Text

Forwards Addl Info Re Engineering Employee Concerns in Response to Ja Zwolinski 870715 Request to SA White.Telcon Can Be Scheduled to Discuss Any Needed Clarification or Further Questions
	ML20237K143
Person / Time
Site:	Sequoyah
Issue date:	08/13/1987
From:	Russell J; TENNESSEE VALLEY AUTHORITY
To:	; NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
Shared Package
ML20237K146	List: ML20237K143; ML20237K152; ML20237K157;
References
	NUDOCS 8708190104
	Download: ML20237K143 (32)
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x TENNESSEE VALLEY AUTHORITY Watts Bar Nuclear Plant P. O. Box 800-l Spring City, Tennessee 37381 August 13, 1987 U.S. Nuclear Regulatory Commission Attn: Document Control Deak Washington, DC 20555 Gentlemen:

In the Matter of the Application of

) Docket Nos. 50-327 Tennessee Valley Authority

)

50-328 EMPLOYEE CONCERNS TASK GROUP (ECTG)

Reference:

Memorandum from John A. Zwolinski, NRC, to S. A. White dated July 15, 1987, " Request for Additional Information on Engineering Employee Concerns" In response to your request for additional information concerning engineering element reports related to Jequoyah restart, TVA has researched the pertinent information and the appropriate responses are included as Enclosure 1 to this letter.

Subsequent to your review of this letter, a teleconference can be scheduled to discuss any needed clarifications or further questions as desired.

Please contact Ron Gagne (615) 365-3788, Watts Bar, if you need further assistance.

Very truly yours, TENNESSEE VALLEY AUTHORITY R. M James R. Russell Watts Bar Nuclear Plant ECTG Licensing Manager cc (See page 2.):

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An Equal Opportunity Employer v

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It U.S. Nuclear' Regulatory Commission l.

cc (Enclosures)

Mr. G. G. Zech, Assistant Director Regional Inspections Division of TVA Projects Office of Special Projects.

U.S.: Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 Mr. John A. Zwolinski, Assistant Director For Projects Division of TVA. Projects Office of Special Projects U.S. Nuclear Regulatory Commission 4350 East West Highway EWW 322 Bethesda, Maryland 20018 Sequoyah Resident Inspector Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy Daisy, Tennessee 37379

Y

l. e TENNESSEE VALLEY AUTHORITY Watts Bar Nuclear Plant

-P. O. Box 800 Spring City Tennessee 37381 August 13, 1987 0;S. Nuclear Regulatory Commission Attn: Document Control Dest j

. Washington, DC 20555 Gentlemen:

In the Matter of the Application of

) Docket Nos. 50-327

{

Tennessee Valley Authority

)

50-328 4

I EMPLOYEE CONCERNS TASK GROUP (ECTG)

I

Reference:

Memorandum from John A. Zwolinski, NRC, to S. A. White dated July 15, 1987, " Request for Additional Information on Engineering Employee Concerns" In response to your request for additional information concerning engineering element reports related to Sequoyah' restart, TVA has researched.the pertinent information and the appropriate risponses are included as Enclosure 1 to this letter.

Subsequent to your review of this letter, a teleconference can be scheduled to discuss any needed clarifications or further questions as desired.

Plewse l

contact Ron Gagne (615) 365-3788, Watts Bar, if you need further assistance.

l l

Very truly yours.

TENNESSEE VALLEY AUTHORITY

_ R. LuA James R. Russell Watts Bar Nuclear Plant ECTG Licensing Manager cc (See page 2.):

\\

flet/h/D g-An Equal Opportunity Employer i

u

.s 2

U.S. Nuclear Regulatory Commission ec (Enclosures):

Mr. G. G. Zech, Assistant Director Regional Inspections Division of TVA Projects Office of Special Projects U.S. Nuclear Re6ulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 Mr. John A. Zvolinski, Assistant Director For ProjJcts Division of TVA Projects Office of Special Projects U.S. Nuclear Regulatory Commission 4350 East West Highway EWW 322 Bethesda, Maryland 20018 Sequoyah Resident Inspector Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy Daisy, Tennessee 37379 I

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I Employee Concern Report 20601(B)' Revision 1-1 h

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1.-

Question: Conclusion C of report 206.1(B) dated February 17. 1987' states

)

that "only the control room drawings will be up-dated to red-lined. status before plant restart." What safety-related systems will still need to'be updated?

Response: The post restart scope for the Design Baseline and Verification Program (DBVP) was docketed to NRC on May 12, 1987 in a letter from R. Gridley (L44 870512 802 - included as attachment A) 2.

Question: What is the schedule for updating the drawings of the safety-related systems?

Response: The post restart schedule for the DBVP was docketed in the same letter noted above (attachment A).

j.

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. Employee Concern Report 20501(B) Revision 3 20502(B) Revision 2 1.

Question: At the time of preparation of report 205.1(B) in December 1986, several reviews concerning calculations were in progress or I

were being planned. Are the results of these reviews available?

Response: All engineering disciplines have programs in place to complete the essential calculation and to review the calculations which support modification to safety systems within the pre-restart phase of the design baseline and verification program. This review is not yet complete but corrective action is being tracked per Corrective Action Plan for ECTG Report 205.01.

Additional information may be obtained from NRC Audit Team leaders Ralph Archtizel, Hans Ashar and Shou Ning Hou who have audited the Essential Calculation Program, Design Baseline Verification Program, Cable Tray Support Analysis, Alternately Analyzed Piping and Supports and the Electrical System Review.

2.

Question: Have the results of the reviews caused modifications to the

. plants?

Response

See response to Question 1.

3.

Question: Has the review caused a change in the conclusions or the extent of the corrective action plan?

Response

See response to Question 1.

6156T Pe.ge 1 of 25 s

Employee Concern Reports 215.9(B) Rev. 1 222.5(B) Rev. 2 1.

Question: TVA has committed (TCAB)-049 APP ASN) to sampling mixed connections. When will this be completed?

l Response: DNE calculation for SQNP units 1 and 2, "Samplin6 Program for l

Welds Mixed with Anchor Bolts " SCG-S134 R1 [B25 870611 301]

is now complete.

2.

Question: TVA has committed (TCAB-49A and APP ASV) to analyzing the mixed connections found at SQNP. Will the calculations use nominal drawing dimensions or the actual weld sizes, both diameters, hole sizes, and base plate thicknesses?

Response: DNE Calculation SCG-S134 was performed by TVA based on the as-built installations, i.e.,

actual weld sizes, bolt diameters, and base plate thicknesses. This information was co31ected by conducting field walkdowns.

3.

Question: AISC Structural Steel for Building, 8th Edition Commentary 1.15.10 allows mixed connections if the welds are made after the bolts are tightened.

Does TVA intend to follow this procedure?

