation above 5% power. 'eak rate test techniques will include:

a.

Liquid Systems A visual examination will be performed on items 1

through 3 of paragraph 2C above with the system at or near operating pressure.

If leakage is identi-fied during these examinati'ons, an integrated leakage rate will be determined by monitoring the applicable sump and tank levels.

For sumps and tanks that do not contain a level indicator, the levels will be determined by physical measurement.

In addition, the local leak rate tests performed on isolation valves will be utilized for the portion of each system located between the containment and Amendment 3

III.D.1.1-8 May 1982

PVNGS LLIR the isolation valves, if-practical.

These tests will be performed in accordance with written Station Manual procedures.

b.

Gas Systems The leakage will be determined by detecting gas leakage at individual valves, fittings, seals, and bolted connections with the system at or near operation pressure.

Leakage will be detected by use of acoustic, bubble, or equivalent method (such as a tracer gas method).

In addition, the local leak rate tests performed on isolation valves will be utilized for that portion of each system located between the containment and the isolation valves, if practical.

~ These tests will be per-formed in accordance with written Station Manual procedures.

The PVNGS design was reviewed to confirm that the design and construction of PVNGS systems minimize unplanned releases of radioactivity including the related incidents identified in NRC letter dated October 17, 1979 to all operating nuclear power plants.

The following summarizes that review:

Radioactive liquid atmospheric tanks are provided with overflows with either no isolation valve or a locked-open valve.

Overflow lines have loop seals and are routed to appropriate radioactive building sumps.,

May 1982 III.D.1.1-9 Amendment 3

PVNGS LLIR The sump liquid is routed to the LRS holdup tanks.

Overflow lines from the refueling water tank and the LRS, concentrate monitor tanks are heat-traced to prevent plugging.

Radioactive liquid pressurized tanks with the exception of the volume control tank and reactor drain tank are provided with relief lines routed to the appropriate sumps.

A summary of the overflow provisions for the radioactive tanks is provided in table III.D.1.1-1.

Storm drains are located away from areas witha high potential for radioactive spills.

No storm drains exist in the immediate vicinity of the Containment, Auxiliary, or Radwaste Buildings.

Radioactive pumps are generally located 'in isolated compartments whose drains are designed to catch all potential leakage.

These drains are routed to the appropriate radioactive building sump.

In addition, certain pumps whose potential for radioactive leakage is greatest are equipped with drip pans with lines hard-piped to'he associated building sump.

A summary of the radioactive pumps and their leakage provisions is given in table III.D.1.1-2.

Radioactive valves're located in shielded compartments such as valve galleries equipped with floor drains that are designed to collect all potential valve leakage.

These drains are routed to appropriate building sumps.

Radioactive tanks located inside the'Auxiliary and Radwaste Buildings are located in compartments with curbs to contain tank leakage.

These compartments 'are also equipped with floor drains routed to the appropriate radioactive building sump'.

Outside Amendment 3

III.D.1.1-10 Nay 1982

PVNGS LLIR the contents of a tank rupture.

Outside CVCS tanks are concrete tanks with steel liners.

The concrete tanks will retain potential liner leakage.

The hot lab, cold lab, decontamination

area, and sample station are equipped with floor drains routed to the non-ESF sump.

There are no piping systems between units which could become contaiminated.

Based on this discussion the North Anna-type event is not expected

.to occur at PVNGS.

May 1982 III.D 1.1-11 Amendment 3

Table III.D.l.l-l RADIOACTIVE TANKS OVERFLOW AND LEAKAGE PROTECTION P&ID CHP-001 Tank Volume Control Tank Atmospheri c or Pressure Vessel PV Overflow or Relief Relieves to vent gas surge header Overflow or Relief Line N-214-HCDA-3/4'ank Location Auxiliary Bldg 120'evel Curb or Enclosed Compartment Enclosed Compartment Coaaaents CHP-002 Refueling Water Tank Overflow to holdup tank sump N-134-HCDA-6" Outside of fuel bldg Concrete w/

Overflow line is steel liner heat-traced.

