Text

Responds to NRC Request for Addl Info Re Proposed Decommissioning Plan.Vols I & II of Activation Analysis, Encl
	ML20077B902
Person / Time
Site:	Fort Saint Vrain
Issue date:	04/26/1991
From:	Crawford A; PUBLIC SERVICE CO. OF COLORADO
To:	Weiss S; Office of Nuclear Reactor Regulation
Shared Package
ML20077B907	List: ML20077B902; ML20077B909;
References
	P-91118, NUDOCS 9105160327
	Download: ML20077B902 (987)
	v • d • e

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P.0, Box B40 oene co e a April 26, 1991 Fort St. Vrain A. Clegg Crawford Unit No. 1 P-91118 El,$'8l',",', m.

U.S. Nuclear Regulatory Comission ATTN: Document Control Desk Washington, D.C. 20555 ATTN: Mr. Seymour H. Weiss, Director Non Power Reactor, Decommissioning and Cnvironmental Project Directortte Docket No. 50-267

SUBJECT:

PSC RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION ON THE FORT ST. VRAIN PROPOSED DECOMMISSIONING PLAN

REFERENCES:

(See Attached)

Dear Mr. Weiss:

The purpose of this letter is to respond to the NRC's aequest for Additional Information (RAI), forwarded to Public Service Company of Colorado (PSC) in Reference 1. The RAI was developed based on a preliminary NRC review of the Proposed Decomissioning Plan for Fort St. Vrain Nuclear Generating Station, which was submitted to the NRC in Reference 2. PSC requested a delay in its response to the RAI until April 26, 1991, in Reference 3. ,

The following are provided as attachments to this letter:

Attachment 1 - Detailed responses to NRC questions provided to PSC in NRC Request for Additional Information (Reference 1).

Attachment 2 - Fort St, Vrain Activation Analysis, Appendix A, '

Vols. I and 11 (Rev. B)

PSC will forward a revision to the Proposed Decomissioning Plan by the end of June to incorporate the comitments made in PSC responses contained in Attachment I of this letter. As previously comitted to the NRC, the detailed Decomissioning Cost Estimate and the Decommissioning Funding Plan will be forwarded to the NRC at later dates, The Decomissioning Cost Estimate is being prepared to the I sample level of detail provided the NRC in Reference 4 and will be /

provided by the end of May 1991. As previously reported to the NRC, most recently in Reference 5, the Decomissioning Funding Plan will bu submitted to the NRC in the third quarter of 1991. .

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P-91118 April 26, 1991 Page 2 If you have any questions related to the contents of this letter, please contact Mr. M. H. Holmes at (303) 480-6960.

Very truly yours, sA. Clegg dw 4A' Crawford Vice President Nuclear Operations ACC:CRB/cb Attachments cc: Regional Adminnt-ator, Region IV Mr. J.B. Baird Senior Resident Inspector Fort St. Vrain Mr. Robert M. Quillin, Director Radiation Control Division Colorado Department of Health 4210 East lith Avenue Denver, CO 80220 IDF per B. Gunnerson memo of April 12, 1991:

B. Gunnerson: d du,w & #4u W Date: wer/9/ rh e Ts c cco,J J. Johns : S / M o>- Date: t//f4/9 / %;r , : .iv<d e..<c /cher K. Dvorak : M by/ Date: 9/2.b /94 V. Walker : //hbA Date: v/' <u/f/

M. Ferris :V$/w$s /pr/AMS Mtalc% Date: //2f/f/

T. Borst :l l S u AyT&RsiJu GAdm Date: 4M/f/

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P-91118 April 26, 1991 Page 3 REFERENCES (1) NRC letter, Erickson to Crawford, dated February 8, 1991 (G-91020)

(2) PSC letter, Crawford to Weiss, dated November 5, 1991 (P-90318)

(3) PSC letter, Crawford to Weiss, dated April 8, 1991 (P-91121)

(4) PSC letter, Crawford to Weiss, dated March 22, 1991 (P-91093)

(5) PSC letter, Crawford to Weiss, dated April 5,1991 (P-91096)

i ATTACHMENT 1 PSC RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION 4

RELATED TO THE FORT ST, VRAIN PROPOSED DECOMMISSIONING PLAN

l ATTACilhtENT 1 TO P.91118 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORhiATION RELATED TO Tile FORT ST. VRAIN PROPOSED DECOhihilSSIONINO PIAN NRC Ouestion No.1 (Section 1.2.2 Site Final Release Criterial:

"In addition to the references PSC stated as guidance for both loose and fixed contamination for unrestricted release, JE Circular 81-01 and IE Information Notice 85 92 provide guidance regarding the pre-release surveys of non-radioactive trash disposal and scrap material. The IE Circular and IE Information Notice should be referenced in this section."

PSC Response:

Final site release criteria are fully identified in Section 4.2 of the Proposed Decommissioning Plan, and include the following criteria:

1. Limits for loose and fixed surface contamination established in accordance with USNRC Regulatory Guide 1.86

" Termination of Operating Licenses for Nuclear Reactors".

2. Limits for direct exposure based on NRC interim guidance of less than 5 microR/hr above background (at one meter) for reactor-generated gamma emitting isotopes.

3. Limits for total concentrations of ,~ adioactive materials above background in soil and water will be based on the '

total effective dose equivalent (TEDE) limits established in NUREG/CR-5512, " Residual Radioactive Contamination from Decommissioning".

4. The guidance provided in NRC Circular 81-07 " Control of Radioactively Contaminated Materials" will be used to ensure appropriate survey methods are employed for the unrestricted release of decontaminated items (e.g., tools and equipment) and scrap materials.

5. The guidance provided in NRC IEN 85-92 " Surveys of Wastes before Disposal from Nuclear Reactor Facilities" will be used to ensure appropriate survey methods are employed for the monitoring of segregated waste prior to disposal in a sanitary landfill.

Section 1.2.2 was intended to provide a brief summary of the Site Final Release Criteria. This section will be updated to reference the detailed description of the release criteria contained in Section 4.2.

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Attachment I to P-91118 April 26, 1991 Page 2 NRC Ouestion No. 2 (Section 1.2.5 Schedule for Decommissionina Activities):

"Sectton 2.3.5.2 is referenced as providing a schedule for decommissioning; this section is omitted from the DP. In addition, the schedule provided in Figure 2.3-16 does not show development of the radiation protection program (radiation protection plan and implementing procedures) as a major task in the Planning and Preparations Phase. PSC must provide a schedule for completion of the radiation protection program (including that of the Westinghouse team) which shows that the program is in place prior to initiation of dismantlement."

PSC Resoonse:

Section 1.2.5 of the Proposed Decommissioning Plan will be revised to change the referenced section from Section 2.3.5.2 to Section 2.3.5, and to change the referenced figure from Figure 2.3-16 to Figure 2.3-15.

Figure 2.3-15 will be replaced by the attached figure that incorporates the schedule for completion of the Project Radiation Protection Plan and implementing procedures. As shown, no dismantlement work will be commenced prior to the completion of this milestone.

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j Attachment I to P-91118 l April 26, 1991 Page 4 NRC Ouestion No. 3 (Section 1.3.1 Decommissionina Cost):

"10 CFR 50.82(b)(4) requires PSC to provide a cost estimate. The NK has provided PSC with guidance in our RAls with respect to the cost estimate provided in the Preliminary Decomissioning Plan. The PSC assumption that a Westinghouse fixed price contract is sufficient for NRC purposes is not acceptable.

The NRC must review the cost breakdown of decomissioning by its varicus phases or stages. This is required to ensure that the cost estimates used by PSC and its contractor are realistic and that the licensee provides an acceptable level of financial assurance. The staff is concerned that if these costs are grossly underestimated the decomissioning process may be stopped midway.'

PEC Resoonse:

PSC and its Contractor are presently compiling and organizing the decommissioning costs for submittal to the NRC in May 1991, in a format and at a level of detail that will meet the needs of the NRC.

A discussion of this breakdown and the level of detail have been separately furnished to the NRC by PSC letter (Crawford to Weiss) dated March 22, 1991 (P-91093), to assure that a common understanding and agreement are reached.

Attachment 1 to P 91118 April 26,.1991 Page 5 NRC Ouestion No. 4 (Section 1.3.2 Decommissionina Fundina Plan):

l "A. 10 CFR 50.82(b)(4) and 50.75(e)(1)(ii) require PSC to provide a funding plan for assuring the availability of sufficient funds for completion of decomissioning and that all funds for decomissioning be in an external decomissioning trust fund prior to the start of decomissionins. This was not provided and must be included."

, 'his section states that r$C has not made a decision on how fort St. Vrain decomissioning activities will be funded. This section also indicates that PSC does not have the funding available at this time to support the activities for decomissioning. When will PSC resolve this issue?"

ESC Resoonse:

A. In PSC letter, Crawford to Weiss, dated Ar ril 5, 1991 (P-91096), PSC outlined its most recent plans for submittal of the Decommissioning Funding Plan. in this Lubmittal, PSC informed the NRC that selection of the DECON (ecommissioning alternative is contingent upon approval by the Colorado Public Utilities Commission of PSC's application to repower Fort St.

Vrain as a ccnventional fossil fueled generating unit. This approval is necessary to ensure that sufficient revenue recovery conditions will be present to offset the additional expenditures necessary to implement the DECON alternative, as well as successfully repower Fort St. Vrain. PSC is aware that the Proposed Decommissioning Plan cannot be approved until the Funding Plan is submitted; PSC plans to submit the Funding Plan not later than the third quarter of 1991.

B. PSC is currently in the process of evaluating a suitable funding plan that will meet the criteria of 10 CFR 50.75(e), as identified by the NRC in NRC letter, Crutchfield to Crawford, dated February 26, 1991 (G-91038). As reported in PSC letter dated April 5, 1991 (P-91096), PSC has initiated efforts to evaluate financing methods that will provide reasonable assurance of the availability of funds to decommission Fort St.

Vrain, that will meet the requirements of 10 CFR 50.75(e) and that will be responsive to the guidance provided in NRC Regulatory Guide 1.159 " Assuring the Availability of Funds for Decommissioning Nuclear Reactors." As noted above, PSC plans to submit the Funding Plan not later than the third quarter of 1991.

Attachment I to P-91118 April 26, 1991 Page 6 NRC Ouestion No. 5 (Section 2.1 Decommissionino Alternative):

"PSC has stated that DECON is currently the option being considered for Fort St. Vrain but this option is based on several conditions and PSC may elect SAFSTOR if an acceptable financing mechanism cannot be arranged. 10 CFR 50.82(b)(1) requires PSC to identify the selected option. PSC should provide a discussion of the impact of changing from the DECON option to SAFSTOR. The NRC would have to start over to review that option."

PSC Resoonse:

PSC management fully realizes the adverse situation created by maintaining the possibility of reverting to the SAFSTOR decommissioning alternative, and recognizes that this may cause some inconvenience to all parties. However, due to the significant cost and potentially adverse financial consequences involved, this is not a decision that PSC will take lightly.

Since Fort St. Vrain is out of the rate base, the majority (over 90%) of the decomissioning costs will be directly borne by PSC's shareholders. Therefore, it is incumbent on PSC management to ensure that selection of the decomissioning methodology is based on the minimum impact to the financial health of the corporation, and therefore, its shareholders.

PSC has performed an extensive preliminary evaluation of the cost of decomissioning, as well as possible methods to finance the balance of the funds required to accomplish imediate dismantlement and decomissioning, per the DECON alternative. As noted in the response to NRC Question No. 4, PSC has reached the conclusion that selection of the DECON alternative is the most attractive -

decommissioning option.

Imoact of Future Chance of Decommissioninn Alternative:

Any change in decomissioning alternative will t:e t major inconvenience to both parties, but there are no restrktions in the rule to preclude PSC's from changing its dacommissioning alternative. PSC's Preliminary Decommissioning Plan was based on the SAFSTOR alternative. A change of decommissioning alternative from DECON to SAFSTOR would require preparation and sutaittal of a completely revised Proposed Decommissioning Pl an. In all probability, a revised Proposed Decomissioning Plan bued on the SAFSTOR alternative would utilize the cost information that will be provided in the Decomissioning Cost Estimate currently being prepared and due to be submitted by the and of May 1991.

A Funding Plan to support a SAFSTOR alternative would then use this cost estimate (with appropriate escalation), and be based on a funding plan that would allow accumulation of the required funds over the remaining term of the license, as recommended in NRC letter

i Attachment 1 to P-91118 April 26, 1991 Page 7 (Erickson to Crawford) dated October 4, 1989 (0-89338). This recommendation was the basis for an exemption request submitted in PSC letter (Crawford to Weiss) dated March 19, 1990 (P-90071).

i It is also noted that without approval of the Funding Plan, the EXISTING Proposed Decomissioning Plan cannot be approved and no physical decommissioning activities can commence. Other than the inconvenience to both parties and the delay of preparing, reviewing and approving a new Proposed Decommissioning Plan based on the SAFSTOR alternative, there will be no adverse effect on the safety 3

of the plant, nor should there be any major concerns on the part of the NRC related to maintaining the plant in a safe condition.

5

Attachment 1 to P-91118 April 26, 1991 Page 8 NRC Ouestion No. 6 (Section 2.2 Facility Descriotion):

"This section should include a discussion of the adequacy of on-site and off-site roadways in relationship to expected transport vehicle loading expected from shipment of heavy loads such as concrete blocks and steam generator packages. If any additional roadway construction or improvement is anticipated this should be described."