Response: The referenced AISC specification commentary discusses both friction and bearing connections. The referenced allowance to combinin6 mixed weldin6 and boltin6 is applicable only to friction connections. The subject of concern for Sequoyah element reports 215.09 and 222.05 is the combination of welding and expansion anchors. Due to the relaxation characteristics of most expansion anchors, their use is properly characterized as a bearinr. connection where such combination with welding is not appropriate.

Thus TVA does not intend to follow the referenced proceduce in designs which are the subject of these element reports.

4.

Question: Our copy of Report 222.5(B) does not contain Page 5 of 12.

Response: The requested page is included as attachment B.

Page 2 of 25

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Employee Concern Report 220.1(B) Revision 1

1.. Question: A portion of the concern states " allow hangers or box anchors or structural features to be acceptable, even when they do not conform to the requirements of detail drawings." Do the drawing notes allow features to be acceptable which do not conform to TVA or AISC design requirements?

Response: The 47A050 series notes provide tolerances and additional instructions for the installation of pipe supports. These notes supplement existing procedures and specifications. They do not allow features to be acceptable which do not conform to TVA or AISC design requirements.

Employee Concern Report 222.2(B) Revision 1 1.

Question: TVA did not respond to the concern, "when all that metal is welded on, the pipe has to get so hot that it could adversely affect the pipe material."

j 1.

Based on actual pipe support measurements for 3/4 inch and 1 inch stainlets steel pipe, could there have been a heat input of sufficient energy and length of time to cause the interior of the pipe to become ' sensitized' and thereby susceptible to corrosion and cracking?

2.

Does the pipe material provide inherent protection against sensitization such as the use of low carbon stainless steel?

l 3.

Does the water chemistry provide minimal corrosion activity by having low amounts of free oxygen and chlorides and sulfates?

Response

Employee Concern Number OW-5-003-001 contained two facets, one of which pertained to the welding attributes quoted in Question 1.

This aspect of the concern was addressed in Element Report WP-15-SQN and was found to be unsubstantiated, therefore the additional questions asked do not require responses. This agreement was reached during a telephone conversation between M. Fields of the NRC and R. Gagne of the Employee Concerns Task Group.

page 3 of 25 m

Employee Concern Report 229.2(B) Revision 0 1.

Question:

Specify maximum and average experienced radioactive airborne concentrations during normal operation in areas of concern.

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Response

For the reasons given in Element Report 229.2(B) (reference 1),

there are only two " areas of concern" where " hot instrument panel and sample sink drains connect to open drain headers.*

i These are (1) the lower elevations (i.e., elevatior,3 693.0 and 679.78 feet) of the Reactor Building outboard the crane wall; and (2) elevation 690,0 feet of the Auxiliary Building, where floor drains may communicate with drains from sample sinks IB and 28.

1.

Reactor Building In the lower compartment of the Reactor Building (i.e.,

containment), the approximate average and maximum experienced radioactive airborne concentrations during power operation, expressed as a fraction of the maximum permissible concentrations (MPCF) listed in Appendix B Table 1, Column 1, of 10 CFR 20 (reference 2), are as follows:

Type of Airborne Average Maximum Radioactivity Concentration Concentration Particulate 0.5 MPCF 20 MPCF Halogens 0.5 MPCF 40 MPCF Noble gases 100 HPCF 500 MPCF Tritium 0.5 to 1,MPCF 5 MPCF reference 3 It should be noted that these airborne concentrations cannot be used to infer an airborne radioactivity hazard to personnel resulting from instrument panel drains connecting to open drains, because:

a.

Personnel do not ordinarily occupy the affected areas during normal operation, although the lower compartment of the Reactor Building outboard of the crane wall may be accessed for limited times under strict health physics supervision.

"Open drain headers" refers to embedded drain headers that may receive drainage from floor drains that have not been plugged.

Page 4 of 25

b. 'The instrument panels in question are essentially racks containing instrument such as gauges, transmitters, and switches, with valves and tubing; the panels are not sample stations. The panel drain connections are capped; therefore, the panel drains normally are isolated from the floor drain header (reference 4).

The panel drain connections are used infrequently and for such maintenance activities as venting and flushing the instrument lines.

c.

Considering the low drainage quantities, partition factors of the isotopic constituents, and long decay times of the stagnant fluids in the instrument tubing, the amount of radioactivity in the instrument panel drains is quite low.

Therefore, compared with airborne radioactivity in the Reactor Building during norn.a1 operation from such sources as reactor coolant system leakage, any airborne contamination hazard presented by these panel drains is insignificant.

These points are stated on page 10 and 11 of Element Report 229.2(B) (reference 1):

"Since the Reactor Building Drainage System has closed equipment drain connections to open drain headers, the notential for backflow and venting exists. Because of the small volumes handled, the low contamination level of the effluent, the large size of the receiving headers, and the elevation differences, backflow of potentially radioactive drainage into the floor drains is unlikely. Similarly, because of the small volumes, the tendency for dissolved geses to remain in the liquid, and the relatively low inventory of dissolved gases that would be radioactive, the venting of such gases through the open flow drains presents an insignificant exposure issue.

Operating temperatures are not sufficient to cause boiling. It must be recognized that the Reactor Building is not normally occupied during operation, when the exposure potential is highest. Any entry to the Reactor Building is made under close administrative control with substantial health physics procedures in place. The exposure potential due to drains is insignificant compared to the other hazards present.

The total exposure within the Reactor Building is subject to continuing ALARA and health physics review.

No changes have been necessary as a result of these reviews."

Page 5 cf 25

2.

Auxiliary Building.

Sample sink cubicles 1B and 2B (Panels 1-L-232 and 2-L-232) are located on elevation 690.0 feet of the Auxiliary Building inside the hot sample rooms (reference 2).

As described on page 14 of Element Report 229.2(B) (refer 2nce 1), the drains from sample sinks 1B and 2B connect to drain headers upstream of floor drain openings.

The rooms in which these floor drain openings are located are listed below (reference 6).

Units 1 and 2 Hot Sample Rooms RHR Heat Exchanger Rooms IA, IB, 2A, 2B Seal Water Heat Exchanger Rooms 1A, 2A Unit 2 Valve Gallery Mixed Bed Demineralized Rooms 2A, 2B Cation Bed Room 2A Seal Water Filter Room 2A Seal Water Injection Filter Rooms 2A, 2B Reactor Coolant Filter Room 2A Unit 2 Valve Vaults (below these demineralized and filter rooms)

The approximate average experienced radioactive airborne concentrations in these rooms is less than 0.25 MPC total (reference 7).

The approximate maximum experienced airborne concentrations in these areas has generally been less than 1 MPC total, except during infrequent activities such as valve grinding (reference 7).

It should be noted that except for the hot sample rooms themselves, and the valve vaults, all the above-listed rooms j

are high radiation areas (restricted), Access Type V, as i

defined in FSAR Table 12.1.2-1 (references 8 and 9).

The filter and domineralizer rooms are completely scaled; access is pocsible only through a sealed shield plug.

