ML20091C824

Forwards Response to Request for Addl Info Re Facility Proposed Decommissioning Tech Specs & Representative Cost Estimate

ML20091C824
Person / Time
Site:	Fort Saint Vrain
Issue date:	07/30/1991
From:	Crawford A PUBLIC SERVICE CO. OF COLORADO
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
P-91248, NUDOCS 9108070185
Download: ML20091C824 (14)

Text

_

O rubiic service-

== = '.-

P,0. Don M o Denver CO 80201 o M O A. Clegg Crawf ord July 30, 1991 v a renia,mi Fort St. Vrain N """' O' * ' " "'

Unit No. 1 P-91248 U. S. Nuclear Regulatory Commission ATTN: Decument Control Desk Washington, D. C.

20555 ATTN: Dr. Seymour H. Weiss, Direetcr Non-Power Reac u r, Decommissioning and Environmental, oject Directorate Do'.:ket No. 50-267

$UBJECT:

RESPONSE

TO REQUEST FOR ADDITIONAL INFORMATION DECOMMIS$10NING TECHNICAL SPECIFICATIONS AND REPRESENTATIVE COST ESTIMATE

REFERENCES:

1.

NRC Letter, Erickson to Crawford, dated June 7,1991 (G-91121) 2.

PSC Letter, Crawford to Weiss, dated June 6, 1991 (P-91198)

Dear Mr. Weiss:

Attached is Public Service Company of Colorado's (PSC's) response to your request for additional information (Refercnce 1) regarding the Fort St.

Vrain

.(FSV) proposed Decommissioning Technical Specifi-Stions and Decommissioning Representative Cost Estimate.

The attached responses discuss several changes to the proposed Decommissioning Technical Specifications.

PSC will incorporate these changes and submit a revised set of Decommissioning Technical Specifications as a proposed license amendment by August 30, 1991.

As is discussed in the attachment, PSC notes that the detailed FSV Decommissioning Cost Estimate was submitted in Reference 2.

The NRC's concerns regarding the representative _ cost estimate had previously been provided to PSC and were considered prior to submittal of the detailed cost estimate, hD Oh rh67 P

PDR l

L

T P-91248 Page 2 1

July 30, 1991 If you have any questions regarding the attached informatton, please contact Mr. M. H. Holmes at (303) 480-6960.

Very truly yours, g/. &$ft /

A. Clegg Crawford Vice President Nuclear Operations ACC/SWC/Imb Attachment cc: Regional Administrator, Region IV Mr. J. B. Baird Senior Resident Inspector Fort St. Vrain Mr. Robert M. Ouillin, Director Radiation Control Division Colorado Department of Health l

l i.

t l

l

Attachment to P-91248 Page 1 RESPONSE TO REOUEST FOR ADDITIONAL INFORMATION The - following are PSC's responses to the NRC's Roquest for Additional Information regarding the Proposed Decommissioning Technical Specifications and Decommissioning Representativo Cost Estimate, dated June 7, 1991:

NRC Ouestion it General Comment The Technical Specifications (TS) should bo expanded to cover dismantlement activitics (i.e.,

cutting) that may be conducted in the PCRV and require the Reactor Building to maintain subatmospheric pressure.

The TS should also address how the PCRV fluid level will be maintained as well as sealing the PCRV and dealing with 1sakago that may occur while the PCRV is flooded.

PSC Response This question will be addressed in two parts.

The first part deals with the scope and extant of the FSV Docommissioning Technical Specifications (DTS).

PSC considers that the proposed DTS submitted on December 21, 1990, include all the dismantlement activities that require the Reactor Building to maintain subatmospheric

pressure, consistent with the accident analysis in Section 3.4 of the Proposed Decommissioning Plan (PDP).

i The NRC requiroments in 10 CFR 50.36 provido that Technical Specification requirements should be derived from the analyses and evaluations in the Safety Analysis Report.

Further,. ANSI /ANS 58.4 guidance indicates that Technical Specification Limiting conditions for Operation are provided for items when they are relied upon in the Safety Analysis.