PSC Response:

During decommissioning operations, PSC does not expect that truck shipments will exceed normal highway axle load limits. It is planned that the large concrete blocks being removed from the PCRV will be further cut and reduced in size, first so that only the radioactive portion of the concrete is disposed of as radioactive waste, and secondly so that the concrete blocks will be of a small enough size for truck haulage within normal axle load limits.

Trucks meeting the highway axle load limit requirements will be used to carry heavy loads such as concrete blocks or steam generator packages.

The large steam generator packages are planned to be shipped by rail. The short section of rail line near the plant will be reinstalled for shipment of the steam generators.

The onsite road within the Fort St. Vrain protected area (that leads to the Reactor Building Truck Bay) is a paved road and is in good physical condition. No additional improvements to this onsite road are necessary. The present access roads from the Fort St. Vrain site to the main interstate highway (1-25) are paved roads, in good physical condition, and are well maintained by Weld County. These access roads have been in continuous use during the life of the Fort St. Vrain facility and have seen considerable truck traffic over the routes.

Discussions were held with the Weld County engineer relative to the condition of the existing site access roads and the expected truck traffic during decommissioning. It Was determined that no new modifications or special improvements would be needed.

_ _ _ _ _ _ _ _ _ - . to P-91118 April 26, 1991 Page 9 NRC Ouestion No. 7 (Section 2.2.2 PCRV and Internal Comoonentsl:

"A. The description of the PCRV does not identify the longitudinal and circumferential tendons that will be detensioned and where the diamond wire saw will be inserted in the tendon tubes. A diagram showing both the horizontal and vertical tendons being removed should be included.

B. A discussion should be included addressing the integrity of the PCRV with the tendons detensioned. The discussion should address the impacts of the removal of several feet of concrete and reinforcement on the integrity of the top section of the PCRV.

C. 1:re spread of contamination during the cutting process must be analyzed. Radiological safety procedures to maintain exposures ALARA must be described in detail."

PSC Resoonse:

A. The sectioned areas on attached Figures 1 and 2 show the top head and core beltline region concrete to be removed from the PCRV. In the top head region, holes will be drilled in the PCRV, as necessary, to provide access to thread the diamond wire to start the concrete cuts. Figure 2 shows the horizontal tendon conduits that are planned to be used to make the horizontal cuts in the core beltline region. Those cuts are planned to be made by feeding the diamond wire through inner horizontal tendons CI 9.5 and CI 9.6 at adjacent layers (e.g.,

layers 5 and 6). The cuts will be made by pulling the diamond approximately halfway through the PCRV at each layer.

Figne 3 is a plan view of the PCRV top cavity and shows the vertical tendon conduits that are planned to be used to insert the diamond wire for the core beltline region vertical cuts.

The vertical beltline cuts are planned to be made by feeding the diamond wire through every third inner vertical tendon conduit.

Tables 1 and 2 (attached) identify the top head and vertical tendons, respectively, to be detensioned and removed. All 24 top head tendon and 90 vertical tendons are planned to be detensioned and removed. Table 3 identifies the horizontal PCRV tendons to be detensioned and left in place, or as noted by the asterisk detensioned and removed.

The above response indic.tes the tendon conduits that are planned to be utilized for these operations. Detailed engineering analyses are currently in progress to confirm the structural integrity of the PCRV based on this dismantlement approach.

I

Attachment I to P-91118 April 26,,1991 Page 10 B. As part if the Fort St. Vrain PCRV decommissioning plan, a hexagonai cut will be made to remove the top head concrete.

The remaining internal components and structures of the PCRV will then be removed and an average depth of the beltline region concrete, including the liner plate, will be removed.

See response to Question 7A for location of concrete cuts in the PCRV.

A structural evaluation has been performed for the modified configuration of the PCRV. The evaluation considered the detensioning and removal of all 24 cross head tendons, all 90 vertical tendons, and all circumferential tendons, (groups 9 through 19). To be conservative, all circumferential tendons (groups 9 through 19) were considered to be detensioned even though it has been shown that only the inner most tendons need to be detensioned.

The modified structure was evaluated for the loadings produced by the dead weight of the PCRV structure and components, the lifting operations of the core support floor, and a design basis seismic event.

A simplified free-body iumped mass model fixed at the basement floor of the PCRV structure was developed for analysis with the STAAD-III ISDS computer code. NRC Regulatory Guide 1.60 design response spectra normalized to the Fort Saint Vrain specific

" double earthquake" ground motions with Regulatory Guide 1.61 damping values were inputs to the analysis along with the other structural loadings.

The results of the analysis provide forces and moments in each

- of the individual cross sections of the PCRV. The forces and moments were used to develop concrete comprescive and reinforcing steel tensile stresses. To be conservative and for ease in stress computation, each cross section was evaluated with only the outer most row of reinforcing steel being active.

Two distinct cross sections were evaluated: the modified top head and the modified belt region. The resulting concrete compressive and reinforcing steel tensile stresses are provided below.

Concrete Reinforcing Cross Compressive Steel Tensile Section loadina Stress (osi) Stress (osi)

Top Head DW <l.0 0.0 LIFT 93.6 0.0 SEISMIC 280.0 983.2 Belt Line DW 107.3 0.0 Region LIFT 74.6 0.0 SEISMIC 934.7 4134.5

Attachment I to P 91118 April 26, 1991 Page 11 The concrete has a compressive strength of f'c - 6000 psi and the (Fy reinforcing

- 40,000 psi ) steel steel,isalthugh conservatively evaluated the FSAR indicatesasthat Grade it is 40

( probably Grade 60 or better.

To determine margins of safety, the worst case load combination of dead weight, lifting loads and seismic event were considered:

U -

0.75 (1.4 DW + 1.7 Lift + 1.87 Seismic)

The total combined compressive stress was compared to the s

" Limit Condition 2" compressive stress allowable of 0.85 f'c or 5100 psi as outlined by the FSAR, Section E.1.2.6.2. The v reinforcing steel margin of safety was determined by directly comparing the steel stress to 90% of the yield stress or 36,000 psi. The margins of safety are summarized below.

I Reinforcing Section Concrete Siggl Top Head 9.87 36.62 Beltline Region 3.36 8.71 Margin - Allowable / Combined Actual Stress In summary, the concrete compressive and the reinforcing steel stresses of the modified PCRV are within allowable limits and provide adequate margin of safety for all loading conditions specified. The potential for cracking of concrete in the modified top head and beltline regions has been reviewed and, considering the relatively. low tensile stress in a conservative -

number of reinforcing bars, cracking due to tension in the concrete is not expected.

C. The concrete cutting operation will be controlled and monitored in accordance with the decommissioning Radiation Protection Program and corresponding implementing procedures. Personnel operating the cutting equipment will be located in low exposure areas to maintain exposure ALARA. To minimize the spread of contamination during the cutting process, potential flow paths (i.e., cooling tubes, tendon conduits, etc.) will be plugged, sealed or the flow otherwise controlled. The spread of contamination will 'be controlled with a water retention and recirculation system. This system will be used to minimize airborne contaminants, contain the water used to cool the diamond wire, and process the concrete slurry created by the cutting process.

Airborne contamination will be monitored during the cutting process, and appropriate engineering controls, including respiratory protection equipment (if necessary), will be used.

Attachment I to P 91118 April 2G, 1991 P - 12 Bioassay will be utilized to assess internal deposition of radioactive materials.

Attachment I to P-91118 April 26, 1991 Page 13 TABLE 1 TOP HEAD TENDONS TO BE DETENSIONED AND REMOVED LOWER TENDONS From Pilaster To Pilaster Tendon Number Face Number Face Number TOR-L-2 1/2 4/5 TOL-U-2 2/3 5/6

-TOL-M-2 3/4 6/1 TIR-L-2 1/2 4/5 TIL-U-2 2/3 5/6 TIL-M-2 3/4 6/1 TIL-L-2 1/2 4/5 TIR-U-2 2/3 5/6 TIR-M-2 3/4 6/1 TOL-L-2 1/2 4/5 TOR-U-2 2/3 5/6 TOR-M-2 3/4 6/1 UPPER TENDONS From Pilaster To Pilaster Tendon Number Face Number Face Number TOR-L-1 1/2 4/5 TOL-U-1 2/3 5/6 TOL-M-1 3/4 6/1 TIR-L-1 1/2 4/5 TIL-U-l 2/3 5/6 TIL-M-1 3/4 6/1 TIL-L-1 1/2 4/5 TIR-U-1 2/3 5/6 TIR-M-1_ 3/4 6/1 TOL-L-1 1/2 4/5 TOR-U-1 2/3- 5/6 TOR-M-1 3/4 6/1

Attachment I to P 91118 April 26,-1991 Page 14 TABLE 2 o VERTICAL TENDONS TO BE DETENSIONED AND REMOVED l

! l l INNER RING Tendon Number Tendon Number Igndon Number VI-l VI-15 VI-29 VI-2 VI-16 VI-30 VI-3 VI-17 VI-31 VI-4 VI-18 VI-32 i VI-5 VI-19 VI-33 l VI-6 VI-20 VI-34 i t

VI-7 VI-21 VI-35 l VI-8 VI-22 VI-36 i VI-9 VI-23 VI-37 l VI-10 VI-24 VI-38 l VI-ll VI-25 VI-39 VI-12 VI-26 VI-40 VI-13 VI-27 VI-41 VI-14 VI-28 VI-42 l

l MIDDLE RING Tendon Number Tendon Number Tendon Number l

L VM-1 VM-15 VM-29 l VM-2 VM-16 VM-30 i VM-3 VM-17 VM-31 VM-4 VM-18 VM-32 l VM-5 VM-19 VM-33 VM-6 VM-20 VM-34 VM-7 VM-21 VM-35 VM-8 VM-22 VM-36 VM-9 VM-23 VM-37 VM-10 VM-24 VM-38 VM-Il VM-25 VM-39 VM-12 VM-26 VM-40 VM-13 VM-27 VM-41 VM-14 VM-28 VM-42 OUTER RING Tendon Number V0-7 l V0-14 V0-21 V0-28 V0-35 V0-42 l

Attachment I to P-91118 April 26,. 1991 Page 15 TABLE 3 HORIZONTAL TENDONS TO BE DETENSIONED INNER HORIZONTAL TENDONS Group No. 9 GrouD No. 10 Group No. 11 Group No. 12 Group No. 13 CI-9.6

Cl-10.6 Cl-11.6 CI-12.6 CI-13.6 Cl-9.5

CI-10.5 (1-11. 5 CI-12.5 CI-13.5 Cl-9.4 CI-10.4 Ci 11.4 Cl-12.4 Cl-13.4 Cl-9.3 Cl-10.3 C1-11.3 Cl-12.3 CI-13.3 Cl-9.2 CI-10.2 C1-11.2 CI-12.2 Cl-13.2 CI-9.1 Cl-10.1 Cl-11.1 Cl-12.1 Cl-13.1 Grouo No. 14 Group No. 15, Group No. 16 Grouc No. 17 Grouc No. 18 CI-14.6 CI-15.6 CI-16.6 Cl-17.2

CI-18.6

CI-14.5 CI-15.5 CI-16.5 Cl-17.1

Cl-18.5

Cl-14.4 CI-15.4 CI-16.4 CI-18.4

Cl-14.3 CI-15.3 CI-16.3 Cl-18.3

Cl-14.2 CI-15.2 Cl-16.2 C1-18.2

Cl-14.1 CI-15.1 Cl-16.1 CI-18.1

Group No. 19 CI-19.6

Cl-19.5

MIDDLE HORIZONTAL TENDOWS

- Group No. 17 Grspo No. 18 Grouc No. 19 CM-17.6

CM-18.6

CH-19.6

CM-17.5

CM-18.5

CM-19.5

CH-17.4

CH-18.4

CM-17.3

CH-18.3

CM-17.2

CM-18.2

CM-17.1

CM-18.1

OUTER HORIZONTAL TENDONS Grouc No. 17 Group No. 18 Group No. 19 C0-17.6

C0-18.6

C0-19.6

C0-17.5

C0-18.5

C0-19-5

C0-17.4

C0-18.4

C0-17.3

C0-18.3

C0-17.2

C0-18.2

C0-17.1

C0-18.1 *

Denotes tendon to be removed for diamond wire cutting.