CHP-003 Reactor Makeup Water Tank Overflows to holdup N-381-HCDA-3" tank sump Outside of fuel bldg Concrete w/

steel line CHP-001 Radwaste Crud Tank PV Relieves to non-ESF sump N-533-GCDA-2" Auxiliary Bldg 100'evel 4" curb CHP-003 CHP-003 Reactor Drain Tank Eguipment Drain Tank PV PV Vents to gas surge tank Relieves to non-ESF sump N-281-HCDB-2" N-347-HCDB-1" Containment 80'evel Auxiliary Bldg 40'evel CHP-003 Holdup Tank Overflows to holdup N-353-HCDA-3" tank sump Outside of fuel bldg Concrete w/

steel liner LRP-001 Low TDS Holdup Tank LRP-001 High TDS Holdup Tanks Overflows to radwaste bldg sump Overflows to radwaste bldg sump N-014-HCDA-6" N-229-HCDA-4" Outside of Radwaste Bldg Outside of Radwaste Bldg Enclosed compartment'nclosed compartment LRP-001 LRP-002 LRP-002 Chemical Drain Tanks Concentrate Monitor Tanks Recycle Monitor Tanks Overflows to aux bldg sump via a funnel drain Overflows to radwaste bldg sump Overflows to radwaste bldg sump N-067-HCDA-3" N-206-HCDA-2 N-195-HCDC-2" N-219-HCDC-1" N 183 HCDA 3u N-205-HCDA-3" Auxiliary Bldg 51'-6" level Radwaste Bldg 100'evel Outside of Radwaste Bldg 6" curb 6" curb Enclosed compartment Overflow lines are heat-traced SRP-001 Low Activity Spent Resin Tank SRP-002 Waste Feed Tank SRP-001 High Activity Spent Resin Tank PV PV Relieves to radwaste bldg sump Relieves to radwaste bldg sump Overflows to radwaste bldg sump via funnel drain N-027-HCDA-2" N-016-HCDA-2" N-204-HCDC-3/4" Radwaste Bldg 100'evel Radwaste Bldg 100'evel Radwaste Bldg 100'evel Curb Curb Enclosed Compartment

Table III.D.1.1-2 RADIOACTIVE PUMPS LEAKAGE PROVISIONS PSlID Pump Drain Pan Drain Line Location Comments CHP-001 Crud Pump CHP-002 Charging Pumps N-554-HCDA-1".

Drains to non-ESF sump N-245-HCDB-1" N-246-HCDB-1" N-247-HCDB-1".

Drain to recycle drain header.

Auxiliary Bldg 100'evel None Auxiliary Bldg 100'evel None CHP-002 Boric Acid Makeup Pumps N-449-XCDA-1/2" N-453-XCDA-1/2" Drain to non-ESF sump.

Auxiliary Bldg 70'-0'evel Equipped with a gland seal loop off the process flow CHP-003 Reactor Makeup Water Pumps No drip pan.

Drain line off gland seal to holdup tank sump Auxiliary Bldg 70'evel Equipped with a gland seal loop off the process flow CHP-003 Reactor Drain Pumps N-476-XCDA-1" N-479-XCDA-1".

Drain to a funnel drain routed to non-ESF sump Auxiliary Bldg 40'evel None CHP-003 Holdup Pumps N-482-XCDA-1/2" N-488-XCDA-1/2".

Drain to holdup tank sump Auxiliary Bldg 40'evel Equipped with a gland seal loop off the process flow LRP-001 LRS Holdup Pumps.

N-031-HCDA-1" N-032-HCDA-1 N-033-HCDA-1".

Drain to radwaste bldg sump Radwaste Bldg 100'evel None LRP-001 Chemical Drain N-079-HCDA-1" N-082-HCDA-1" Drain to a funnel drain routed to radwaste bldg sump Auxiliary Bldg 40'-0" level None LRP-002 LRP-002 Concentrate Monitor Tank Pumps Recycle Monitor Pumps N-117-HCDC-lu N-620-HCDC 1" Drain to radwaste bldg sump N-186-HCDA-1".

Drains to radwaste bldg sump Radwaste Bldg. 100'evel Drain line is heat traced Radwaste Bldg 100'evel None PCP 001 PCP-001 Fuel Pool Cleanup Pumps Fuel Pool Cooling Pumps No drip pan No drip pan Fuel Bldg 100'evel Fuel Bldg 100'evel Equipped with a gland seal loop off the process flow which drains to fuel building sump Equipped with a gland seal loop off the process flow which drains to fuel building sump SRP-001 Resin Transfer/

Dewatering Pump N-081-HCDA-1" ~

Drains to a local stub-up routed to radwaste bldg sump Radwaste Bldg 100'evel None SRP-002 Waste Feed Pump N-068-HCDC-1".

Drains to a local stub-up routed to radwaste bldg sump Radwaste Bldg 100'evel None

PVNGS LLIR ACCIDENT CONDITIONS Position (1)

Each licensee shall provide equipment and associated training,and procedures for accurately determining the airborne iodine concentration in areas within the facility where plant personnel may be present during an accident.