In the PDP accident analysis, Section 3.4.8 analyzes a loss of AC power during the cutting of a large activated graphite reflector block, and concludes that the loss of ventilation results in an acceptably small potential releaso.

For all of the analyzed accidents, the proposed DTS adequately bound activities that may be conducted within the Reactor Building.

i The proposed DTS-therefore ensure that off-site dosos to the public are well below 10 CFR 100 guidelines and within a small fraction of the EPA guidelines provided in EPA-520/1-75-001-A, dated January 1990.

Attachment to P-91248 Page 2 The proposed DTS provido requirements for the Reactor Building to be maintained at subatmospheric pressure during all activitios involving activated graphite blocks.

The activation lovel of other PCRV materials, including graphito, concrete, and various metallic items, is significantly loss, as identified in the PDP activation analysis.

In the ovent of a

load drop accident or roloano of cutting debris involving other PCRV matorials, the resultant donos are low enough that subatmospheric conditions are not relied upon.

In addition to the requ romonts for subatmospheric conditions and Reactor Building confinomont integrity, the DTS provide Administrativo controls for a radiation protection program and - for a Decommissioning Safoty Review Committee.

These Administrativo Controls will ensure that activities are conducted in accordance with Radiation Work Pormit controls, as applicabic.

Also, activities that could create the greatest potential for airborno contamination, such ac certain_ cutting operations, will utilizo enginoored controls for radioactive containment.

PSC considers that the raquirements of the proposed DTS are consintent with the safety analysis provided in the PDP, and that the Administrativo controls provide sufficient assurance of radiation protection measures, such that no expansion of the DTS scope is required.

The second part of the question deals with controls on the PCRV shielding water.

Prior to the initial fill of the PCRV, all panotrations which are below the PCRV water lino and have had their instrumentation removed will be scaled.

Scaling will be accomplished 01ther by wolding on cover platos, by cutting and capping (with wolded caps), or by installation of blind flangos.

All scaling devices will be designed and tested per applicable requirements.

It should also be noted that there are two indopondent PCRV water cleanup and clarification systems, so that repair and maintenance on one train will not affect operation of the other.

During the initial f !.ll of the PCRV, the seals will be monitored for leakage and, if leakago is detected, they will be repaired prior to substantially increasing the level.

Af ter the PCRV has boon flooded, the PCRV water cleanup and clarification system will be placed into operation.

Water level in the PCRV will be monitored on the control panel for this system, located on the refueling dock.

Attachmont to P-91240 Page 3 The PCRV water cleanup and clarification system will be designed so that portions of it can be isolated with valves and drained and repaired if.tecessary.

The system will bo pressure tested prior to the introduction of contaminated water and it will be checked for leakago during operation.

PSC proposes to add a discussion j n the Design Features of the DTS, addressing water leakago prevention provisions, but va do not consider that a Limiting Condition is required.

The Loss of PCRV Shiolding Water accident scenario postulated in Section 3.4.7 of the PDP assumes that the entire water inventory of the PCRV is roloased due to a pipe rupture.

The dono analysis conservatively assumes that the theoretical maximum amount-of tritium is transferred to the PCRV shielding water from the graphite blocks.

As such, any leakage that may occur while the PCRV _is flooded would be bounded by the accident analysis in Section 3.4.7 of the PDP.

PSC proposes to add the following to the DTS Design Features:

"4.3 P.CRV Water LeaPago Proventi.QD The PCRV will be filled with water to provide chielding for workers during initial PCRV internal dismantlement activities.

To prevent leakage from the PCRV, all penotrations which are below the PCRV water line and have had their instrumentation removed aro sealed.

Scaling is accomplished with either welded cover platos, welded caps, or blind flangen.

There are two independent trains in the PCRV water cleanup and clarification system, to allow for maintenanco and repair.

Each train has sufficient valvos and drains to allow isolation as requi *ed."

l l

ay &a aa-4

-v g

ny q,

a--

g g

i Attachment to P-91248 Page 4 URQ_QQgytion 21 Pago 3.0-1, Section 3.0.4 Justification for the extension period of 25%

should be provided.