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Attachment I to P-91118 i

April 26, 1991 i Page 16

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Attachment I to P 91118 April 26, 1991 Page 19 !
NRC Ouestion No. 8 (Section 2.2.3 B0P Contaminated Componentsl:
"This section Ifsts the systems that PSC considers to be the potentially contaminated balance of plant (B0P) outside of the PCRV.
In the sections that follow, a discussion of each system is included. The discussion should include a description of the estimated level of contamination, the approach for decontaminating the system, the estimated volume of radioactive waste for each system, and if the system is to be used in the decommissioning of FSV."
PSC Resoonsql The decontamination and dismantlement of contaminated or potentially contaminated balance of plant (B0P) systems will be done by either (1) decontamination in place, (2) removal and decontamination, or (3) removal and disposal as radivactive waste. The method of decontamination for each of the BOP systems is described in the following paragraphs.
In general, contaminated or potentially contaminated piping, components, and structures, will first be surveyed to determine if decontamination or removal efforts are required prior to release for unrestricted use, Piping that needs to be removed will be drained of fluids, to the extent practical: these fluids will be collected, analyzed, and l l disposed of accorCing to project radioactive waste procedures. Pipe !
will be sectioned using thermal or mechanical cutting means such as I oxy /acetylena, plasma, portable band saw, powered hack saw, or j hydrauDc shears, dt:pending on the pipe size. All potentially , -!
contam natrA 9 pen pipe ends will be covered to prevent the spread of I contarinatt , during handling. Small valves will be removed intact with she piping, while large valves and bulky items will be removed and placed into appropriate waste shipping contatners.
Decontamination of piping systems is not anticipated.
After completion of removal activities, those systems and portions !
of systems that remain in the plant as released for unrestricted use wil3 he de-energized and sealed closed. Blank flanges, pipe plugs, or caps will be installed at open pipe and component ends. Piping systems which are to remain in place will be sealed with plugs or caps which may be removed for subsequent survey. Wire leads will be ,
lifted and taped at the electrical components, power distribution '
boxes, and junction boxes. Embedded pipe meeting criteria that dllows release for unrestricted uso Wii a be Capped, tagged and abandoned in place. 1 1
The aporoach for decontaminating each system is explained telow.
Tables 1 and 2 summarize the estimated volume of radioactive waste and estimated level of contamination for each system.
I
- _ _ _ _ - _ _ - _ _ _ . to P-91118 April 26, 1991 Page 20 System 13 - Fuel Handlina Eauioment The contaminated fuel handling equipment includes the fuel handling machine (FHM, Figure 2.210 of the PDP), five reactor isolation valves (Figure 2.211 of the PDP) and two refueling sleeves (Figure 2.2-12 of the PDP). However, the residual radiation and contamination levels for this equipment are low enough to allow manual disassembly on the operating floor.
Privr to disassembly, the FHM will be used to remove selected PCRV internal components. These may include the control rod metal clad reflector blocks (MCRBs), and the non-control rod hexagonal MCRBs.
The fuel handling machine will be disassembled into its component parts as necessary for decontamination or disposal.
Sleeves will be attached as necessary to maintain a contamination envelope. The body of the FHM will be decontaminated and will be left on the refueling deck if release for unrestricted use limits are achieved. If further disassembly is required for release, the lead shot will be removed and the body will be segmented to segregate the contaminated material from the uncontaminated components. The contaminated scrap will be disposed of as described in Section 3.3 of the Proposed Decommissioning Plan.
If the Reactor Isolation Valves cannot be readily decontaminated by manual means, the valves will be removed from the operating floor and the lead shot removed The shot is not expected to be contaminated or activated. The valve bodies will be disposed off-site according to Section 3.3 of the Proposed Decommissioning Plan.
The Refueling Sleeves will be decontaminated by hydrolaser techniques or sandbl asting, then surveyed for release for unrestricted use, if they cannot be decontaminated, they will be disposed of as described in Section 3.3 of the Proposed Decommissioning Plan. The purge vacuum system will be removed and disposed of as described in Section 3.3 of the Proposed Deccmmissioning Plan.
System 14 - Fuel Storace Facility The fuel storage facility (See Figure 2.2-13 of the Proposed Decommissioning Plan) consists of nine fuel storage wells (FSWs) constructed of carbon steel liners suspended in concrete pits.
The FSWs were used for storing new and irradiated fuel during normal plant operation and may be used to temporarily store MCRBs or graphite reflector blocks during decommissioning. All fuel will have been removed from the Reactor Building prior to
Attachment I to P 91118 April 26, 1991 Page 21 initiation of decommissioning activities. TW actual contamination levels in the FSWs will be determined af ter the fuel has been permeriently removed.
When the FSWs are no longer needed, each of the nine inner storage wells will be decontaminated to the criteria for release for unrestricted use, surveyed, and the top access plugs replaced. The outer wells and the reactor plant water cooling system are not contaminated and no outer well decontamination or dismantling is expected to be required. The water cooling system piping at the Dottom of each pit will be cut open for survey.
Decontamination of the FSWs will be accomplished using a HEPA filtered vacuum. Following var"uming, the wells will be mechanically blasted with sind grit or cleaned using a hydrolaser. Spent sand will be collected in catchments placed at the bottom of the well. The well drain pipe will provide k water drainage during hydrolaser operation. After sandblasting or hydrolasing, the five standoff plates at the bottom of the wells will be removed manu:lly. This will provide access for the final release surveys.
f 4tnor cogonents will be shipped as radioactive waste rather han decontaminated. The well plugs will be decontaminated and
.eplaced and sealed after the release surveys have been completed.
System 16_ ,Jmiliary Eauioment The auxiliary equipment consists of the Auxiliary Transfer Cask (ATC), (Figure 2.2.12 of the PDP), ten Equipment Storage Wells (ESWs), (Figure 2.2-14 of the PDP), the Hot Service f acility (HSF), (Figure 2.2 15 of the PDP), and three shielding adapters (Figure 2.2-16 of the PDP).
The ATC was used to transfer the control rod drive assemblies, reifueling sleeves and the shield plugs to and from the ESWs.
The ten ESWs are carbon steel structures embedded in concrete.
They were used to store the control rod drives ar.d the refueling sleens. The Het Service Facility is constructed of concrett and steel shielding and was used for inspection, repair, mainter.ance, testing and decontamination work.
Figure 2.216 of the Proposed Decommissioning Flan is a general layout of the location of the various fuel handling and storage system components and associated auxiliary equipment on the refueling floor.
All the components of the ATC above the top base (32 ft. 11 in, above the operating floor) will be removed using the Reactor Building crane. A containment sleeve will be used to seal the
Attachment I to P 91118 April 26, 1991 '
Page 22 contaminated ports in the cask and the hoist assembly floor as they are separated. The hoist cover and lift extension will then be lowered to the operating floor and disassembled within a contamination control envelope. The components will either be packaged and shipped for burial or to a licensed facility for processing and final disposition, or decontaminated and released D r unrestricted use.
The remaining structure of the ATC will be decontaminated on site. The internal bore will be decontaminated using mechanical means such as sand blasting or hydrolasing to the criteria for release for unrestricted use. After internal decontamination, the crane will be used to lay the cask body over onto the floor for disassembly and decontamination of the bottom flange. When all surfaces meet the criteria for release for unrestricted use, it will be lifted by the crane and returned to storage on the operating floor.
The three shielding adapters will be decontaminated manually to the criteria for release for unrestricted use.
The ten Equipment Storage Wells are internally contaminated and will be decontaminated to the criteria for release for unrestricted use and abandoned in place. Contamination levels in the ESWs will be determined when they are no longer needed.
After the plugs are removed, the ESWs will be vacuumed using a HEPA vacuum assembly similar to that for the FSWs. After vacuuming, the ESWs will be further cler.ned using mechanical methods as necessary to reduce the contamination to the criteria for release for unrestricted use. After decontamination, the wells will be surveyed for release for unrestricted use. The top, access plugs will be decontaminated. -
replaced and sealed.
Following final use of the Hot Service Facility for de:ommissioning activities, the cell will be - decontaminated.
All equipment will be removed, packaged and shipped for burial or to a licensed facility for processing and disposition. All.
surfaces will be decontaminated using appropriate techniques for the- activity including sand - blasting, hydrolasing, scabbling or other mechanical means. The contaminated shield <
window sections will be removed and decontaminated or disposed of as radioactive waste.
System 21 - Helium Circulator Auxiliariet The auxiliary equipment for System 21 was used to provide a supply of high pressure water for the helium circulator bearing lubrication and a supply of purified buffer helium to )revent in-leakage of bearing water into the primary coolant 1elium.
The major equipment items include buffer helium recirculators, heat exchangers, filters, pumps, helium dryers, chemical mme.,.,..e--me,-c-u-.w-,--y ,-,-on,r*-e-..-.- -..~--*re, e,,-, ,----r--v-
Attachment I to P 91118 April 26. 1991 Page 23 i injection components, containment tanks, and compressors (See
, Figure 2.2 17 of the PDP).
Following the defueling of the reactor, the helium circulator ,
system will no longer be used. It has no function in the '
decommissioning of the facility. j The helium circulator auxiliary equipment is not expected to be contaminated above the criteria for release for unrestricted i use based on historical survey data. However, this system will I be surveyed to determine the acceptability for release for I unrestricted use following final use of the system.
At present, no waste is expected to be generated from this '
system.
System 23 - Helium Purification Auxiliaries The helium purification auxiliary equipment was used to assist in purification of the helium used as the primary reactor coolant. The major equipment items include filters, adsorbers, '
heat exchangers, compressors, and dryers (See figure 2.2-18 of the PDP). ,
Most of the system outside the PCRV is not expected to be '
contaminated above the criteria for release for unrestricted use based on historical survey data. This system will be surveyed to determine the contamination level following final use of the system.
The maximum surface contamination levels and estimated volume of radioactive waste for the helium aurification auxiliary equipment are ideMified on attached Tables 1 and 2.
System 24 Helium Storaae 9.ystem The primary purpose of the helium storage system was to provide for both storage and transt4r of helium from the reactor vessel to the storage tanks. In addition, the helium storage system was used in testing the control rod reserve shutdown system and for Various FHM purging operations. The primary equipment ;
items include-a helium transfer compressor,. storage tanks, oil absorber, and hi h pressure helium supply tanks (See Figure 2.2 19 of the PDP .-
Following the defueling of the reactor, the helium storage system will no longer be used. it has no function in the decommissioning of the facility.
The 108 helium storage bottles and associated system components are not expected to be contaminated based on historical survey data. The systcm will be surveyed following final use of the s
l l
Attachment 1 to P 91118 April 26, 1991 Page 24 system. The results of this survey will be used to confirm the non contaminated condition or determine specific decontamination steps for specific components.
No radioactive waste is expected from this system.
System 46 - Reactor Plant Coolina Water htts The reactor plant cooling water system (Figure 2.2 20 of the PDP) provides cooling water for process heat removal from all auxiliary equipment in the reactor plant. Three loops are provided that form the PCRV circuit (liner cooling tubes), the PCRV auxiliary circuit (closed loop for various systems / components) and the service water circuit (open loop for various systems / components). The major equipment items include surge tanks, pumps, demineralizers, filters, heat exchangers, chemical injection (tank and pump) and recondenser chiller.
The portions of the system external to the PCRV are not expected to be contaminated above the criteria for release for unrestricted use based on historical survey data. The system, however, will be surveyed following its final use. Cleanup requirements will be determined based on that survey.
No radioactive waste is expected to be generated in the decommissioning of the portions of this system external to the PCRV.
The reactor plant cooling water system will not be used for cooling of plant components during decommissioning. It will be .
disconnected and isolated from the PCRV and from the FSWs before decommissioning of those systems occurs.
System 47 - Purification Coolina Water System The purification cooling water system (two loops) providet cooling water to the helium purification system heat exchangers. The major components are pumps, expansion tanks, exchangers and associated piping (See figure 2.2-21 of the PDP).
This cooling water system is not expected to be contaminated above the criteria for release for unrestricted use based on historical survey data. The system, however, will be surveyed following its final use. Cleanup requirements will be determined based on that survey and performed before final survey for release.
There is not expected to be any radioactive waste generated in the decommissioning of this system.
l l
~
Attachment 1 to P 91118 April 26, 1991 Page 25 The purification cooling water system will be isolated from the helium purification system before it is decommissioned. The purification cooling water system has no other use during the decommissioning.
System 61 Decontamination System The decontamination system consists of a water heater, a drying air heater, a filter, pumps, a solution tank and a chemical injection system (See Figure 2.2 22 of the PDP).
The decontamination system will be surveyed to determine the extent and location of radioactive contamination following final system use. The decontamination system components are small and will be removed and packaged in LSA shipping containers with other contaminated components and piping. The decontamination solution tank may be removed in one piece for shipment, or may be segmented and packaged in LSA shipping containers.
System.62 Radioactive liauid Waite System The major equipment items in the Radioactive liquid Wa:te System include a waste sump (1000 gallon tank), pumps, filters, two 3000 gallon receiver tanks, two demineralizers, and a 3000 gallon waste monitor tank (See Figure 2.2 23 of the PDP).
The liquid waste system is expected to be used for its original function during decommissioning operations. Therefore, it will be one of the last systems to be decommissioned.
A characterization survey of the radioactive liquid waste system will be performed, when the system is no longer needed, to determine the extent and location of radioactive contamination.
The contaminated radioactive liquid waste system components are small and include: two liquid transfer pumps, two liquid waste sump pumps, two liquid waste filters, and two liquid waste demineralizers. The liquid waste monitor tank and the two liquid waste receivers may be decontaminated and abandoned in place, shipped as one piece containers, or segmented and packaged in LSA shipping containers depending on the extent and location of radioactive conttmination. The liquid waste sump will be considered for either (1) decontamination to the release for unrestricted use levels and abandonment, or (2) segmented and packaged as LSA waste.
l l
Attachment I to P 91118 April 26. 1991 Page 26 System 63 Radioactive Gas Waste System The- major equipment items in this system include pre filters. I filters, exhaust blowers, tanks compressor (See Figure 2.2 24 thePDP). of(vacuum, surge, and drain), and ,