Entrances to the j

heat exchanger rooms are kept locked except when entry is

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authorized under a radiation work permit. During the periods i

of access, continuous health physics monitoring is required (reference 10).

Except for the valve gallery and valve vaults, Page 6 of 25

1 those rooms contain only tents or vassels. Therefore, maintenance access to these rooms typically is infrequent. Any airborne contamination that might be generated as a result of sample sink drainage into the open drain header would be insignificant from a health physics standpoint.

The hot sample rooms are radiation areas, Access Type III.

Continuous occupation is fat permitted; access is under administrative control.

The response to Question 2 provides further information related i

to possible generation of airborne radioactivity in these j

spaces.

2.

Question: Describe how ventilation affects radioactive airborne concentration in areas of concern.

Response

1.

Reactor Building Lower Compartment Sequoyah FSAR section 9.4.8 describes the containment air I

cooling system. The CRDM air cooling system and the lower compartment air cooling system provide ventilation of the lower areas of the Reactor Building during normal operation.

Three of the four lower compartment fan-coil assemblies supply a total of 195,000 cfm airflow continuously, and two of the four CRDM air cooling fan-coil assemblies supply a total of 62,500 cfm continuously. These airflows are sufficiently high so that essentially there is no stagnant air in the lower compartment.

Thus, any radioactivity that might become airborne as a result of open floor drains is dispersed immediately (reference 11).

The Sequoyah Reactor Building is provided with a porge ventilation system designed to maintain the environment in the primary and secondary containment within acceptable limits for personnel access during inspection, testing, maintenance, and refueling operations; and to limit the release of radioactivity to the environment.

The purge system is operated on an interim basis to reduce Reactor Building airborne radioactivity concentrations to the levels listed in the response to question 1.

Further information on this system can be found in FSAR section 9.4.7.

2.

Auxiliary Building Elevation 690.0 feet I

As described in the response to question 1, the drains from sample sink cubicles IB and 2B connect to open floor drain headers.

Page 7 of 25

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1 Tha samples taken at these stations are from the following sources (references 5 and 12):-

' ' ' Outlet of boric acid blender i

Accumulator tank header outlet l

-Containment floor and equipment sump discharge l

Accumulator tanks 1, 2, 3, and 4 Steam generator blowdowns 1, 2, 3, and 4 As listed in Appendix C of Element Report 229.2(B), the sample-from the boric acid blender and accumulator tanks are considered " hot" because they do not communicate directly with radioactive systems.

The samples from the containment floor and equipment sumps do contain radioactivity. However, most of the gaseous activity i

vents off in the containment sumps to the Reactor Building atmosphere before samples are drawn. This gaseous activity is then processed by the Reactor Building lower compartment ventilation systems described in part 1 of this response.

Therefore, samples drawn from the containment floor and equipment sumps in sample sinks 1B and-2B do'not present a significant airborne contamination hazard in the Auxiliary Building.

The-level of radioactivity in the steam generator blowdown samples depends on the amount of primary-to-secondary steam generator tube leakage and on primary coolant activity. The amount of allowable tube leakage and the specific activity of the primary coolant are limited by the plant technical specifications. The tube leakage activity is then highly diluted by the secondary side water.- Therefore, the radioactivity concentrations in steam generator blowdown samples are relatively low compared with the sample drawn in sample sink 1A, 2A, and 1C (which are also located in the hot sample room and wh!ch are used for sampling reactor coolant and connected systems). Further, the steam generator blowdown samples may be recirculated through the steam generator blowdown radietion monitor (references 5 and 12). The design allows purging the sample lines prior to drawing a sample, thus j

minimizing the amount of fluid entering the sample sink drain.

I Therefore, the amount of radioactivity entering sample sinks 1B and ?B is relatively low.

The sample sinks are totally enclosed by a cubicle with a fume hood on the ventilation exhaust (reference 13).

The exhaust from each sample sink is l

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page 8 of 25

equipped with a HEPA filter and a booster exhaust fan before it joins the main exhaust duct (reference 14). As described on i

page 14 of Element Report 229.2(B) (reference 1), sample sinks are provided with fume hoods such that ".

any entrained gases (the noble gases would be the major radioactive constituents) would be released at the sample extraction point and vented off in the fume hood."

Thus, for liquids that enter the drain header from sample sinks 1A and 2B, only lodines and particulate would remain.

However, these constituents have a relatively high partition factor; only a small fraction - about 1 percent - of the radioactivity would be available to become airborne.

The response to Question 1 lists the rooms having floor drain openings that connect to the same drain headers as sample sinks 1B and 2B.

SQN FSAR section 9.4.2 describes the ventilating system serving elevation 690 feet of the Auxiliary Building.

All the rooms in question, except for the hot sample rooms themselves, are normally inaccessible areas of the plant (see response to Question 1).

To control airborne activity, the ventilation air is supplied to clean areas, then routed to areas of progressively greater contamination potential. Areas of the building which are subject to radioactive contamination are maintained at a slightly negative pressure to limit outleakage.

In addition, the system has the capability of isolating the contaminated areas from the outdoors. All exhaust air is routed through a duct system and is discharged into the Auxiliary Building exhaust stack.

Upon indication of high radiation in the exhaust air from potentially contaminated areas of the Auxiliary Building or upon an isolation signal from either reactor unit, the Auxiliary Building supply and exhaust fans are automatically i

stopped and low leakage dampers located in the ducts which penetrate the Auxiliary Building are closed to complete the isolation barrier. Two 100-percent capacity gas treatment system filter trains consisting of profilters, HEPA filters, and carbon absorbers are automatically energized, and a reduced j

quantity of building exhaust is diverted through the filter

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trains and discharged into the Shield Building exhaust vent.

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All the rooms in question are connected to the exhaust system only (reference 14).

The rooms are maintained at a negative pransure with respect to the surrounding areas.

Thus, except l

during infrequent periods when these rooms are accessed under l

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.___________________________ _ _ ~

health physics controls, no personnel ere exposed to any airborne radioactivity resulting from sample sink drainage into the open floor headers.

Finally, when ALARA pre-job planning (see response to Question 3) determines that airborne radioactivity is expected to result from a planned maintenance activity or plant modification, local, portable ventilation may be employed to minimize occupational exposures (reference 15).

3.

Question: Third paragraph on page 15 states that " potential exposure of operating personnel is consistent with ALARA guidelines and accepted health physics practices." This is misleading.

Regulatory guide 8.8, Rev. 3 states in paragraph C that "the goals of the effort to maintain occupational exposures ALARA are (1) to maintain the annual dose to individual station personnel as low as is reasonably achievable and (2) to keep the annual integrated (collective) dose to station personnel (i.e., the sum of annual doses (expressed in man-rems) to all atation personnel) as low as is reasonable achievable.

Therefore, potential exposure to operating personnel, which can be avoided, should be avoided in order to comply with R.G.