(2)

Each applicant for a fuel-loading license to be issued prior to January 1,

1981 shall provide the equipment,

training, and procedures necessary to accurately deter-mine the presence of airborne radioiodine in areas within the plant where plant personnel may be present during an accident.

PVNGS Evaluation Prior to fuel load, procedures will be developed for determin-ing airborne iodine concentration.

Silver-Zeolite or charcoal cartridges will be used in conjunction with a portable pump.

The cartridges will be removed to the counting laboratory for gamma spectrum analysis.

Cartridges will be purged with clean bottled compressed air prior to counting.

A low background counting facility is available on the site.

The results of airborne concentration can be obtained within 15 to 30 minutes after collection of iodine on filtered cart-May 1982 III.D.3.3-1

'Amendment 3

PVNGS LLIR removal, transport, and analysis of filter.cartridges.

There will be at least three such portable airborne monitors avail-able at each unit which meet the NUREG-0737 recommendations.

PVNGS response to this item is included in the evaluation of section II.F.1 requirements.

Amendment.

2 III.D.3.3-2 November 1981

PVNGS LLIR IV.

RESPONSES TO NRC RE UESTS FOR INFORMATION Please indicate your commitment that the Independent Safety Engineering Group (ISEG) which you plan to establish per-your letter of April 6, 1981 will meet all of the guidance provided in Section I.B.1.2 of NUREG-0737 (November 1980).

Also,'lease commit that the members of this group will satisfy the qualifi-cation requirements for Staff Specialists as defined in Para-graph 4.7.2, ANS 3.1-1978; and.that the group will provide to management, no less frequently than monthly, a summary of their activities to advise management on the overall quality and safety of operation.

RESPONSE

Refer to amended section I.A.1.3.

A general description of our organization which combines the STA program with the ISEG program is given in amended sec-tion I.A.l.l. and I.B.1.2.

Please indi'cate your commitment that the PVNGS procedures limiting overtime (as committed to in your letter of April 6,'981) will conform to the guidance contained in Section I.A.1.3

'I of NUREG-0737 (November 1980).

RESPONSE

The response is provided in amended section I.A.1.3.

May 1982

'V-1 Amendment 3

PVNGS LLIR Q

~

(

Q"""""

I

~

(I.A.1.1)

'I Please describe where the Shift Technical Advisor Group and the Independent Safety Engineering Group are located in the APS Nuclear Operations organization.

RESPONSE

~ Refer to amended sections I.A.l.l and I.B.1.2.

(I.B.1. 2)

Please furnish copies of the following procedures you committed to prepare in response to the TMI Task Action Plan Items listed below:

I.A.1.2 Shift Supervisor Administrative Duties I.A.1.3.1 Overtime I.A.1.3.2 Shift Manning and Movement of Key on-Shift

'Personnel I.B.12 I.C.2 I.C.3 I.C.4 I.CQ5 Independent Safety Engineering Group (Group Charter)

Shift Relief and Turnover Procedures Shift Supervisor Responsibilities Control Room Access Procedures for Feedback of Operating Experience to Plant Staff

RESPONSE

These procedures will be available onsite for review six months prior to fuel load.

Amendment 3

IV-2 May 1982

PVNGS LLIR 1

d d

(III.D.1.1)

January 8,

1982)

Your response to our earlier request, for additional information relating to this action plan item has not addressed the following items.

These should be addressed:

a.

Leak test methods for liquid and gaseous systems.

b.

C.

Applicability to Palo Verde of North Anna and related incidents (identified in NRC's letter dated 10/17/79 to all operating nuclear power plants).

Measured actual leak rates from all applicable systems with the system in operation (at this time, at. least a

commitment, must be made that these will be submitted according to the schedule given in NUREG-0737).

d.

e.

Frequency of the periodic integrated leak tests.

Major features of the continuing Leak Reduction Program.

f.

Leak testing for the containment sampling system.

g.

Leak testing for Residual Heat Removal System.

RESPONSE

a.

.The response is given in amended section III.D.l.l.

b.

The response is given in amended section III.D.l.l.

c.

The response is given in amended section III.D.l.l.

d.

The response is given in amended section III.D.l.l.

e.

The response is given in amended section III.D.1.1.

May "1982 IV-3 Amendment 3

PVNGS LLIR f.

The response is,given in amended section III.D.l.l.

g The function of a "residual heat removal system" is performed. by the shutdown cooling system which consists of portions of the high pressure and low pressure safety injection and containment spray systems, which are included in the leakage reduction program described in section III.D.l.l(2).

Amendment 3

IV-4 May l982