Following final use of the system, the radioactive gas waste system will be surveyed to determine the extent and location of i radioactive contamination. The large components such as the two gas waste surge tanks, the gas waste vacuum tank and the two gas waste compressors may be decontaminated and abandoned i
- in place, shipped as one piece units, or segmented for packaging and shipping. The other components are small enough to be shipped in LSA shipping containers with other contaminated piping.
Decontamination of these systems will be by manual mechanical methods depending on the levels of contamination found during the characterization survey. The system will not be used in the decommissioning of the plant.
System 72 - Reactor Buildino Drain System The major equipment items include drain tanks, sump, pumps, piping and filters (See Figure 2.2 25 of the PDP). Two gravity flow drains are provided to direct drainage from the Reactor Building equipment, piping and floor drains to either the radioactive liquid waste sump for potentially contaminated liquids or the Reactor Building sump for all other liquids.
The drain system will continue to be -used for its original function during much of the decomissioning work and will be one of the last systems to be decommissioned.
When no longer required to remain operational, the system will be surveyed, and a decontamination and decommissioning decision will then be made. Contaminated piping or components will be either removed and shipped in LSA containers, or decontaminated to the criteria for release for unrestricted use and left in place.
Tiie portion of this system that drains to the Reactor Building sump is- not expected _ to be contaminated. The portion of the system that drains to the radioactive liquid waste sump i s expected to be contaminated and is included in dismantlement and removal plans.
System 73 - Reactor'Buildino Ventilation System The Reactor Building- HVAC system ventilates various - areas of the Reactor Building with heated or cooled air. All ventilation air, whether outdoor or recirculated, is filtered before distribution. In addition, ths reactor plant HVAC
--,,, e e,--.,w e- wm m<-e- e. , - --- . w. e ,,w a -,m y--e --~,,+,.-,<..-,,; m---- .--yrr- 4.- ,-mwe---w-,,-e,-ttww--waw we vw. -'n
Attachment 1 to P 91118 April 26, 1991 l Page 27 I s stem maintains building differential pressure control. As s own in Figure 2.2 26 of the Proposed Decommissioning Plan, i this system consists of several air handling units and filters.
The only part of the system considered to contain possible contamination is the Reactor Building exhaust filters, hot service facility vent, and the analytical room vent. The reactor. plant exhaust filters are composed of banks of moisture i separators, HEPA filters and charcoal adsorbers. ,
Based on historical data, the ventilation system is not l expected to be contaminated above the release for unrestricted i use limits. This system will be maintained during l decommissioning to provide ventilation for decommissioning operations. The ventilation system will be included in the characterization survey. Following final' filter changeout and Interior cleaning, ultimate decommissioning activities will be .'
based on survey data taken at that time.
$vstem 93 - Instrumentation and Controls The portions of the instrumentation and control system that are -
of interest are the process monitors, moisture monitors and other instruments that penetrate the PCRV.
Moisture monitors will be removed during dismantling the PCRV.
All other instrument interfaces to contaminated or potentially ,
contaminated systems will be addressed with the respective systems and. all interfaces will be either removed or verified to be below the Ilmits for release for unrestricted use. All systems are scheduled for inclusion in the characterization
- survey. Contaminated system components will be decontaminated or disposed of as LSA waste.
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Attachment I to P-91118 April 26,.1991 Page 28 TABLE 1 B0P WASTE VOLUME ESTIMATES SYSTEM NO. ITEM VOLUME (FT3) 13 Fuel Handling Machine 192 13 fuel Handling Machine Sand 420 ,
13 Reactor Isolation Valves 960 l 13 Refueling Sleeves 192 i 14 Fuel Storage Well Sand 750 16 Auxiliary Transfer Cask Sand 15 16 Hot Service Facility 384 16 Hot Service Facility Sand 500 16 Equipment Storage facility Sand 225 23 Hydrogen Getter Units 4 23 Hydrogen Removal filter 96 23 Purified Helium filters 14 23 Helium Regeneration Pit Sand 135 61 Decon Chsmical Supply Pump 2 61 Decon Recycle Pump 2 61 Decon Solution Tank 366 62 Liquid Waste Demineralizers 192 62 Liquid Waste Filters 15 62 Liquid Waste Monitor Tank 576 62 Liquid Waste Receivers 1152 62 Liquid Waste Sump Pumps 5 62 Liquid Waste Transfer Pumps 96 62 Liquid Waste Sump Sand 23 63 Gas Waste Compressors 2058 63 Gas Waste Surge Tanks 2646 63 Liquid Drain Tank 20 63 Gaseous Waste Vacuum Tank 980 72 Reactor Building Drain System 125 73 Ventilation Filters 192 93 Instrumentation and Control 225 Core Support Vent filters 15 Small and Large Bore Piping
El TOTAL 13,153 Cumulative from all BOP systems
Attachment I to P-91118 !
April 26, 1991 Page 29 TABLE 2 ESTIMATED LOOSE SURFACE CONTAMINATION LEVELS SYSTEM CONTAMINATION NO. SYSTEM / ITEM DPM/100 CH2 13 Fuel Handling System 200,000 14 Fuel Storage Facility (each FSW) 300,000 l 16 Auxiliary Equipment Auxiliary Transfer Cask Internals 230,000 Hot Service facility 2000 - 20,000 Equipment Storage Wells (each ESW) 480,000 Shielding Adaptors 5,000 23 Helium Purification Auxiliaries 5,000 46 Reactor Plant Cooling Water System (Pending Survey) 47 Purification Cooling Water (Pending Survey) 61 Decontamination System 100 - 1,000 62 Radioactive Liquid Waste System 100 - 2,000 63 Radioactive Gas Waste System (Piping) 100 - 2,800 72 Reactor Building Drain System (Pending Survey) 73 Reactor Building Ventilation (Filters) 100 93 Instrumentation and Controls 500 - 400,000
h Attachment I to P-91118 April 26, 1991 Pye 30 NRC Oyestion No. 9 (Section 2.3.3.1 Overview of PCRV Dismantlement Activities):
"A. This discussion states that the steps discussed for the PCRV dismantlement are very preliminary and may change during the planning stage af ter the engineering drawings are developed.
Also, this section states that many of the assumptions are preliminary estimates, including the wall thickness of activated concrete that will require removal. PSC needs to identify when the detailed engineering analysis and drawings will be completed and how these studies will be evaluated against the initial assumptions for which the decommissioning plan is based.
B. PSC needs to address how the large uncertaintles discussed in the above questions are addressed in their cost estimate. In particular, are the costs based on removing a fixed thickness of the PCRV or are they based on achieving the required level of decontamination? Also will activated concrete be sampled and analyzed after the spent fuel has been removed."
PSC ResDonse:
A. PSC and Westinghouse have agreed upon the specific work scope, and this work scope includes dismantlement of the PCRV. This work scope and the preliminary methods are defined in the Proposed Decommissioning Plan, previously submitted to the NRC.
The activation analysis predicts that approximately 24 inches of PCRV concrete sidewall will be required to be removed.
Westinghouse, by virtue of the proposed removal methodology, will remove a minimum depth of approximately 27 inches of concrete within the existing fixed price.
To support planning and engineering efforts, Westinghouse is preparing a mockup of the PCRV concreto walls to test the diamond wire cutting techniques on the concrete and rebar expected to be found in ti.e PCRV walls. PSC is also preparing a detailed cost estimate that will be sutmitted to the NRC in May 1991, that will identify the expected costs associated with the concrete removal process.
PSC is continuing to perform verifict. tion of its activation analyses to verify that the amount of concrete to be removed was properly identified and that th6 specified work scope is adequate to remove -this amount of concrete. It is PSC's opinion that these steps provide reasonable assurance that all activated concrete will be removed The engineering required to finalize the PCRV dismantlement methods and techniques, and the technical requirements for dismantlement and decommissioning the components, systems, and facilities is scheduled to be completed as part of early I
Attachment I to P 91118 April 26. 1991 Page 31 project activities. Upon completion of this engineering effort, the necessary equipment will be procured, leased, or designed and fabricated, and the detailed work instructions will be written. This latter effort will be performed on a schedule to meet the needs of the dismantlement operations in the field. For example, the removal of the (primary) steam generators will not start until approximately 2 years after the beginning of physical dismantlement activities at the site.
The dismantlement methods, techniques and technical requirements will be established early; however, the design, l procurement, and fabrication of tools, hardware, and waste shipping containers, and the writing of the detailed work instructions will be done on a schedule consistent with the performance of the work in the field. As the detailed engineering analysis, drawings, and work instructions are !
completed, the initial assumptions upon which the !
Decommissioning Plan is based will be reviewed. j If a change is required, it will be evaluated against the approved Decommissioning Plan and the Decommissioning Technical Specifications. Proposed changes will be administered in accordance with the guidelines identified in Section 1.4 of the Proposed Decommissioning Plan.
B. The fixed price contract for the decommissioning of Fort St.
Vrain is written so that the Contractor is responsible for dismantlement, removal, and decontamination of the PCRV and its components to the extent it has been specified to be contaminated and activated. For example, the Core Support Floor (CSF) is required to be removed and disposed of by the Contractor within the scope of work and within the fixed price.
The Contractor originally proposed a technically feasible method and now is considering alternate methods that could be used. If the Contractor chooses to utilize a different method that is technically satisfactory, it will be done at no additional cost to PSC. In the area of the contaminated or activated side wall concrete, PSC performed a conservative activation analysis to determine the depth of concrete to be removed. The activation analysis is discussed in detail in the response to NRC Question 29. In specifying this depth in the contract, considerable uncertainty was removed from the scope of work, which a fixed price contractor would have had to account for by including additional costs for the possibility that greater amounts of concrete would have to be removed. A specific, detailed cost study is presently being performed that will provide additional assurance that the uncertainties which exist in the Plan are bounded by the cost estimate.
PSC does plan to sample and analyze the PCRV concrete to confirm the depth of activation after the spent fuel has been removed.
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I Attachment 1 to P 91118' April 26, 1991 Page 32 NRC Ouestion No. 10 (Section 2.3.3.6.1 PCRV Floodinal:
"PSC should discuss how the PCRV penetratfons will be sealed prior to flooding and what testing will be performed to verify sealing."
l PSC Resoonse, i The PCRV penetrations listed in FSAR Table 5.8-1 will be sealed except for the top head penetrations which will not require sealing since the PCRV water level will be maintained below that level.
The remaining penetrations will be sealed as follows:
(1) Steam Generator Penetrations: After removal of the secondary modules, the penetrations will be sealed with welded cover plates. The cover plates will be designed to withstand a 115 foot head of water. The root passes will be liquid penetrant tested.
(2) Access Ooenina: The bottom dome of the access opening will be '
removed to increase the size of the opening.- A cover plate will be welded over the opening. The cover plate will include the suction and fill nozzles to the water shield and filtration system. The root pass will be liquid penetrant tested.
(3) Helium Circulator Penetrations: After removal of the circulators, the openings will be sealed by either bolted or welded cover plates. Welding will be performed in accordance with applicable codes.
(4) Remainina Penetrations: The remaining penetrations will be
sealed by either one or a combination of the following:- cutting and capping just outside the PCRV or by installation of bolted and gasketted blind flanges. Where welding is utilized, all root passes will be liquid penetrant tested as per applicable code requirements.
During the initial fill of the PCRV, all possible sources of leakage will be inspected and repairs made as necessary.
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Attachment I to P 91118 April 26. 1991 Page 33 NRC Ouestion No.11 (Section 2.3.3.6.2 Installation of Water Cleanuo :
and Clarification System):
l "A. Will the water cleanup and clarification system be cross connected to the existing FSV Ifquid waste processing system for monitoring and controlling releases of contaminated !
water effluents? If not, describe in the appropriate section 1 how the contaminated water will be monitored and releases controlled to 10 CFR Part 20, Appendix B, release levels.
B. PSC should provide information on the types of filters and ion exchange resins that will be used in the water cleanup and clarification system, how this cleanup system is consiste:;t with NRC waste treatment system recommendations in Regulatory l Guide 1.143, and what procedures will be used for handling l wastes from this system." l PSC Response:
A. The PCRV Water Cleanup and Clarification System will be cross-connected to the fort St. Vrain Radioactive Liquid Waste System via the reactor building sump discharge line at a point in the system that will allow utilization of the existing flow rate and radioactivity monitoring and controls.
B. The final design of the PCRV Water Cleanup and Clarification System is scheduled to be completed by August 21, 1991.
Filters used will be specified to accommodate carbon dust, rust particles and debris from cutting operations. Ion exchange resins will be specified to accomplish removal of radionuclides from the water. The recommendations of Regulatory Guide 1.143 as well as ALARA considerations will be used in the design.
Radiation protection procedures will be written and implemented for the handling of wastes from this system.
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Attachment 1 to P 91118 l April 26, 1991 '
Page 34 i
NRC ' Ouestion No. 12 (Section 2.L3.7 PCRV Too Head Concrete and !
Liner Removal):
"A. Describe the radiation protection features to be provided for the sultiple work stations on the -PCRV work platform to minimize personnel exposures to airborne particulate radioactivity, tritium, and direct radiation during removal of core components from the open PCRV.
B. PSC should provide specific information on the methods - and i equipment that will be used to remove the remaining concrete
~
layer and liner plate. For e ; ample, how will the circular ;
trough around the outer periphery be cut? What thermal-cutting methods will be used? How will these methods affect the integrity of the remaining concrete in the head?"
PSC ResDonse:
A. The primary radiation protection feature for workers on the work platform is provided by performing core dismantlement underwater. All highly activated core components will remain !
submerged in water. This will provide attenuation from any
, direct radiation to workers on the platform. The shield water also minimizes airborne particulate radioactivity. By dismantling the PCRV core underwater, particles that would potentially become airborne in a " dry" environraant will remain suspended in the PCRV shield water. When core components are lifted out of the shield water, their surface will be wet.
This will also help to minimize particulate airborne contamination.
Another radiation protection feature for the workers on the * '
work platform is the PCRV Water Cleanup and Clarification System. A- description of this system is provided in Section 2.3.3.6.2 of the Proposed Decommissioning Plan. In this system, demineralizers will strip soluble radionuclides from the shield- water. Any activated solid particles will be removed by stepwise filtration. Tritium inventory will be controlled by bleeding a side stream of demineralized and filtered water, then diluting the effluent to below 10 CFR 20 limits prior to release. The PCRV will be flooded well before beginning core dismantlement operations that require shielding.
The . PCRV- Water Cleanup and C1arification System will- be - in operation imn.ediately after the' vessel is flooded and throughout dismantlement operations. Monitoring of the water system for tritium and gamma emitting radionuclides wil.1 be performed at regular intervals to ensure proper worker protection is provided. ,
The work platform floor will cover the entire exposed water surface. A basic design element of the platform requires that it- rotate so that workers can be positioned- over one of the