PSC RegggMg1 The extension period of 25% allowed by DTS Section 3.0.4 is consistent both with current FSV requiremonto and with generic HRC guidance for surveillanco frequencies.

Section 2.18 of the existing FSV Technical Specifications allows survol11ance intervals to be extended up to 25%, an doos Specification 4.0.2 of the Westinghouse Standard Technical Specifications, Revision 5.

The NRC guidance provided in Gonoric Lotter 89-14, " Lino-Item Improvements in Romoval of the 3.25 Limit on Technical SpecifAcations Extending Survoillanco Intervals",

also allows a maximum allowable extensloh not to exceed 25 percent of the specified surveillanco interval.

Based on acceptable past practico and on NRC guidanco, PSC considers the allowablo 25% oxtension period appropriato for the FSV DTS.

NRC Ouestion 3 Page 3.1-3,

Background

This section states that now outer truck doors may be added.

Are the new doors required to maintain Reactor Building integrity?

If so, the TS should specify the use of the now doors in controlling releases of radioactivity.

PSC Rosnonsgi LC 3.1 requires that Reactor Building confinomont intogrity be maintained with (among other considerations) either the innor or the outer truck bay closures closed.

The word

" closures" is intended to apply to either the historical hatches and doors or any futuro redundant door installation, as described in the Bases section cited by the NRC.

In this caso, the now outer truck doors would be required for Reactor Building integrity when all innor doors are open.

_m

2 Attachment to P-91248 Page 5 Since these new outer truck doors are considered " closures",

as-defined in the

Bases, they are included in the requirements of LC 3.1.b.1, for assuring Reactor Building confinement integrity, and no additional DTS requirements are needed.

hRC Ougation 4:

-Page 3.1-4, Bases The TS state that the Reactor Building louvers may be open while activated graphite blocks are being dried or stored.

PSC should provide 'tn analysis of the potential release of tritium during the drying process.

P.S.C ERA 2911Alti Although the Reactor Building louvers may be open while activated graphite blocks are being dried or stored, the Reactor Building internal pressure will be maintained subatmospheric whenever activated graphite blocks have been

-removed from the PCRV shielding water and remain inside the Reactor Building, in accordance with LC 3.2.

Therefore, all-gaseous affluents created as a result of decommissioning operations will pass through the Reactor Building-ventilation exhaust system, as was done during normal plant operations.

However, the-ventilation filters have no provision for removing tritium and consequently, no credit is taken -for confinement of tritium.

The position of the

louvers, therefore, has no effect on the amount of tritium released during the drying process.

PSC has' reviewed the amount of tritium that could ?otentially-be released during the drying process.

The quantnty of PCRV shielding water being evaporated from the surface of the graphite blocks is relatively small, compared to the amounts of tritiated water vapor assumed to be evaporated in the Loss-of PCRV Shielding Water accident analyzed-in Section 3.4.7 of the PDP..The PDP analysis assumed -that tritium would be evaporated-from-an 848 square meter-pool, and concluded that the dose to an individual 100 meters from the -Reactor Building would be 34.8 mrem '.or a twc hour period.-.This is a very small. fraction of the 1 Rem whole body dose criteria of-the EPA Protective Action Guidelines cited in the PDP.

Since the quantities of tritiated water vapor released-during drying operations. are bounded by the PDP accident analysis, the consequences of drying operations are also bounded by the l

PDP accident analysis.

l i

Attachment to P-91248 Page 6 NRC Ouestion St Page 3.1-1, Action The completion time allowed to respond to the condition listed should be reevaluated.

If the reactor building confinement integrity cannot be maintained, it is recommended that activities be suspended immediately.

PSC Responset PSC proposes.to revise the completion time to suspend activities in the. event that Reactor Building confinement integrity is lost, from 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

This completion time is consistent with that proposed in LC

3. 2 for the condition _ where-Reactor Building pressure is not subatmospheric.

The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />! completion time allows -an orderly suspension of

_ activities within a _ reasonably conservative time - frame, so

-that further problems are-not created out of actions taken in a-more hurried manner.