8.8.

Please clarify the TVA position regarding ALARA.

Response: TVA's corporate commitments to the ALARA philosophy are given in (1) TVA code VIII, Occupational Radiation Protection (Ref.

16); (2) Office of Nuclear Power (ONP) Policy 5.7, Radiological Control (Ref. 17); and (3) the ONP Radiation Protection Plan, noction A, Nuclear Power Plants (Ref. 16).

These three documents apply to all TVA nuclear power plant sitos.

i The Policy section of TVA Code VIII, issued by the Office of the General Manager and approved by the Board of Directors, states, in part:

"TVA endorses the policy of the Nuclear Regulatory Commission for maintaining occupational radiation exposures as low as is reasonably achievable (ALARA). TVA believes the occupational radiation exposure to individuals and the collective exposure to all individuals (including contractors employed by TVA) must be maintained ALARA.

In addition to being a policy ALARA is a viable program requiring management commitment, implementing procedures, and employee involvement to be successful.

Page 10 of 25

TVA includse ALI.RA policy in.its design, purchasing, contracting, construction, maintenance, and operating activities related to its nuclear power program and all other operations or activities which utilize or produce sources of ionizing radiation, including.

. radioactive wastes.

The TVA ALARA program can only be effective through concern and commitment by both employee and management.

Accordingly, affected TVA managers maintain continual oversight and carefully evaluate means by which occupational radiation exposure to their employees can be minimized, exercise sound judgment in the weighing of factors competing with ALARA, institute planned programs for occupational radiation exposure control, and encourage responsible employee participation in ALARA programs.

TVA encourages employees to be aware of potential radiation hazards and to promptly bring these to the attention of their supervisors for appropriate correction."

The Delegation section for Code VIII states, in part:

"Within ONp, the Division of Nuclear Services' Radiological Control Organization plans, develops, reviews, and coordinates radiation protection policy and evaluates policy compliance and the effectiveness and consistency of radiation protection program implementation.

It evaluates plans for proposed and modified nuclear facilities and recommends changes as appropriate.

It provides health physics support to operating nuclear power plants.

. and conducts the environmental radioactivity monitoring programs and environmental radiological assessments for TVA nuclear power facilities.

Each nuclear power plant site implements a planned program of radiation exposure control including incorporation of ALARA requirements in appropriate design, Operating, and maintenance procedures, and ensures that such procedures are followed.

Each plant is responsible for implementation of the radiation protection program, identification of areas needing ALARA improvement, institution of corrective measures, and participation of plant employees in improving the ALARA program.

Each plant maintains routine radiation surveillance and radiation access control of plant areas...

The Division of Nuclear Engineering develops nuclear plant designs and design criteria, which, consistent with other design constraints, include features that will allow the l

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plent to b3 opsrated consistent with ALARA.

It ensures timely processing of ALARA-related requests for design changes and design philosophy improvements.

It reviews TVA and' industry operating experience and coordinates the i

evaluation of design alternatives and new design technology

)

as they relate to ALARA.

It conducts interdivisional and interdisciplinary design reviews on specific plant systems and vendor efforts on new designs."

ONP Policy 5.7 states:

"It is the policy of the Office of Nuclear Power to minimize individual collective exposure to radiation resulting from any activities associated with nuclear plant operation. This policy includes exposure to the populace in proximity of the nuclear plants as well as the onsite employees and contractors."

Part of the intent of this policy is that:

" Reasonable costs be allowed to install (as a backfit if necessary) equipment or structures or instrumentation that will reduce exposure or substantially improve exposure control through better knowledge of processes, exposure rates or cumulative exposures."

The policy provides for establishment of a radiological control program that includes, in part:

" Methods to control radioactive contamination at the source with emphasis on repair, replacement, or redesign of equipment in conjunction with containment of leakage.

Methods to minimize radioactive maintenance inplant discharges.

j

Establishment of comprehensive goals by which to measure performance and reports to management to indicate performance."

The ONP Radiation Protection Plan (RPP) provides the detailed policy requirements for the radiological program within the Office of Nuclear Power, Subsection 4.0 of Section A describes j

the ALARA program for all TVA nuclear power plants.

Pertinent portions of this subsection are quoted below:

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Page 12 of 25

"4.2 Supervisors in ONp who are responsible for radiation protection program implementation or responsible for work performed in radiologically controlled areas shall maintain oversight and in the course of work identify methods for reduction of occupational radiation exposures within their area of responsibility.

4.3 General employee training shall include an awareness program to inform and train all employees who frequent radiologically controlled areas of the ALARA philosophy, encourage them to avoid unnecessary exposure, and make recommendations to reduce exposure and improve the radiation protection program.

4.4 Each site director shall ensure than annual ALARA goals are established for his plant including personnel exposure, radioactive effluents, and the minimization of radioactive waste generated, and the volume of radioactive materials in storage areas.

4.7 Consistent with the ALARA philosophy, engineering controls shall be given precedence over administrative controls in both design and operation of nuclear facilities.

4.8 pre-job planning shall be carried out to identify and to minimize the hazards associated with jobs involving significant exposure potential. A pre-job plan shall be prepared and significant hazards identified for each job where.

(4) The intake of airborne radioactivity (excluding noble gas) by any indj'idual is expected to exceed 10 MpC-hre during one working day.

4.9 A post-job report shall be prepared if pre-job planning is required by this section and if the estimated dose differs from the actual dose by 25 percent or greater. The report shall campare estimated doses with actual deses and explain deviations and document the ALARA actions implemented and their effectiveness. The analysis and the post-job report shall be retained for future ALARA planning.

page 13 of 25

4.10 D3 signs for ntw facilities and modifications to existing faellities that involve controlled areas l

or have radiological protection implications shall reflect ALARA design criteria and incorporate reduction of occupational radiation exposure whenever reasonable.

4.10.4 Each site shall perform cost effectiveness evaluations on all ALARA-related changes and modifications. For ALARA design features and modifications which do not have significant manpower, operational, or maintenance implications, a value of $1000 per life of plant man-rem has been established.

Cost effectiveness evaluations for 1

other design features and modifications shall i

consider, as a minimum, does implications and

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impacta on maintenance, operations, and modification personnel."

Further, subsection 4.6 of RPP Section A requires each site l

director to submit an annual ALARA report to the Manager of Nuclear Power.

The annual ALARA report includes, among other j

things, an evaluation of employee ALARA suggestions, pre-job planning / post-job evaluation (see 4.8 and 4.9 above), and a I

description of design changes initiated primarily for ALARA purposes.

In addition. TVA is currently in the process of revising its I

ALARA documentation. When completed, the process will result in a tiered set of generic documents (policy, directive, elements, and standards) as well as in site-specific procedures and instructions on ALARA practices.

These documents will

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ensure that TVA will more closely implement the guidance of Regulatory Guide 8.8 at Sequoyah.