i Attachment I to P-91118 April 26, 1991 Page 35 three covered tool slots over the work region. The platform floor will provide additional shielding against direct radiation from the core components being removed.
A fourth radiation protection feature for workers on the work platform is the platform's ventilation system. Air supply ductwork will be located above the work platform. Exhaust ducts will be located between the bottom of the work platform and top of the flooded PCRV cavity. A positive downward flow of air over the workers and through the tool slots will ensure that a clean supply of air will be delivered to the top of the work stations. This air will then flow underneath the platform and sweep across the top of the water. Exhaust ducts below the plat form will carry the air through HEPA filters to the existing plant ventilation exhaust system for additional filtering and discharge. The air circulation flow direction will minimize accumulation of airborne contaminants under the platform.
Finally, procedures will be developed for workers to operate the work platform, and to wipe and bag core components, tools or other equipment removed from the PCRV. The procedures will be written to minimize exposure of work platform personnel. In particular, the procedures will provide for monitoring components that have the potential for significant personnel exposure during dismantlement. Handling of core components will be performed in a manner to minimize both the time spent and the number of personnel in the vicinity of the work station, and maximize the distance froc tto radiation source to personnel. Temporary shielding will also be used as required to minimize personnel exposure.
B. The PCRV concrete head will be cut by the diamond wire method and removed in sections leaving a hexagonal cavity.
Engineering evaluations are underway to determine if the lower portion of the PCRV concrete head should be cut coincident with the top of the PCRV liner, or if up to 12 inches of concrete should be left uncut.
After the concrete head sections are removed, the remotely operated hydraulic mechanical breaker will cut a circular groove or trough in the remaining concrete to expose the top knuckle radius of the main cavity liner. Two additional parallel concrete cuts will be made by the hydraulic breaker to divide the circular region into three sections. Small portions of concrete will be removed to expose selected areas of the cavity liner so holding and lifting points can be attached in each section. Rubble and debris will be cleaned up by conventional methods. This will be completed before breaching the PCRV liner. No significant degradation is expected to surrounding concrete except for a somewhat rough and irregular surface. Experience shows that overbreak and crack propagation l
l
)
Attachment 1 to P-91118 April 26, 1991 Page 36 from a hydraulic breaker is minimal. The PCRV will be flooded with water and the PCRV Water Cleanup and Clarification System will be cperational prior to entry into the top head region once the PCRV concrete upper head has been removed.
Following all concrete removal operations, holding and lifting hold points will be attached to the liner. Shielding will be provided, as needed, during subsequent cutting operations.
Ventilation system ducting will be directed into the cavity to keep a clean flow of air over the liner head if needed.
Oxyfuel Cutting - Acetylene (OFC-A) will be used to make intermittent cuts in the exposed liner plate. The torch )ath for the circular cut will be made so all of the top lead knuckle radius is removed. The inside of the PCRV cavity contains three layers of insulation, a steel plate, another layer of insulation, and a second steel plate. Tests are planned in the Decommissioning Planning Phase to determine whether the OFC A torch can cut the interior insulation when the initial intermittent section cuts are made. if the cutting operation does not extend through all layers during the initial torch pass, insulation layers will be mechanically cut and the steel plates cut with a torch.
When the intermittent cuts are completed, a crane can be attached to the lif t hold points for one of the three sections.
Exposed liner not previously cut will now be removed to free this section of the top head. This process will be repeated for the remaining two sections.
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Attachment I to P-91118 I April 26, 1991 Page 37 i
NRC Ouestion No. 13 (Section 2.3.3.8 Dismantlina PCRV Core l Components):
"A. Describe the radiation protection provisions for dewatering graphite blocks and other core components to minimize personnel {
exposures to airborne particulate radioactivity, tritium, and ,
direct radiation during preparation for compliance with '
shipping and disposal requirements.
B. This section refers to Table 2.3-1, " Estimated Contact Dose Rates for Graphite Blocks," which estic'.tes the dose rate from the graphite blocks. The calculations that support the estimated doses are not reference) or included. PSC must provide the basis to support this estimate.
C. The decommissioning plan states that special procedures will be developed for removal of high activity core components (500 mReWhr or greater). pSC should identify those components and provide a discussion on how the components will be removed and handled including the specific methods to be ustd. Specific methods for ensuring occupatfonal safety should be discussed, especially when components are handled outside of the water shield.
D. The decommissioning plan states that existino lifting holes will be used to remove the various reflector Olocks. How will submersion in the flooded PCRV affect the integrity of the 1iftiny holes?
E. PSC should provide a detailed discussion on how the Hastelloy cans will be removed from the graphite blocks (e.g., spearing with a pick) and how occupational safety will be ensured.
F. pSC should provide specific information on how reflector blocks will be packaged."
PSC Response:
A. Graphite block water testing is currently being performed to determine the water absorption characteristics of the graphite blocks and the test results will be used to determine the best methodology for dewatering graphite blocks. Preliminary results indicate that the graphite blocks absorb minimal amounts of water.
For graphite blocks with low exposure rates, a simple wipe down of the blocks with long handled tools and packing for disposal with water absorbing material may be adequate. See the response to Question 130 for handling methods for graphite blocks with high exposure rates. Implementing procedures will provide for the use of shielding, engineering controls, and
i Attachment I to P-91118 l April 26, 1991 1 Page 38 '
respiratory protection equipment during the processing of i graphite blocks and other core components.
B. Calculations to determine the contact exposure rate for various PCRV components were carried out in a conservative manner to estimate shielding requirements for handling. For the graphite blocks, uncertainty exists in the impurity levels that produce the radioactivity. Inspection of the calculated radioactivity levels indicates that tie major exposure contributor is Co-60.
Therefore, all calculations are based on the radiation from this isotope.
For the purpose of developing a conservative estimate, the graphite blocks were considered to - be infinite in size and contain a uniform _ density of Co-60. Using- formulas for an ,
infinite slab source and data for gama absorption of Co-60 in I graphite, it is found that the exposure rate is equal to 7.4ES R/hr per Ci/cc for Co 60 gama rays emanating from graphite (ref: The Photon Shieldina Manual by A. Foderaro). Co 60 values in the graphite are based on conservatively estimated cobalt impurity levels in the graphite since no measurements are available at this time.
Results for exposure rates from other components (Hastelloy cans, boron pins, etc.) are based on similar simple formulas, eithar infinite slabs or point sources. Due to the _much smaller size of the cans and pins, contact exposurc rates are very high but the exposure will decrease much more rapidly with distance than for the large graphite blocks.
Table 2.3-1 of the Proposed Decomissioning Plan identifies estimated doses based on preliminary estimates made during the proposal preparation phase. Since that time, the activation analysis has been updated (EE-DEC-0010 Rev. B) and Table 2.3-1 (Table 4 of this document) will be updated to reflect these improved-estimates.
It should be emphasized that these exposure rate values are intended to be estimates. Every ar. tempt was made to make these estimates conservative by using conservative impurity estimates and conservative assumptions as to source size. Measurements are therefore expected to show that, the estimates are high and that actual exposures will be-lowe . In addition, preliminary plans for the dismantlement have provided contingencies for higher than expected radiation leic!s. Therefore, it is unlikely -that measured levels will necessitate increased shielding requirements and result in increased costs. Primary reliancs in determining the final handling procedures will be placed on the actual measured radiation levels.
C. Table 2.3 1 of the Proposed Decomissioning Plan provides j estimated contact exposure rates for the various core
Attachment I to P-91118 April 26, 1991 Page 39 components. Tables 3.3-1 and 3.3 5 of the plan are based on
} the estimates of Table 2.3-1 and are used to determine waste classification, volume reduction and transport container type.
Table 5 summarizes this information and includes methods of how highly activated components will be removed from the core.
These methods consider ALARA, cost and benefit trade-offs, and will include applicable regulatory requirements. Presently, there are three handling situations being evaluated:
1) graphite blocks with no pins or Hastelloy cans inserted,
( 2) blocks with pins inserted which will require dumping,
3) the bottom reflector blocks containing Hastelley cans.
Common to all three situations will be the use of the PCRV water shield to minimize exposure to personnel working on the work platform.
1. When the radiation levels of graphite blocks require the use of shielding for removal, a shielded container designed to fit through one of three work openings on the platform will be lowered to the top of the core. A remote handling tool attached to a crane hook and guided by workers on the platform will be utilized to attach and load blocks into the shielded container designed to carry blocks. When loaded the container will be remotely covered, lifted above the PCRV water line but below the work platform and allowed to drain. The container will be checked by radiation protection personnel to ensure acceptable radiation levels are not exceeded. The container will then be removed from the PCRV and taken to an area where the contents will be classified, prepared for shipment, and loaded into a shielded shipping container, ,
2. For graphite blocks containing pins, one approach is that the pins be dumped out of the block under water before proceeding with the removal of the block as outlined above.
A shielded container will be lowered to the top of the core. A pin dumping station will be designed into the work platform. As discussed in Method I above, a long handled tool will be used to move the block to the dumping station where it will be set into the dumper and unlatched. The pins will be dumped underwater into a shielded container supported by the work platform. After the. pins have been dumped, the handling tool will be re-attached and the block loaded into the shielded containar. The graphite block will then be handled as in Method I above.
The pins would be handled in a separate shielded container.
An alternative method is to lift the blocks with the pins into an adequately shielded container, and transport this container to the Hot Service Facility where the pins will be removed. The graphite blocks (without their pins) will be taken to a shipping area, classified, and prepared for
Attachment I to P 91118 April 26, 1991 Page 40 shipping. As an alternative, the pins may be left in the graphite blocks and disposed of with the blocks as radioactive waste.
3. The Hastelloy cans in the bottom reflector blocks are 0.51 inches in diameter and 8 inches long. They are inserted into 0.53 inch diameter holes in the graphite blocks.
These cans are not expected to fall out of the block; therefore, a dumping or tipping operation to remove the cans will not be attempted. An adequately shielded container will be utilized to transport the bottom reflector blocks with the Hastelloy cans. The water level in the PCRV will be maintained such that exposure to workers on the platform is maintained ALARA. As in the previous two cases, a shielded container designed to carry the bottom reflector blocks will be lowered by crane through the work opening to the top of the core. When i loaded, the container will be remotely closed underwater !
and withdrawn from the PCRV. Health Physics personnel will monitor radiation levels during withdrawal operation and just before the container is pulled above the water level.
The container will be maintained above the water level and below the work platform until it has drained. The shielded container will be transported to the Hot Service Facility where the Hastelloy cans will be removed. The blocks with their Hastelloy cans removed will be taken to the shipping area, classified, and prepared for shipping. As a further alternative, the Haste 11oy cans could be left in the graphite blocks and disposed of with the blocks as radioactive waste.
D. Submersion of the graphite block in the flooded PCRV is not expected to degrade the structural integrity of the lifting holes. Discussion with the gra)hite block manufacturer has indicated the integrity of the block would not be affected by submersion in water even at an elevated pH. This will be confirmed by observations during the graphite block testing currently being performed by Westinghouse. In another test program of graphite blocks, the British Telecommunications Research Center assessed the durability of graphite blocks under simulated deep ocean pressures. The blocks were immersed in water and hydrostatic pressure applied up to 13,456 psi.
The results demonstrated that the structural integrity of the blocks was maintained and no evidence of cracking was observed.
E. The bottom reflector blocks with Hastelloy cans inserted represent the highest level of irradiated components to be removed from the core. The bottom reflector block will be removed from the PCLV as described in the response to Question 13C. At present, several approaches for removal of the Hastelloy can are being considered for performance in the l controlled environment of the Hot Service facility. Possible
Attachment I to P 91118 April 26, 1991 Page 41 approaches include piking, cleaving / cutting, drilling /
broaching, tipping or leaving the cans in place for disposal l with the blocks. The approach selected will best meet the requirements of ALARA and personnel protection, minimal handling, and cost and benefit trade-offs. Once the cans have been removed, the graphite blocks will be stabilized, classified, and prepared for shipment.
F. Packaging will be selected based on the reflector specific l activity, exposure rate and disposal option. Reflectors with j significant exposure rates, in excess nf 500 mrem /hr, will be packt.ged in appropriate disposal contsiners for transfer to the dtsposal site. Reflectors with exposure rates less than 500 aRem/hr may be packaged for transport to an off site vendor for volume reduction and subsequent disposal. In all cases, packaging selected will comply with requirements specified by 49 CFR,- 10 CFR 71, and the Disposal Facility Site Criteria.
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Attachment I to P 91118 :
April 26, 1991 i Page 42 I
TABLE 4 ESTIMATED CONTACT DOSE RATES Of CORE COMP 0NENTS Estimated Transport No. of Contact Cask Description Blocks Dose Rate Reouired l 1) Defueling Blocks 1,482 <1 mrem /hr No l
l 2) Top Reflector Blocks 1,215 500 mrem /hr No 1
3) Bottom Reflectors 589 500 mrem /hr No
4) Hexagonal Radial Reflectors 480 000 mrem /hr No (Removable & Permanant)
5) Large Side Reflector 312 <30 Rem /hr Yes l Blocks
6) Reflector Keys 24 100 mrem /hr No
(
7) Side Spacer Blocks (a) with Boron Rods 1,152 30 Rem /hr Yes (b) w/out Boron Rods 1,152 <3 Rem /hr Yes (c) Boron Rods 309.792 60 Rem /hr Yes
8) Bottom Reflector Blocks (a) with Hastelloy Cans 276 300 Rem /hr Yes l (b) w/out Hastelloy Cans -
276 500 mrem /hr No l
(c) Hastelloy Cans 20,061 10,000 Rem /hr Yes
9) Core Support Blocks 24 1,000 Rem /hr Yes Hastelloy Keys l 10) Core Support Blocks & Posts 244 15 mrem /hr No
11) Metal Clad Reflector Blocks 6 300 Rem /hr Yes (Non control rod) l l
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Attachment I to P-91118 April 26, 1991 Page 43 TABLE 5 DISPOSITION OF HIGHLY ACTIVATED CORE COMPONENTS Handling Stabilize Shipping Core Component Method in HSF Container
1) Top Reflector Blocks 1 LSA STD
2) Bottom Reflectors 1 LSA STD
3) Radial Reflectors Hex. 1 LSA STD Removable & Permanent
4) Large Side 1 Yes Type A or Cask Reflector Blocks
5) Side Spacer Blocks 1 or 2 a with Boron Rods, or Yes Type A or Cask b without Boron Rods Yes Type A or Cask c Boron Rods Yes Type A or Cask