Also, a-1 hour completion time. avoids the ambiguity that_ is - inherent with "immediate" action requirements.

1 NRC Ouestion 6:

Page 3.2-1, Actions The. Required Action and the completion Time listed in the table for Action A.1 is not consistent with the required Action described on page 3.2-4 for the same activity.

This inconsistency must be resolved.

PJC Resoonset PSC proposes to-revise the A.1 Action discussion in the Bases to be consistent with the required Action table.

The second-sentence of-the A.1 Action discussion on Page 3.2-4 will be revised to read c.s-followst The one hour completion time to susoend activitieg involvina _ chysical handlina of ACTIVATED GRAPHITE BLOCKS within the Reactor Buildina minimizes _the. time exposure of the-Reactor Building to atmospheric or greater conditions-and:is a conservative time frameL(changes underlined).

i.

This-revision ensures that the Bases discussion and the l-Required Action Table are in agreement.

l l

=

. - --.a

. - ~. -..

Attachmont to P-91248 Page 7 l

NRC Ouestion 71 page 3.2-5, Surveillance Requirements, SR 3.2.1 The Reactor Building subatmospheric pressure surveillanco should be every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during critical activition requiring subatmospheric pressure.

Action B.1 states that subatmospheric conditions can be maintained for about 12 houro.

i PSC ResDonset PSC agrees to reviso Survoillanco Requirement SR 3.2.1 to require a "Once por 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />" verification that the Reactor Building pressuro_is subatmospheric.

NRC Ouestjon 81 Page 3.3-2, Table 3.3-2 The required channel calibration frequency should be on a 6-month interval during decommissioning activities.

l PSC Regponset PSC agrees to reviso Table 3.3-2 to require a 6-month channel calibration frequency for the specified radiation monitors.

l l

l 1

Attachment to P-91248 Pago 8 IfRC Ouest19D_1L Pago 3.4-2, Surveillanco Requirements, SR 3.4.2 The surveillance frequency should be daily while water is being used for shielding.

PSC Roeponsol PSC considers that a requiremont to sample the PCRV shield water daily while it is being used for shielding is an unnecessary burden, sinco each samplo and analysis requires approximately four hours and the PCRV shield water system is expected to be in uso for approximately one year, although the tritium concentration is expected to be very low after about 40 days.

PSC considers that daily sampling until the initial tritium level has boon substantially reduced, followed by wockly sampling until tritium concentration is decreased below 0.01 microcuries/cc, is acceptable and consistent with NRC regulatory guidanco, as follows:

The majority of the tritium within the PCRV is contained within the activated graphite blocks.

PSC anticipatos that the release of tritium into the shield water will occur within a very short timo after the graphite blocks are immersed.

During this initial immersion

period, daily sampling is warranted to monitor the tritium level and ensure that the maximum limiting concentration is not excooded.

Tritium levels in the PCRV shield water will be reduced by a food and blood operation.

As shown on the attached Figure (provided in the Proposed Decommissioning Plan as Figure 3.J-1),

tritium concentration is expected to peak within 10 days after flooding the PCRV, and to be substantially reduced (to less than 0.1 microcuries/cc) within 40 days after flooding the PCRV, Tritium concentration is expected to continue to decrease thereafter.

NRC regulatory guidance f or tritium monitoring programs for occupational oxposure is containod in Regulatory Guido 8.32, "Critoria for Establishing a Tritium Bioassay Program."

This guidance providos critoria for tritium concentrations above which bioassay programs should be established, and frequencies at which routino bioassay sampling should be conducted.

1

Attachment to P-91248 Page 9 The most conservativo criteria in Regulatory Guido (RG) 8.32 for workers where tritiated water is in contact with tho air calls for sampling if the tritium concentration exceeds 10 mic rocurie s,'ec, on a onco por two wooks frequency.

Even if workors can como in contact with tritiated

water, the concentration limit is 0.01 microcurios/cc.

Below those tritium concentrations, a

routino survey program is not required by RG 8.32.