Further, TVA Corporate Health Physics conducts periodic site audits to identify areas for further ALARA program improvement (reference 15).

Further information on standard TVA practices for implementing j

ALARA in plant design is given in TVA Mechanical Design Guide j

DG-M18.7.1, Radiation Protection (ALARA) Guidelines (also cited in App. A. 6.c of Element Report 229.2[B]) (Ref. 19).

This I

design guide is applicable to all TVA nuclear power plants to the extent, that while it was ".

. written to cover all of the ALARA design features involved in a new plant design effort,

. any single design feature covered can be considered for existing operating plants or for those under construction."

1 (Section 1.1 of the Design Guide) j j

Page 14 of 25

As discussed in these esferences~, the decision to incorporate new design features to further reduce occupational radiation exposures in operating plants is based on benefit-cost analyses. This approach. is acknowledged by Regulatory Guide 8.8 (Ref. 20), Regulatory Position C, in the paragraph immediately following that quoted in the RAI. As stated on page 2 of the Regulatory Guide "' Reasonably achievable' is judged by considerin6 the state of technology and the economics of improvements in relation to al the benefits of these improvements." With respect to control or airborne contaminants, Position C.2.d of the Regulatory Guide states:

" Effective design features can minimize the occurrence of occasional increases in air contamination and the i

concentrations and amounts of contaminants associated with any such occasional occurrences.

Designs that permit repeated, identified releases of large amounts of radioactive materials into the air spaces occupied by personnel are contrary to a program to' maintain occupational radiation exposures ALARA."

Therefore, avoidance of any potential exposure to operating personnel is not required by Regulatory Guide 8.8, particularly where such avoidance is to be achieved by design features.

Section D. Implementation, of the Regulatory Guide, expressly recognizes that for plants such as Sequoyah, which were designed before Regulatory Guide 8.8 was issued:

. 4no substantive design changes will be required.

.unless the design change can prevent substantial man-rem exposures that cannot be prevented by procedural measures and the design change is consistent with the cost-effectiveness principle of maintaining occupational radiation exposures ALARA."

Accepted health physics practices other than design features may be employed in the ALARA effort.

These include, for example, training and instruction of personnel in how to minimize their exposure, access control of radiation areas, installed radiation monitoring systems augmented by periodic surveys, and preparation and planning of maintenance and other activities. The portions of the NSRS Investigation Report I-85-921-SQN quoted on pages 12 and 13 of Element Report 229.2(B) (reference 1) describe some of TVA's procedural practices that minimize the amount of radioactive liquid entering the floor drain system from the instrument and sample panels.

Page 15 of 25

_)

Stetion 9.1 of TVA's Radiation Protection (ALARA) D3 sign Guidelines (reference 19) describes the radioactive 1

contamination reduction features applied to radioactive liquid drains. Subsections 9.1.1c and 9.1.1d state, inpart:

" Drainage having the potential to flash should flow through j

a solid pipe to a collection tank.

.it is generally

]

acceptable practice to use floor drains for drainage that does not have the potential to flash."

As described in the responses to questions 1 and 2, the potential for flashing is low; therefore, it is concluded that the existing drain configuration is consistent with ALARA guidelines and existing health physics practices.

References 1.

Sequoyah Restart Program Element Report No. 229.2(B), Revision 0, Instrumentation and Control Design - Radioactive. Panel Drains into Floor Drains (12/02/86) 2.

Title 10, Code of Federal Regulations, Part 20, Standards for Protection against Radiation, Appendix B, Concentrations in Air and Water Above Natural Background 3.

Telecon from T. McDonnell, Bechtei, to J. Leamon, TVA SQN Health Physics, IOM 1409 (07/02/87 at 1:15 p.m.)

4.

TVA drawing series 47W600, Mechanical Instruments and Controls, SQN:

Sheet 12 R15 Sheet 23. R11 Sheet 28, R18 Sheet 64, R?O Sheet 75, R21 Sheet 88, R6 5.

TVA drawing series 47W343, Mechanical Sampling Water Quality System, SQN:

Sheet 2. R1 Sheet 3, El Sheet 4, R2 Sheet 5. R1 6.

TVA drawing 47W479-5, Revision 14 Mechanical Drains and Embedded Piping (Floor Drains) 7.

Telecon from T. McDonnell, Bechtel, to J. Leamon, TVA SQN Health Physics, IOM 1416 (07/07/87 at 2:45 p.m.)

Page 16 of 25

8.

SQN FSAR Figures 12.1-1 cnd 12.1-5, Radiation Protection Design Features (for pending FSAR Amendment) 9.

SQN Final Safety Analysis Report, Table 12.1.2-1 Access Control Areas. Revision 2

10. ONP Radiation Protection Plan, Section A, Nuclear Power Plants.

Revision 3, 07/15/86 Subsection 5.0, Radiologically Controlled Areas, and Subsection 6.0, Radiation Work Permit System

11. Telecon from T. McDonnell, Bechtel, to L. Klaes TVA NEB Environmental Control Systems, IOM 1410 (07/06-87 at 8:30 a.m.)

12. TVA drawin6 series 47W625, Mechanical Radiation Samplin6 System, Auxiliary and Reactor Buildings, SQN:

Sheet 1, R26 Sheet

2. R30 Sheet 11, R21

15. Telecon from H. Mahlman, TVA, and R. Serbu, TVA DNS Rad. Con. to T.

McDonnell, Bechtel, 07/10/87 at 1:10 p.m. (IOM 1468)

16. TVA Code VIII, Occupational Radiation Protection, Office of the General Manager, September 10, 1986 l

17. ONP Policy 5.7, 0-Q, Radiological Control, 12/06/86

18. ONp Radiation Protection Plan, Section A.. Nuclear Power Plants, R3, 07/15/86, Subsection 4.0, ALARA i

19. TVA Mechanical Desl6n Guide DG-M18.7.1, Radiation Protection (ALARA)

Guidelines, Revision 0, 11/20/81 (ESS 811202 204) 1

20. U.S. NRC Regulatory Guide 8.8, Information Relevant to Assuring that Occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As Reasonably Achievable, Revision 3, 06/78 Employee Concern Report 235.8(B) Rev. 1 1.

Question:

Does the Central Laboratory qualification test include the energy contribution resulting from superheated steam during the l

steam line break accident?

Response

a.

The TVA Central Laboratory qualification test (C186-86-1155) does not envelope the temperature and 1

\\

l 1

Page 17 of 25 I

l

pressure profiles for a " Main Steca Line Break" (HELB) in the Main Steam Valve Rooms (Reference Drawing 47E235-79, R4),

s b.

The TVA Central Laboratory qualification test only envelops the temperature profile for the first two hours caused by "the Most Severe'Steamline Break (HELB)" in the containment. The pressure was n.t recorded during the qualification test (Reference Drawing 47E235-44 and -45, both R3).

Employee Concern Report 235.11(B) Rev. 2 1.