6) Bottom Reflector Blocks 1 or 3 a) with Hastelloy Cans, or Yes Type A or Cask b) without Hastelloy Cans Yes LSA STO c) Hastelloy Cans Yes Type A or Cask
7) Core Support Blocks 1 Yes Type A or Cask Hastelloy Keys
8) MCRBs (Non-control rod) 1 Yes Type A or Cask
,-n---
Attachment I to P 91118 April 26, 1991 Page 44 N3C Ouestion No. 14 (Section 2.3.3.9 Core Barrel Removal):
"What thermal cutting methods and underwater operations will be used to remove the core barrel?"
PSC Responsgi Underwater and in-air cutting approaches will be evaluated during the Phase I design effort. Semi-remote and manual methods are being evaluated for each of these approaches. If a semi-remote underwater method is chosen, a mast mounted plasma torch will be used.
Semi remote in-air methods being considered include a track mounted oxy fuel cutting (acetylene) or a plasma torch, if a manual method is selected, only in-air cutting will be performed, in air techniques being evaluated include oxy fuel cutting (acetylene),
plasma torch, oxylance, thermitic rod, and carbon arc. Other factors that will be included in the selection of the cutting technique include cost benefit analyses, estimated radiation levels in the core barrel region when the cutting is performed, ALARA, and cutting speeds.
Regardless of the cutting technique selected, containment of radioactive contamination will be of primary importance. For either method identified above, fumes given off will be collected by installation of appropriate containment, exhaust fans, filters, and ducting in the work area.
4
l Attachment I to P 91118 !
April 26, 1991 1 Page 45 l NRC Ouestion No. 15 (Section 2.3.3.10.2 Removal of the Core Succort Floorh "A. This section discusses several methods for removal of the core support floort however, a removal method has not been selected.
PSC has not indicated what will be the controlling factor affecting the selection of the method of removal. The method of removal should be selected and analyzed. These factors need to be addressed in this sectfon.
B. How will the impact of selecting a method for removal of the '
core support floor be factored into the decommissioning schedule and cost? ,
l C. What methods will be used to remove large pieces of debris that fall from the PCRV head and core support floor during their l removal?
l D. What specific remotely operated cutting methods will be used l for cutting the stenin generators and core support columns from '
the core support floor? How wi11 the core support floor be ,
sectfoned and handled? How wil1 occupatlonal safety be ensured?
PSC Resoonse:
A. The removal of the Core Support Floor (CSF) is a difficult operation because of its size and weight. The CSF is a large disc approximately 29 feet in diameter by 5 feet thick and weighing 270 tons. The existing crane in the Reactor Building-only has a capacity of 170 tons and cannot lift the CSF in one piece. Therefore, the CSF will have to be cut into pieces within the PCRV if it is to be removed by the Reactor Building crane. This will require working in the radiation field that exists due to the CSF and the other components remaining in the PCRV and with limited access on the sides and no direct access to the bottom of the CSF. It is desirable to raise the CSF to the PCRV top head region for greater access and to- reduce the radiation effects of other PCRV components. This can be accomplished by using a " strand jacking" system, which uses multiple cables hooked to the CSF and attached to hydraulic jacks positioned on beams above the PCRV. The jacks would raise the CSF and place it on supports on the ledge in the cavity where the PCRV top head was cut and removed. Shielding and radiological containment would be installed as necessary, and the CSF would be cut into segments small enough for handling by the Reactor Building crane. This method will provide greater access for the segmenting operations, and the work can be accomplished in a lower radiation field. -
B. The fixed price contract includes removal and disposal of the Core Support Floor. If the contractor chooses to use an
Attachment 1 to P 91118 i April 26, 1991 Page 46 alternate method, there is no change in cost to PSC. Likewise, Ae decomissioning contractor is comitted to an overall completion schedule for the decomissioning activities.
Regardless of the method used for CSF removal, the ccntractor must complete the overall decomissioning on schedule or pay penalties to PSC.
C. Experience with diamond wire cutting during the D.C. Cook steam generator replacement program has shown that large pieces of debris were not generated by the cutting process. The diamond wire cutting process cuts the concrete leaving a solid block held together by rebar. However, if a piece did fall into the PCRV, it would be removed with appropriate hand held and/or crane held tools, grapples, cables or buckets as appropriate.
D. A semi remote in air cutting method will be used to cut the steam generator ducts from the core support floor. Techniques to be evaluated in the decommissioning Planning Phase are oxy fuel cutting (acetylene), oxylance, and plasms torch.
An in air method will also be used to remove the core support floor columns from the CSF. One technique to be evaluated during Phase I is to drill through the CSF with a 16 inch to 18-inch oversize core drill to free the columns from the floor.
An alternate technique to be evaluated is to use an oxylance to cut out a section of the steam generator ducts for access to the annulus between the CSF and the PCRV liner.
Stress ar,alysis of the core support floor columns will determine the number of columns required to support the CSF at this stage of dismantlement. A lifting strueure will hold the core support floor in place prior to cutting the remaining columns.
After the columns have been cut, the CSF will be lifted and supported incide the PCRV. After the dry cutting of the CSF columns and the steam generators, the water level in the PCRV will be raised to a level above the CSF. The CSF lifting process is expected to occur with the PCRV flooded to a level above the CSF The CSF will then be lifted into the PCRV tcp head region and cribbing will be lowered into place. The CSF will then be positioned on the cribbing while inside the PCRV top head area. The.outside of the steel liner pl .e will be scored and the diamond wire cutting method will be used to section the remainder of the floor while it is positioned on the cribbing in the PCRV top head cavity.
The expected exposure rate from the top surface of the CSF is expected to be approximately 400 mR/hr. Temporary shielding will be used as necessary for all in-air operations.
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Attachment I to P-91118 April 26. 1991 Page 48 HRC Ouestion No. 16 (Section 2.3.3.12 Final Dismantlina. Decontamination, and Clean-uo Activities)
"A. This section states that if the tendon tubes prove to be unsuitable for the diamond wire cuts, new vertical holes will have to be core drilled. PSC should include a discussion of the criteria that will be used to determine if the tendon tubes are acceptable.
B. PSC should also address the impact of core drilling new holes on the decomissioning schedule and cost. The discussion should address the additional exposure and spread of contamination resu! ting from the core drilling. C. How and where will the diamond wire cuts be made in the final PCRV dismantlement?" PSC Resconse: A. Diamond wire cutting through a tendon tube will be tested and evaluated on a concrete mock-up. The results of the mockup testing will provide the necessary criteria to determine tne suitability of the tendon tubes for sectioning the concrete. B. New holes (14 total at a length of 40 feet per hole 560 linear feet) would be drilled through non-activated concrete, therefore not risking the spread of contamination. Approximate drill locations for the new beltline vertical holes are shown in Figure 4. This operation would be performed after the graphite core elements have been removed from the reactor and radiation levels are minimal. This activity can be performed in parallel with work on the core support floor, steam generator removal and clean out of the lower plenum, therefore having minimal effect on schedule. The cost of this additional core drilling is estimated to be $50,000 to $75,000. Thi contingency is included in the Contractor's estimate. C. See response to question 7 A. 1
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Attachment I to P-91118 April 26 1991 Page 50 NRC Ouestion No.17 (Section 2.3.4.1 BOP Dismantlement)
"PSC should provide the specific methods that will be used for dismantlement of BOP piping and equipment (e.g., plasma arc cutting, 1 saws, shears, etc.). What types of packaging will be used for wastes from these operations?"
PSC Resronse: The specific methods used in the dismantlement of contaminated piping systems such as those at fort St. Vrain include those methods cited above and each method may have a potential use in the decommissioning of Fort St. Vrain. The actual method used for a specific application can be better understood by a discussion of the characteristics of the various methods or tools and the radiological controls associated with them.
1. Saws - There are a variety of saws commercially available which are effective for cutting pipe from small diameters up to 12 inches in diameter. The portable band saw is an extremely effective tool for cutting pipe less than three inches in diameter because it is light weight, hand held, and very portable. Some of the other saws are not as easily moved, require more complicated holding fixtures and sometimes require air or higher voltages to operate.
Another advantages of saws is the ease of containing the spread of cuntamination. The metal filings created in the sawing process are easily captured.
2. Thermal Cutting Processes -
Thermal cutting processes (most commonly either oxy / acetylene er plasma) are also very portable, rnd can cut an almost unlimited size range and variety of materials. The biggest draw-back for thermal cutting are the complications involved in containing the spread of contamination. With the use of appropriate containments and filtered exhaust systems, the spread of contamination can be controlled, however this concern limits the use of thermal cutting processes.
3. Shears - The use of shears is a very effective method to minimize the spread of contamination. This is offset however by the size of the equipment. For anything but small diameter pipe or tubing, the equipment required is very heavy and bulky and it can be very difficult to position the shear head at the desired location.
Accordingly, the use of shears is reserved for unique applications and may not be applicable to any of the systems at Fort St. Vrain. The selection of the proper tool will be made based on the particular task being performed in conjunction with the Radiation Work Permit and ALARA review to assure that the optimum balance
l Attachment I to P-91118 April 26, 1991 Page 51 between effective cutting and proper radiation control practices are achieved. Typically, on the removal of piping systems at Fort St. Vrain, radiation fields are expected to be low and therefore, the commitment to maintain exposures ALARA is not expected to be a major factor in the selection of cutting methods. The packaging of waste from the removal of contaminated piping systems will primarily be achieved by cutting pipe into approximate 7-foot lengths which will allow it to be packaged in LSA boxes (approximate dimensions of 4' x 8' x 3'). Other contaminated components associated with the piping systems can also be packaged in this size of container.
~ _ _ _ _ _ _ _ . _ _
Attachment 1 to P-91118 April 26, 1991 Page 52 NRC Ouestion No. 18 (Section 2.4,9 Decommissionino Safety Review Committee):
"The Decommissioning Safety Review Comit tee membership should include the PSC Radiation Protection Manager."
PSC Resoonse: In PSC's response to NRC Question No. 31(B), it is noted that the PSC Facility Support Manager will also function as and meet the requisite qualifications for the position of PSC Radiation Protection Manager, in this combined role, this individual is currently required to be a member of the Decommissioning Safety Review Committee. The Proposed Decommissioning Plan description of the Decommissioning Safety Review Committee will be revised to specifically require membership of the PSC Radiation Protection Manager and the PSC Decommissioning Organization Chart (Figure 2.4-1) will be revised to reflect the dual roles of the PSC Facility Support Manager, as well as the reporting chain of command in the role as the PSC Radiation Protection Manager in accordance with the attached figure.
i Attachment I to P-91118 April 26, 1991 Page 53 FIGURE 5 DECOMMISSIONING ORGANIZATION CHART PRESIDENT PUBLIC SERV!CE COMPANY OF COLORADO VICE PRESIDENT NUCLEAR OPERATICNS DEC04MI5510NING SAFETY REVIEV COMMITTEE
- WESTINGHOUSE , PROGRAM MANAGER DECOMMIS$10NING PROJECT DIRECTOR PROJECT ASSURANCE OPERATIONS ENGINEERING FACILITY $UPPORT MANAGER MANAGER MANAGER MANAGER DECOMM15$10NING RADIATION PROTECTION MANAGER l
Attachment I to P-91118 April 26, 1991 Page 54 NRC Ouestion No. 19 (Section 2.6 Trainina Proaram)1 "The training requirements for radiation protection staff have not been identified. This section should include a description of initial training, qualification, and retraining requirements, and any special training required in relationship to implementation of the revised 10 CFR 20." PSC ResDonse: Section 2.6 of the Proposed Decommissioning Plan will be revised to identify the training requirements of the radiation protection staff. These revised training and qualification requirements will be consistent with the guidance contained in NUREG 0761 (Draft, 1981) regarding recommended requirements for radiation protection training and qualification. This revision will identify initial training and qualification requirements, identify those individuals or groups that require ou lification, frequency of retraining or requalification, train;ag documentation, requirements for contractor personnel, and special training requirements for the following radiation protection staff: (1) General employee training (2) Radiation workers (3) Respiratory protection (4) Radiation protection technicians (5) Specialized technician training (e.g., dosimetry, respiratory protection, bioassay, counting room, environmental monitoring) (6) Radiation protection supervisor / foreman (7) Radiation Protection Manager (8) Emergency response personnel The radiation protection staff will include personnel with prior experience and training in decom:nissioning, operational radiation protection and radwaste. All radiation protection technicians will be required to successfully complete initial classroom and on-the-job-training. Other radiation protection staff members will be provided applicable training, either formal or on-the-job, as necessary, to ensure that they can perform their assigned job functions in a safe and competent manner.
Attachment I to P-91118 April 26, 1991 Page 55 NRC Ouestion No. 20 (Section 2.3.6 Occupational Exoosure Estimate):
"This section should include a descriptton of the basic assumpttons used in estimating exposures for each identifled task in sections 2.3.3 and 2.3.4. This section indicates that the radiation levels were not based on actual measurements. The section should also include a discussion of how PSC will evaluate and factor in actual 1 measured radiation levels against the initial assumptions." \
l PSC Response: The estimate of Occupational Radiation Exposures (ORE) for decommissioning of Fort St. Vrain was based on historical survey data for the balance of plant (B0P) systems external to the PCRV. General area radiatien levels for 80P systems measure less than 1 mR/hr. Therefore, the removal of contaminated 80P components and decontamination of remaining structures is not expected to contribute significantly to worker exposures. The radioactive fuel currently remains in the PCRV. Consequently, the only method for determining radiation levels of components internal to the PCRV is by activation and plateout analysis. The results of these calculations indicated that exposure rates on contact (1 cm) from component surfaces ranged from less than 1 mR/hr for carbon steel reflector keys to 10 R/hr for large permanent graphite reflector blocks to 10,000 R/hr for activated Hastelloy metal cans. Based on these analyses, the water shield used for flooding the PCRV and shielded transport containers will be used extensively to minimize worker exposures during decommissioning operations. ~ The assumptions identified in the following paragraphs will be incorporated into the detailed cost estimate currently being prepared for submittal to the NRC in May 1991. Assumotions for Exoosure Estimates The assumptions and methods for calculating the occupational radiation exposures consisted of: (1) Utilizing the calculated exposure rates for reactor internal components, (2) utilizing projected times for completing each task where the potential for radiation exposures exists, and (3) utilizing engineering experience gained from similar projects in operating nuclear plants. The assumptions used in the calculations were as follows:
1. For PCRV operations only, the time workers spend in the work station radiation environment was assumed to be 50% of the time scheduled to complete the task.