Based on the

above, PSC proposes to reviso Surveillanco Requirements SR 3.4.1 and 3.4.2 to require daily campling during initial filling of the PCRV with shielding water and until tritium concentration decreason below 0.1 microcuries/cc for throo consecutivo days.

After this period (estimated to be approximately 40 days after flooding),

weekly sampling will be required until tritium concentration decreases below 0.01 microcuries/cc.

After this point is reached, no further sampling will be required.

This proposed revision to SR 3.4.1 and 3.4.2 is concorvative with respect to RG 8.32 in that (1) sainpling is required above. 0.01 microcuries/cc, where RG 8.32 only requires sampling above 10 microcurios/cc for comparable applications, and (2) sampling is required on a daily or wookly basis, where RG 8.32 only requirca onco por two wooks or quarterly sampling.

NRC Ouestion 111.

Page 3.4-4, Applicability The LC should be applicable as long as the PCRV has water in it.

PSC Egaggnggi PSC agrees to revise the Applicability of LC 3.4 to "Whenover there is chielding water within the PCRV."

This agreement is subject to the position taken in response to Question 9 above regarding sataple frequencies.

Attachment to P-91248 Page 10 l

NRC Outstion 111 Page 5.0-3, Administrative Controls, 5.3.7

+

The decommissioning audits should be performed at least once overy 6 months for the activities listed in this section.

These activities address significant safety areas that are critical during decommissioning and dismantlement.

PSC Responso!

PSC considers that performing audits on decommissioning activities every 6 months represents an unroanonable burden on our resources.

PSC proposes to reviso Administrative Controls section 5.3.7 to require that audits of decommissioning activities be performed on a one year frequency.

This one year audit frequency is consistent with the current FSV Technical Specifications.

FSV Administrative Control 7.1.3.c requires that the Nuclear Facility Safety Committoo audit conformance of facility operation to the Technical Specifications and various other requirements at least once por year.

This is also consistent with the audit requirements of the Westinghouse Standard Technical Specifications, Revision 5,

Administrative Control Section 6.5.2 (NUREG-0452).

4 Attachment to P-91248 Page 11 NRC Comment:

Decommissioning Representative Cost Estimate The sample cost estimate (WBS NO. 2.3.4.3) described the task to be performed, identified the estimated duration required to complete the

task, estimated crew size necessary to l

perform the task, equipment, and supply. requirements.

The example also estimated _ volume of waste, radiation levels, and radiation exposure resulting from performing the task.

The example also indicated that transportation and burial cost would be developed for each WBS although it was'not included in the example provided.

By providing the cost estimate for each of the identified -areas, NRC's concerns should be adequately addressed.

However, the example provided a

description for an approach for performing the WBS and stated that_ if an alternative - approach is _used.the contingency allowance would be sufficient to cover any differential cost, etc.

This is not an acceptable approach.

If an alternative approach is - being considered, the estimate must address.all the areas -discussed above or identify differential cost

- compared to the initial approach.

PSC Response:

The detailed Fort St. Vrain Decommissioning Cost Estimate was-submitted-to the NRC in PSC letter, Crawford to Weiss, dated June 6,' 1991 (P-91198).

The NRC concerns identified above-had-been relayed to PFC. and Westinghouse during the preparation and prior to submittal of the - detailed cost estimate.

In. preparing the detailed Work Breakdown. Structure (WBE)

Dictionary descriptions, WBS.

Element Descriptions and individual'WBS Element Cost Estimates, alternative approaches were evaluated for technical and ALARA feasibility.

However,-

for cost' estimating purposes, if'an alternative approach was considered in the WBS Dictionary and Element _ Description, l

- only-the highest cost option was included in the WBS Element Cost Estimate.

Therefore, the total - Decommissioning Cost Estimate represents a conservative uppor bound on the cost of decommissioning, an m

t Attachment to P-91248 Page 12 Curies 350 300 250 200 150 100 50 0

0 10 20 30 40 50 60 70 80 Days After Flooding PCRV ESTIMATED TRITIUM ItiVE!1 TORY 111 PCRV WATER SYSTEM I

l

. _ _ _ _ - _ _ _