Question: On page 5 of the report, the chronology itemizes an 11/86 K-form which was submitted to TVA. Describe the contents of l

this item.

Response: A K-form is the TVA Employee Concerns Program form used to document the employee concern for disposition.

This form, itemized on page 5 of the report, contains the quotation of the empicyee concern and indicates that Engineering has been j

assigned the responsibility on 11/86 for investigating the concern. The concern is quoted on page 2 of the report.

2.

Question: This employee concern was initiated by TVA operators because of the misleading indication of equipment condition given a W-2 control switch failure.

Corrective action (e) on page 12 states that "All plant operators will be notified of the possible failure mode of W-2 switches and the effect on control board status light indication." Since the operators initially discovered this and have expressed concern for this unacceptable condition, it appears that this is not corrective action but rather acknowledgement of the operator concern. Can the operators perform their duties effectively given potentially misleading egolpment condition information? Please discuss this aspect.

Response: TVA will review all W-2 switches including those previously determined not to require the corrective action change and not replaced.

If this review shows additional switches which should be modified, the operators will be notified of the potential switch problem and a continuity check performed every 31 days and after every switch manipulation for the identified switches.

These actions will ensure that the operators will Page 18 of 25

P F

not have to parform'their duties with misleading information.

This is the work. defined on addendum 1 for the ECTG. report.

235.11-SQN-01, 02, 03, 04, and 05 under I.A.

This work will.be complete by restart.

3.

Question: Supply the projected completion dates for 235.11(B) corrective action items SQN 01, 02, 03 and 04.

Additionally TVA reports which present scope, methods and findings for each corrective action should be submitted to NRC..

. Response: All remainin6 work to complete CATDs 235.11-SQN-01, 02, 03 and 04 (item I.C. on addendum 1) will be complete by 4/88.

Employee Concern Report 241.1(B) Rev. 1

.1.

Question: Did TVA provide Raychem with samples of all affected cables containin6 splices?

Response: No; however, TVA sent the followin6 representative samples of cables.that were to be installed in all TVA plants.

Samples Type and IVA No.

Function Size Conductors Rating Mark No.

1 Med. Voltage Power 4/0 AWG 1/C CPSJ-kV WNC 2

Low Voltage Power 500 MCM 1/C CPJ-00 V WDP 3

Low Voltage Control 12 AWG 2/C

-PJJ-00 V WGB l

4 Signal 16 AWG 2/C Signal-600 V WVA Cables actually tested by Raychem for TVA were:

Test Report TSL-618 (1975): Samples 2, 3, 4 above; Splice insulation used in test - WCSF-Type N TVA Cable 5-8 kV Spilce Testing (1976):

1/C-500MCM XLPE (Shielded); splice insulated with WCSF-Type U from HVS-A--22 5-8 kV splice kit.

2.

Question:

Is Raychem heat shrink tubing WCSF-N the only tubing used for all SQN splices?

Response

Standard Electrical Drawin6 SD-E 12.5.3, R0, effective 09/13/77, stated that splices made in manholes, or cable trenches shall be waterproofed by use of Raychem's WCSF-Type N.

Also WCSF-Type U cable sleeves were acceptable if used with PaSe 19 of 25

I i

i l

cdh2siva. tape S-1024.

Lstor, revisions _of the detwing allowsd l.

the use'of WCSF-Type U in accordance with G-38 Section

(

3.4.1.1.

Additionally,'Raychem Jacket Repair Sleeves were i'

allowed to be used in lieu of heat shrinkable tubing.

However, since splicing materials were not documented during installation it cannot be determined, other than by actual inspections by qualified personnel, if materials other than WCSF-Type N were used as splicing insulation at SQN.

The evaluation team limited its analysis to the scope of the concern splices in manholes and handholes.

3.

Question: Does TVA consider the 100 day test to be a sufficient basis for accepting flooded splices for periods much greeter than 100 days? If yes, please describe the basis.

If not, please indicate how floodirg is detected and eliminated.

Response: Yes, TVA considers the 100-day test to be a sufficient basis for accepting flooded splices for periods greater than 100 days. The reason for this basis is as follows.

During the manufacturing process, cables are typically subjected to an accelerated water absorption test as described in ICEA/ NEMA Standard Publication for Wire and Cable. The test calls for immersing and cable in tap water for 14 days.

Based on a discussion with a cable manufacturer, test results confirmed that immersing cables in tap water in nxcess of 14 days did not significantly change the electrical characteristics.

Because of the similar behavior of cable splices and cables subjected to immersion in tap water and cons)dering the insignificant impact on the electrical characteristics on cables exposed to water immersion in excess of 14 days, it can be concluded that flooding the SQN cables and splices in excess of 100 days will not impact the electrical characteristics of the cables and splices, and they will perform their safety function.

The review of splices in the manheles and handholes provided evidence to document that the splices were submerged for a period of time.

The following tests or being performed to confirm the functionality of these splices after this period of l

submergence.

j i

)

i Page 20 of 25

l All ERCW pumps and MCC cables are being meggered and hi-potted. All fire pump cables are being meggered. Fifty percent of the diesel supply cables are being meggered and hi-potted.

Fifty percent of the diesel auxiliary board cables j

are being meggered. Splice degradation due to immersion will be detected by;the above tests.

TVA now plans to perform maintenance inspections of manholes on an annual basis.

4.

Question: TVA Electrical Standard Drawing SD-E12.5.3 (APP A 5.1. 9/13/77)'

specified splicing procedures and materials for cables of the 5-15KV class. The Raychem tests were apparently only tested to 8 KV.

Please explain the adequacy of the splices for 8-15 KV applications.

Response: At SQN, there are no system voltages higher than 6.9 kV involving cables routed through the underground raceway system.

5.

Question:

Did the Raychem testing include aamples of Raychem heat shrink tubing over 3M tape splice kits?

' Response:

No.

6.

Question: Page 6 of the report states that splices have been in service under adverse conditions for some time without a failure and also that no splice installation records were identified.. What

'are the adverse conditions and how is splice integrity checked and verified especially since there are incomplete records.

Response: The adverse conditions, as specified in the report, were that cables in trays were covered with water and/or mud for an d'

indeterminate amount of time.

Splice integrity is addressed in the corrective action plan (CAP) relative to CATDs 241 01 SQN 01, 02, and 03 submitted on transmittal TCAB-83 dated 03/31/87.

7.

Question: How will manhole and handhole splice electrical integrity be verified?

Response

SMI--317-62 (RIMS Number B25 870618 002) documents the result of the walkdown to inspect splices in all class 1E/CSSC manholes and handholes. During this inspection moisture was found between the Raychem heat shrink sleeve and the cable jacket in some ERCW pump and MCC cables and one fire pump cable. As a re3 ult all ERCW pump and MCC cables are being meggered and hi-potted. All fire pump cables are being meggered fifty percent of the diesel aux board cables are being meggered. Fifty percent of the control cable splices with moisture are being destructively examined. If any of the control cable splices are unacceptable additional splices will be inspected.