2. The " crew averaged" radiation levels were determined by assuming the total exposures estimated for completing a task would be uniformly distributed among the crew.
Attachment I to P-91118 April 26, 1991 Page 56 j
3. The graphite reflector blocks will be removed without the use of shielded transfer casks. (Shadow shields, distance and long handle tools will be used while removing highly radioactive components.)

4. The activated boronated pins will be loaded into shielded containers under water and transferred to the Hot Service Facility for processing.

5. The water level in the PCRV will be maintained such that the general area exposure rate on the work platform will be less than 2.0 mR/hr.
l
6. The workers will be trained in ALARA principles and good I work practices to minimize occupational exposures. ;
Dose Estimates The estimated occupational ~ radiation exposure for each major activity where the potential for worker exposure exists is given in Table 2.3-2 of the Proposed Decommissioning Plan. The total 3 cumulative occupational radiation exposure for the entire decommissioning project is estimated to be .433 person-Rem, due almost entirely to PCRV dismantlement and associated waste handling i activities.
]
The exposures for completing tasks inside the PCRV were determined from discussions with craft personnel based on the number of workers involved, the location of workers, and duration of tasks in proximity to the high radiation fields. In general, the radiation fields are expected to vary significantly for workers at different ,
-locations- in- the work station area. Consequently, the " crew-averaged" radiation fields were determined by distributing the total estimated exposure over the entire work station crew.
ALARA Work practices In order to complete each individual task and keep personnel exposures to a minimum, the following work practices will be implemented:
1) Pre-job briefings will be held with craft and radiation protection personnel to assure that ALARA practices have been adequately' factored into the work packages for completing tasks.'

2) Personnel exposures will be monitored on a regular basis for potentially high exposure tasks to identify any irregularities that may indicate excessive personnel exposures. In the event that an irregularity is found it will be investigated immediately and corrective actions implemented.
Attachment I to P-91118 April 26, 1991 Page 57
3) Tag lines will be attached and used when rigging and lifting high exposure rate components (e.g. steam generators) from the PCRV to keep workers as far from the source as possible.

4) Only essential personnel will be allowed inside the Reactor Building when high exposure rate components are being removed. Casual observers will not be permitted.

5) Bagging techniques will be designed for quick installation.

6) long-handled wipe tools will be used when appropriate to wipe down wet components removed from the PCRV.

7) Shadow shields (lead blanket curtains or equivalent) will be used, where appropriate, to reduce radiation fields at the work stations.
Exoosure Rate Measurements The radiation levels from activated structures inside the PCRV will be measured as segments of the reactor are dismantled. The measurement of exposure rates at the individual work stations will be performed and compared with calculations prior to commencing each individual task. Adjustments will be made in exposure projections as necessary. Temporary shielding will be used at the work stations to minimize personnel exposures based on actual exposure rate measurements.
Attachment 1 to P-91118 April 26, 1991 Page 58 NRC Ouestion No. 21 (Section 2.5.2 Scooe of Work):
"The decomissioning plan states that the decomissioning project will be conducted in two phases. PSC should explain the nature of these two phases."
PSC Resoonse: The two phases of the Fort St. Vrain Decommissioning Project are identified in Section 1.2.5: Phase i Decommissionina Plannina Phase, with an estimated duration of 18 months, consists of initial site characterization, preparation of work scope planning, work specifications and procedures, and equipment and material staging. There will be N0 physical decommissioning activities performed as part of this planning phase, although some component removal and disposal activities may occur prior to commencement of Phase 11 (described below) as described in Section 1.5 of the Proposed Decommissioning Plan. Phase II Decontamination and Dismantlement Phm with an estimated duration of 39 months. Actual dismantlement, decontamination, and physical decommissioning activities will occur as part of this phase. The actual physical decommissioning activities are scheduled to commence after: (1) NRC aoproval of the Proposed Decommissioning Plan, and (2) removal of all irradiated fuel from the Reactor building. It is important to note that Phase I and Phase 11 activities are not conducted in series. These two phases have considerable overlap (see PSC response to NRC Question No. 2). Further detailed descriptions of the work scope to be performed in each project phase are provided in Appendix 1. Additionally, Section 2.3.1 of the Proposed Decommissioning Plan will be revised to distinguish the two project phases.
Attachment I to P-91118 April 26, 1991 Page 59 NRC Ouestion No. 22 (Sections 2.5.3.2 and 2.5.3.3 Manaaement Resoonsibilities):
" Waste processing is listed as a responsibility of both the Technical Services Manager and the Operations Manager. Is this correct? How are these responsibilities broken down?"
PSC Response: The Westinghouse Technical Services Manager is responsible for the technical aspects of waste processing. This involves specifying requirements, methods, regulations, and procedures, The Westinghouse Operations Manager is responsible to accomplish the activities in the field, in accordance with the specified requirements. These include assuring craft labor parforms the day-to-day activities of decommissioning, decontamination, waste handling (hence, waste processing) and site maintenance. These activities include placing waste in proper containers, operation of the water processing system, waste segregation, and on-site waste minimization. l
Attachment 1 to P-91118 April 26, 1991 Page 60 NRC Ouestion No. 23 (Table 2.5-1 Westinahouse Team Exoerience):
"This table shows that Westinghouse has experience in removing reactor vessels from the West Valley Fuel Reprocessing Plant. What reactor vessels were removed at this facility? What was their size, weight and material?"
PSC Resoonse: Westinghouse has dismantled approximately 50 stainless steel chemical processing vessels at West Valley. These vessels were part of the chemical reclamation process used by the former owner. All vessels were remotely handled, packed and stored on site and are waiting final disposal. The average capacity of the vessels were about 1,000 gallons. The largest was the fuel dissolver vessel that measured 8 feet in diameter, 20 feet tall, and weighed approximately 13 tons. The capacity of that vessel was about 5,000 gallons. No reactor vessels were dismantled by Westinghouse at the West Valley Fuel Processing Plant. The specific entry in Table 2.5-1 of the Proposed Decommissioning Plan will be deleted.
l Attachment 1 to P-91118 April 26, 1991 Page 61 NRC Ouestion No. 24 (Section 3.1.1 Facility Ooeratina History):
"During the operating life of FSV has any disposal been conducted at the site pursuant to 10 CFR 20.304 or 10 CFR 20.302? If so, provide a description of waste buried, including location, concentrations, radionuclides involved, curie' quantities, and waste volumes."
PSC Resoonse: PSC has not conducted any disposal operations at the site pursuant to 10 CFR 20.304 or 10 CFR 20.302.
Attachment I to P-91118 April 26, 1991 Page 62 NRC_0uestion No. 25 (Section 3.1.2.1 Description of Instrumentation and Survey Technioues): "A. PSC stated in this section that natural background is in the range 0.004 to 0.032 mrem /hr. Please provide full details of how the values of background radiation levels and their ranges are to be used in connection with release criteria and survey procedures. B. The section states that alpha contamination is not present above natural background levels at FSV. Please provide in detail the basis to suppor'; this statement. Provide survey results supporting the state.nent about alpha contamination. C. Was the August 1990 smear servey analyzed for beta and alpha acttvity or only beta? This should be clarifled and the data should be provided." PSC Resoonse: A. The background range of 0.004 to 0.032 mrem /hr stated in this section has been validated as part of the background determination in the initial site characterization program, utilizing applicable guidance from draft NUREG 2082 methodology. The background levels determined will be integrated into the implementing procedures that will specify the detailed surveys for release of materials, equipment or areas for unrestricted use. B. PSC has been performing routine alpha surveys at Fort St. Vrain since the early pre-operational stages of the plant. During ' this period, there have been no measurable indications of alpha radiation or contamination as a result of any plant activity. PSC has available in the records storage facility approximately 18 years of survey and sample analysis records that document these results. C. Page 3.1-2, paragraph 4 of the Proposed Decommissioning Plan states that these smears were counted for beta activity only. The data is provided in Figures 3.1-1 through 3.1-19. I
i Attachment I to P 91118 April 26, 1991 Page 63 NRC Ouestion No. 26 (Section 3.1.2.2 Turbine Buildina Survey Results):
"This section states that in all cases the radiation levels are less than 1000 dpm/200 cm2 PSC should provide the basis that supports these readings. Did the August 1990 survey provide the basis for this information?"
PSC Resoonsel As noted in Figures 3.1-3 through 3.1-7, contamination levels in the Turbine Building are all less than 1000 dpm/100 cm2 This conclusion was confirmed by the August 1990 survey performed in support of the Proposed Decommissioning Plan, i
Attachment I to P-91118 April 26, 1991 Page 64 NRC Ouestion No. 27 (Section 3.1.3: Current Environmental Radiolooical Status)1 "Will the current Radiological Environmental Monitoring Program (REMP) be continued in whole or in part during the decommissioning phase to monitor any increase in off-site environmental radiation levels due to decomissioning activities? If so, will the final results of the REMP be included in the final radiological survey report? If not, how will the potential radiological environmental impact be assessed?" PSC Resoonse: The current REMP will be continued in part specifically tailored to accurately monitor the environmental radiation and radioactivity levels, and to determine the effect on the radiological conditions of the environment due to decommissioning activities. The results of the REMP will be included in the final radiological survey report.
Attachment I to P-91118 April 26, 1991 Page 65 NRC Ouestion No. 28 (Section 3.1.4.1.1 Comouter Codes):
"This section should include a description of the computer codes used, a description of the code verification and validation, the assumptions used, material specification data applied, and any additional applicable documentation as well as references. Detailed information regarding the codes and their use should be clearly referenced."
PSC Response: The four main codes and data libraries used in the activation analysis were: BUGLE-80, COMMAND, ANISN and REBATE.
1. BUGLE-80 BUGLE-80 is a cross section data library maintained by the Radiation Shielding Information Center (RSIC) at Oak Ridge National Laboratory (ORNL). BUGLE-80 is a coupled 47 neutron, 20 gamma ray, P3, cross section library for radiation transport analysis for neutron, gamma-ray or coupled neutron-gamma problems.

2. COMMAND COMMAND is a cross section library processing module in the AMPX-Il modular system from RSIC at ORNL, COMMAND module is a utility code for collapsing cross section sets.