Cables and splices with unacceptable test results will be repaired.

Page 21 of 2F

L i

l l

Employee Concern Report 241.2(B) Revision 2 1.

Question: A September 15, 1986 TVA memo from Wallace to Wilson stated that "PIDG connectors on surge suppression networks associated i

with solenoid valves required to perform a safety function must be replaced or soldered prior to restart." corrective action 2 states that safety related PIDG terminals..." will be replaced or soldered if the solenoid valve circuit current exceeds the inductive rating of the circuit contact...".

Explain how this relates to the crimped connections limited current carrying condition. Can the current carrying capacity be conservatively estimated for the crimped PIDG terminals?

Response

If the inductive rating of the circuit (relay) contact exceeds the actual current that the solenoid valve circuit will draw, then the surge suppression network is not necessary to prevent contact damage. On the other hand, it does no harm to leave this suppression network in the circuit where it will act as an enhancement feature. Since the surge suppression network is unnecessary, the adequacy of the PIDG terminals being used in that part of the circuit becomes a moot point. Therefore, it is not necessary to estimate the current carrying capacity of the PIDG terminal crimped to a solid c'onductor.

Employee Concern Report 241.4(B) Revision 0 1.

Question: A problem description regarding use of the "old" Amphenol connector.

Response: The root cause of the Amphenol connector problem was connector thread mismatch which resulted from the vendor using national pipe thread on the conduit fitting which mated with the machine threaded Amphenol connector.

The forced threading of the conduit coupling on Amphenol connectors during assembly caused Amphenol connector thread damage. As a result of thread mismatch damage, the assembly became disconnected and damaged the electrical wiring harness between the engine control cabinets and the hydraulic generators on the diesel engine.

2.

Question: A description of how this problem is avoided with the new corrected connector assemblies including confirmatory test or similar data.

Response: To eliminate the thread mismatch, TVA installed an adapter, part # B02020, recommended by Morrison - Knudson service bulletin SB-1027-1.

New wiring harnesses were installed to replace the damaged harnesses.

Installation completed per MI 6.20.

In addition, electrical tests were completed for diesel generator verification per SI 7-1 and completed satisfactorily.

Page 22 of 25

3.

Question: A documented trail which shows replacement of old connectors with new and successful installation test data or equivalent.

Response: TVA has taken positive steps to correct the problem. These steps were:

08/27/85:

Potential reportable occurrence (PRO) 1-85-272 was generated to document the failure of D/G 2B-B to respond correctly during testing per Surveillance Instruction SI.1.

09/04/86:

Another PRO, PRO-1-85-288, was issued to cover the concern of all loose connectors in all four diesels.

09/06/85:

Maintenance report was written to install Raychem Ziplock sleeve on all connectors.

09/11/85:

Purchase order for adapters as recommended by the diesel manufacturer was issued.

09/11/85:

At the same time, maintenance requests (MRs) were written to install wiring assemblies upon arrival.

10/17/85:

New cable assembly installed on D/G 2B-B per MR A539668 (Attached) 10/25/85:

New cable assembly installed on D/G 2A-A per MR A539669 (Attached) 11/08/85:

New cable assembly installed on D/G 1A-A per MR A539670 (Attached) l 11/13/85:

New cable assembly installed on D/G IB-B per MR A539671 (Attached)

In addition, electrical tests were conducted, after installation, per SI 7-1 and completed satisfactorily. MRs and SI.1 Test Package are included as attachment C.

Page 23 of 25

Employee Concern Report 241.5(B) Revision 0 1.

Question: On the January 9 and January 20 evaluation team walkdowns, was penetration corrosion specifically reviewed?

Response

See responso to Question 3.

2.

Question: Which team member (s) are expert in detecting and diagnosing this corrosion condition?

Response

See response to Question 3.

3.

Question: What percentage of penetrations were inspected by the eyeluation team on January 9 and 20?

Response: The evaluation team did not conduct walkdowns, but reviewed a (Q1,2&3) representative sample of TVA field verification walkdown records that were part of the SQN environmental qualification program. The reviews were directed at :ay deficit'cies identified during the walkdown that could be related to corrosion problems. There were no deficiencies related to penetration corrosion recorded on the field verification walkdown reccrds. The results of the reviews of these walkdown records are documented in the memos referenced in that paragraph (Appendix A, 7.b and 7.c).

The SQN environmental qualification (EQ) program verified that every splice inside containment was environmentally qualified or replaced with one which was qualified. The program con;isted of an inspection by TVA electricians and evaluation by qualified EQ personnel.

In the process of this program 100%

of the penetrations were inspected.

It is physically impossible to actually see the conductors inside the penetration to check for corrosion.

However, the completion of a satisfactory EQ program will prohibit corrosion by guaranteeing that moisture intrusion is not possible.

4.

Question: The GE penetration problem exhibited a decrease of insulation resistance between conductors. Was an insulation resistance check made at Sequoyah?

Response

No specific insulation resistance check was made on the SQN penetrations as a result of the GE penetration problem outlined in IE Bulletin 77-07.

The Bulletin did not require an insulation resistance check. A follow-up " Operating Experience Information Report, dated June 30, 1978, on the problems with containment low voltage control penetration assemblies" was issued by the NRC.

The report concluded that the penetration oroblem is unique to the particular design feature of the Series 100 GE penetration at Millstone Unit 2, and the NRC believes there is no generic issue associated with low voltage penetration assemblies.

Page 24 of 25

m

.However, specific insulation resistance tests were.made on-containment penetration modules.after installation of the penetration assemblies.

In; addition, insulation'reaistance

. tests were'made'on the low voltege cables connected to the penetration modules after completion of cable pu111n6 to. verify the cable 1 nsulation'and penetration module integrity. This 1

response is based on the requirements outlined in "SNP E

Inspection Instruction No. 10. R10. Interconnecting.Cablef

. Termination ~and Insulation Inspection," and our experience with other nuclear power plant construction practices..

i..

J l

Page 25 of 25

_.______._.___.__.__.mm______.__

1 l

l Attachment A

g7/87/1987. 16:35 DNE-SOEo 683 870 7750 P.03 L44 8 7 0 512. 8 0 2 "' '

5N 1578 Lookout Place MAY 12 '987 l

U.S. Wuolear Regulatory Coosaission ATTH: Document Control Desk

~

Washington, D.C.

20355 Centlemens l

In the Matter'of

)

\\

Tennessee Valley

Authority Docket Wes. 50 327

)

t'

$0-328 BASILINE AND YERIFICATICW PROGRAM (D

[

670Ng go Referencet TVA letter fram R. Orldiey to NRC dated Ap[1128,1*??

/\\

In accordance with a TVA cosesitaent within the reference acope and schedule fot Phase II (postrastart) of seguoyah unit 2., enclosed please note that implementation of DSVP activities for the Phase II sy' stems differs, in ocne cases, from Phase I.

scope of Phase 11 may be revisions of Phasa I procedures or may be newly de Also, a,8ystem Modification tussnary Raport will be issued for eac oped.

portion thereof in lieu. of a System Evaluation Report (SYSTER).

s em or pheth 2% scope of systems, only implemented chantes will be r Under the updated design criteria; however, unlaplemented and partially implem e

controlled portion of the protestart program. changes for TVA believes that these differences from Phase 3 tave been justifie results of the Phase I effort, in that the DSVP processes will be optimize

~

while retainin*, the most valuable aspects of Phase I activities.

Enclosuro 2 consists of new commitments made within this i

If you have any questions concerning this issue, please call

~

Both L Nell, of the Seguoyah Sitt Licensing Staff, at (615) 870-1459 l

Very tnaly yours, l

{

78NH888E8 VALLEY AUTHORITY j

0% assa 4 A. Dorner.d my R. Cr141ey, Director i

Nuc1 ear safety and Licensing inclosures een see page 2

07/37/8987 86:36 DE-SOEP 615 B70 7750 P.04 4

\\

I.

II I

8 l

q U.S. Nudear Regulatory Chission MAY 121987 cc (2nclosures):

t Mr. G. c. Zech, Assistant DirectoIr*

for Inspection Programs Office of Special Projects U.S. Nuclear Regulatory Commission 101 Marietta Street, NW, suite 2900 Atlanta, Georgia 30323

%~-

Mr. J. A. Zwolinski, Assistant Director For Projecte p;

Division of TVA Projacts

^2 Offlee of specist Projects U,8. Nuclear Regulatory Commission

,4350 East West Highway EW $32 Bethesda, Maryland 20014 9;

sequoyah Resident Inspector Sequoyah Nuclear Plant.,

2600 Igou Ferry Road 3

Soddy Da,isy, Tennessee 37379 l

MJBspLH E8DtACP ec (Enclosures):

3

't RIM 1, HR AN 72A-C N. L. Abercrombie C&PS-4, sequoyah C. R. Ayees, HR 5N 52A-C R. 8. Jhristanbury, Eli 533 c-M 4

W. H. Hannum, BR IN 768-C W. R. Harding, OSPS-A, seguoyah

7. A. Ippolito, Bethesda Licensin office g;

N. C. Kananas, LP AN 45A-C (2)

J. A. Kirkebo. W12 A12 C-X G. R.~Hullee, 716C EB-C D. R. Nichols.,LP 5N 3028-0 L. M. Nobios, P03-2, sequoyah M. B. Whitaker, Jr., LP $N ?038-C D. L. Williams, W10 885 C-X d

h 9

\\

'N s.

07/27/1987 86 30 DNE-SOEP '

615 070 7750 P.05 4

.p ENCLOSURE,1 PostRE$ TART PHASE (PMAst !!)

DESICW BA8EL!NE AND VERIFICATION PROGR.AH (DSVP) 1HIRQDE.Tf04 1

This discussios reavides the information describing'the postrestart phase of the DBVp.

It defines the specifie systems that are included, attributes of the Phase I scope that are applied, and a schedule for completion.

Phase II will establish the functional baseline for the remaining y

safety-related systems or portions of systems (see attachment'l for list).

Verification will be performed for Postoperatins License (01.) changes to these systems.

The Transitional Design Chanse Control process will be utilized in

{

U preparation for the Permanent Desi n Change Control syster. -

1 5

g The scope of the program is divided into design control activities and 3

5 baseline end, verification activities.

c De' sign control Implementation of improved design control practicos which be5an in Phase %

s#

. will continue.

The Transitional De Phase II until the permanent Desi n, sign contro1 Systern will be utilized during 4

5 Change control System is fully l

A implemented.

The' Permanent DesL n' Control program will be put into effect 5

during Phase II.

Implementation of this program will be5 n by L

4 December 31, 1987, and requires developin5 and ir.plementing new design control j

procedures.

The Change Control Board will remain in effect to assist in controlling the q

change process.

Baseline and Verification i

A.

Safe Shutdown and Accident Hitigation (88/HA) Systems

~

These systems or portions of systems were the subject of Phase I.

Work in Phase !! to complete this dicope is as follows:

1., The C/R data base will be reviewed, updated, and maintained.

l' q

2.

The Restart Design Basis Document (RDBD) design criteria will be revised as appropriate to incorporate additional C/Rs os required.

i l

They. will also be revised considering the INPO Desi n Criteria audit.

I 5

recommendations and testing requirements.

3.

The postrestart modifications and documentation corrections i

identified in phase I will be completed.

4.

Configuration Control Drawings (CCDs) for primary Control Room As-Constructed Drawings (CRADs) reviewod in Phase I will be issued.

1

(

07/87/8987 86837 DNE-SOEP 615 870 7750 P.06

.t 5.

Remaining Safety-Relat'ed Bystems This scope cove s the remaining safety-rsisted rf. items and the safety-related portions of the 88/PA systems escluded in Phase I.

The remaining safety-related sys5ms or portions of safety-related systems are shown on attachment 1.

A new system boundary calculation will be Prepared to define in detail the review boundaries.

F.or this scope, the following four major activities will be carried out:

1.

Design Critoria/ Design Basie Review, update, and maintain the existing C/R data base.

Develop new 0

or expand existing design critoria to cover this system scope.

L 2.

System Walkdown/ Test

d f

system.wsikdowns will be performed to verify the functional e

. configuration *cf the process flow for mechanical and fluid systems.

Instrumentation and control (!&C) systems will include confirmation d

' by walkdown/ tests.that the control device functions and that its

, associated root valve is verified.

For electrical and I&c modifications, results from preoperational j

l q

tests, postmodifiention tests, surveillance instructions, special

sts, and engineering reviews will be evaluated to establish prepar

+

action of modifications since CL.

f

~.mensional data will be obtained, as required, for input to engineering evaluations needed to technically justify the modifiestions.

t f

3.

Evaluation of Engineering change Not.Loes (ECNs) and Other Bystem chanies 9

A technical review will be conducted of implemented ehentes against-the updated design criteria.

The evaluation process will determine if the modifications affect the syst6m's espebility to operate safely and meet the design basis.

i This evaluation will be based upon documented er.gineerin5 analysis, j

I

07474 997 86138 DNE-SOEP 619 B70 7750 P.27 l

t.

I e

The evaluation of other change documents such as Nonconformance Reports /significant condition Reports (NCRs/SCRe), Temporary C

)

Alteration Control Forms (TACTS), Local Design Change Requests (LDcRs), Field Change Roquests (FCRs), and Field Change Notices (FCNs), ensures that engineerin5 evaluations have been performed.

D that. appropriate corrective actions have been taken, and that the h

ML20237K143

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