3. ANISH ANISN is a multigroup one-dimensional discrete ordinates code that solves the energy dependent, one-dimensional Boltzmann transport equation with general anisotropic scattering for
- slab, cylindrical or spherical geometries. ANISN is maintained by the RSIC at ORNL.
4. REBATE REBATE is an activation code for the calculation of radioactivity and gamma ray source strength for one or two dimensional gamma transport analysis. REBATE was obtained from the Princeton Plasma Physics Laboratory.
Additional information on the above codes can be found in the selected sections of Appendix A. Volume 1, of PSC's Activation Analysis provided as Attachment 2 of this submittal. The first step in the activation analysis was to determine the neutron flux throughout the internal components and the PCRV. The ANISN code was used to calculate the neutron flux. Actual neutron flux data, taken from the opet atinr.al data from Fort St. Vrain, were used to set the proper neutron source input to ANISN. Additionally, cross section data, taken from the BVGLE-30 data library, was used as input to the ANISN code. The BUGLE-80 cross section data library was collapsed to 16 group cross secticns, using the COMMAND code
)
a Attachment 1-to P 91118 April 26; 1991-Page 66 prior to use as input to the ANISN code. The ANISN code used P3 scattering and the quadrature of the transport solution was S8. The second step in the analysis was to determine the activation of selected components in the PCRV due to the neutron flux generated in the first' step of the analysis. This was accomplished using the REBATE code and some data handling routines. Material compositions, previously calculated neutron fluxes, component geometry and power history of Fort St. Vrain were used as inputs to - the REBATE code. The REBATE code also calculated the gamma source (ll-group) generated by each PCRV component. The final step in the analysis was to calculate the exposure rates inside the. PCRV. The 11 group gamma source calculated in the previous step was used as input to the ANISN code for the determination of the spatial gamma ray flux resulting from the radioactive decay of the activation products in the various components left within the confines of the PCRV. Verification of the-accuracy of PSC's activation analysis, performed by the codes described above, was achieved by actual _ measurements of: (1) the' activation levels of Charpy V-notch specimens of PCRV liner material, that were irradiated over the life of the core in the PCRV top head wells adjacent to the top-head liner, and (2) the activation levels of _ PCRV prestressing tendon wires removed from vertical . tendons near the inner barrel. section of the PCRV at the core elevation. These tendon wires were also irradiated over the entire- life of the core. The results of these analyses were recently- submittc to the NRC in PSC letter (Crawford to Weiss)_ dated April _15, lw91 (P-91137), and indicated excellent agreement
~
between actual measured activ,ation levels and those activation . levels predicted by PSC's activation analysis. Additiunal details of the code inputs, models, and assumptions can
..be found in selected sections of Appendix A, Volume 2, (attached) of the PSC Activation Analysis.
Attachment I to P-91118 April 26, 1991 Page 67 tiRG Ouestion No. 29 (Section 3.1.4.1.7 and Accendix II Activation Analysis Results): "A. NUREG/CR-3474 presents the results of an assessment of the potential problems posed to light water reactor (LWR) decommisstoning by long 1ived acttvatton products produced in the major construction materials for the LWR. PSC has applied these activation results to FSV (HTGR). Provide the basis that suppu.ts the use of the LWR activation analysis and include in the discussion an estimate of the conservatism, if any, resulting from applying the LWR analysis to the FSV HTGR. B. NUREG/CR-3474 references a two part study completed by Woolam for British Magnox reactors. These reactors are gas cooled graphite reactors. Were the results of this study evaluated and compared to NUREG/CR-3474 and/or FSV? C. The material composition data for the activation analysis was based on data taken from NUREG/CR-3474 and not on an elemental composition from material certifications. What is the effect of using the highest elemental concentrations of Co, Ni and Nb rather than average values such as that used for Niobium? D. The activation analysis should include a detailed discussion of how actual measured activation levels will be incorporated into the activation analysis. The discussion should address the estimated number of samples pSC intends to take, the location of these samples, and when the samples will be taken. E. What is the effect of using the highest elemental concentrations of Co, Nb, Ni and Eu on the waste ' classifications?" 11C Resoonse: A. The specific results of NUREG/CR-3474 were not applied to the Fort St. Vrain HTGR. An independent activation analysis unique to Fort St. Vrain was performed (see EE-DEC-0010, Appendix B to the Proposed Decommissioning Pl an) . Neutron transport calculations were performed to determine the flux level in the reactor and separate activation calculations were performed to estimate the specific activity of the PCRV and its internal components. NUREG/CR-3474 was evaluated to ensure that all nuclides of importance to decommissioning found in the NUREG were considered in the Fort St. Vrain activation analysis. Data from NUREG/CR-3474 was used for estimates of the average concentrations of trace elements in the PCRV concrete. The average trace element concentrations at Fort St. Vrain in NUREG/CR-3474 were obtained from samples from bioshield concrete from twelve different nuclear plants. It was assumed that the average trace element concentrations would not vary I
Attachment I to P-91118 April 26, 1991 Page 68 significantly from these values. Sensitivity studies were performed (see Response to NRC Question 29.C.) to determine the effect of using higher than average concentrations of isotopes of concern. At the time the activation analysis was performed, no sample data from actual PCRV concrete was available. PSC is planning to take concrete surface samples to test for trace element concentrations (see Response to Question 29.D.). Compositions of other materials were taken from material specifications of the each component. Trace element assumptions for these materials, especially cobalt, were compared to those in NUREG/CR-3474, when applicable. B. The actual results of the study, in terms of site specific activity levels and exposure rate assessments are not directly comparable to the Fort St. Vrain analysis since those items are greatly influenced by site specific parameters such as operating history and reactor core specific flux levels. The most valuable information included in the study is the data presented on the identification of important isotopes, normally found as trace elements in components, to decommissioning / dismantlement and waste disposal. Part I of the study was reviewed on this basis and compared with similar data found in NUREG/CR-3474 and with the assumptions made in the Fort St.- ) Vrain activation analysis. A discussion of the comparison is I provided below. Part I of the study, "Heasurements of Neutron Induced Activity in Samples from the Reactor Island," summarized the experimental data gathered for the determination of material composition, including trace elements, of mild steel, concrete, graphite and Magnox Fuel Cladding. The report also addressed the potential contributions to the exposure and disposal problems for the Magnox reactor. Part 11 of the British study, "A Summary of Neutron Induced Activation, Waste Disposal, and Dose Equivalent R;ses for the i Island Structure," summarizes the calculational me'.hodology and
results of the calculations. The study concluded that a combined theoretical and experimental assessment is the only satisfactory means of obtaining an accurate radioactive inventory of the reactor structure. Calculations could not be relied upon alone due to the difficulty in determining the neutron flux in complex geometries and uncertainties in unknown trace elements. Experimental data by itself is also

not sufficient because of the limited number of samples for which it is practicable.
The conclusions drawn from the British study and NUREG/CR-3474 l are consistent with those used in the Fort St. Vrain activation analysis. In mild steels Co-60 was found to be the dominating l l l
Attachment I to P-91118 April 26, 1991 Page 69 ' gamma emitting isotope. Tha British study found mild steel samples to contain 130 ppm cobalt. NUREG/CR-3474 found cobalt in rebar and vessel steels to range from 80 to 151 ppm. The Fort St. Vrain activation analysis used a conservative valte of 200 ppm for carbon steels. Samples taken from the British reactor found that the most important isotopes in the concrete were europium and cobalt, with calcium of concern in terms of waste disposal. These results were consistent with those found in NUREG/CR-3474 and with the results of the sensitivity studies performed for Fort St. Vrain (see response to question 29.C.). All studies indicated that Co-60 would be the dominant gamma-emitting isotope in the short term and that Eu-152 would dominate in the long term. Comparisons between the British study and the Fort St. Vrain activation analysis (the activation analysis used NUREG/CR-3474 for average trace element assumptions for concrete) are as follows: British Study FSV / NUREG/CR-3474 Co ppm 3.0 9.8 Eu ppm 0.2 0.55 Ca % 8.0 12.4 (from FSV FSAR) The British study also obtained experimental data from graphite. The specific type of graphite sampled in the study was unknown, but the important gamma emitting isotope to dose was determined to be Co-60. In the British study the cobalt abundancy in the graphite was found to be 0.01 ppm. The Fort St. Vrain study assumed an abundancy of 0.01% of the fe content. PSC has completed the theoretical study for the activity in the reactor structure and is in the process of gathering experimental data to aid in eliminating uncertainties in the trace element compositions (see Response to Question 29.D.). C. The only material composition data extracted from NUREG/CR-3474 and used in the Fort St. Vrain activation analysis was for the PCRV concrete trace elements. Other component material concentrations were determined from standard and/or specification of the component. The sensitivity of the amount of concrete required for removal and the trace element assumptions used in the activation analyses was studied using the available trace element data in NUREG CR/3474. The activation analysis assumed that the trace element constituents in concrete were at the average trace element impurity level found in samples from the bioshield concrete of twelve different nuclear plants. In order to assess the effect of varying the trace element concentrations
1 Attachment I to P-91118 April 26; 1991 , Page 70 ' ! to the amount of concrete requiring- removal, a worst case scenario was . considered. The elements whose activation products are most important to exposure rate were assumed to be at the maximum values given in NUREG CR/3474 for h21h the rebar and the concrete. The exception to this was the cobalt level in the rebar, which was conservatively set at 200 ppm. The elements considered to be most important, in the short and/or long term were cobalt (Co 60), niobium (Nb-94), silver (Ag-108) i and europium (Eu-152), where the isotope in parenthesis i indicates the isotopv of concern (Nickel was also included in l the sensitivity study). The increase in number densities of I the above elements varied from a factor of 1.5 to 2.2. The i results of the study are as follows: Exoosure rate 5 years after shutdown (microR/hr): Max Trace Previous Calculation Elements Trace Elements-24" Concrete 6.0 3.4 Removed 26" Concrete. 3.2 n/a Removed Exoosure rate 60 years after shutdown-(microR/hr): Max Trace Previous Calculation Elements Trace Elements
'8" Concrete 7.4 3.5 Removed 10" Concrete 3.7 1.7 Removed The above results indicate that the effect of using the-maximum value for all trace elements results in an increase of the required depth of concrete to be removed of approximately two inches in both the long and short term. In the short term, the dominant gamma emitting isotope is Co-60 both from the rebar and concrete. The increase in the cobalt number density was approximately 1.66 and the corresponding increase in - exposure rate at 5 year after shutdown was 1.76. In the long term Eu-152-is the dominate-isotope. The increase in the europium number-density was 2.2 and the corresponding increase in exposure rate at 60 years after shutdown is 2.11.
The results indicate that the increased depth of required concrete to be removed is directly related to the increase in the levels of cobalt (short term) and europium (long term). It is unlikely that the cobalt and europium concentrations in PCRV 7**-* --m-- - r--- -m -
1 Attachment I to P 91118 l April 26, 1991 ' Page 71 concrete are widely different than those sampled in NUREG/CR-3474. Surface samples of unirradiated concrete are ; currently being taken and analyzed for trace element j composition (see response to Question 29.D.). ; D. Response to this question has been provided in PSC letter j (Crawford to Weiss) dated April 15, 1991 (P-91137). j l E. The waste classification of the rebar/ concrete will not change j due to increasing the elemental concentrations of Co, Nb, Ni a and Eu to the highest concentrations given in NUREG/CR-3474. ] The following tables contain activity information for the most highly activated (predicted to be in the top head) concrete at three years after shutdown. The isotopic limits of 10 CFR 61 are included for comparison. The conservative assumption that all Nb-94 and Ni-59 are located in the rebar was made since no limit for these isotopes exists in nonmetal waste. Similarly, all the C-14 and Ni-63 are assumed to be in the concrete, since the limit for nonmetals is more conservative than for metals. The information on the attached tables demonstrates that the predicted isotopic specific activities are well below the allowable limits for class A waste, the most limiting isotopes being two orders of magnitude less than the limits. The trace element sensitivity study (see response to Question 29.C.) showed that the variations in elemental number densities did not vary by this amount for Co, Nb, Ni and Eu. Therefore, the waste classification for the concrete would not be affected by increasing the average elemental concentrations to the highest elemental concentrations. , .
Attachment ! to P-91118 April 26, 1991 Page 72 TABLE 6 PREDICTED 10CFR61 Limits WASTE IS0 TOPE CLASS A CLASS C ACTIVITY QM111 C-14 0.8 8.0 1.14E-04 uC1/cc Ni-59 22.0 220.0 8.64E 07 uCi/cc Nb 94 0.02 0.2 8.53E-07 uCi/cc I TABLE 7 PREDICTED 10CFR61 Limits WASTE ISOTOPE CLASS A CLASS B CLASS C ACTIVITY QM111
< 5 yr 700.0 6.40E+00 uti/cc H-3 40.0 1.97E-01 uti/c:
Co-60 700.0 2.34E-01 uti/cc Ni-63 3.5 70.0 700.0 1.18E-04 uCi/cc Sr-90 0.04 150.0 7000.0 7.63E-12 uCi/cc
Attachment I to P-91118 April 26, 1991 Page 73 NRC Ouestion No. 30 (Section 3.1.5 Initial Site Characterization Plan): "This plan indicates that an initial site characterization will be performed; however, the DP did not indicate when this characterization information will be provided to the NRC. Also, include a discussion of the basis for selection of the location for samples and how this information will be factored into the initial analysis." PSC Response: The Initial Radiological Site Characterization Program is being performed to identify the contamination status of all areas of the site, within and outside the buildings. This primary purpose of the site characterization is to define the initial condition of the site before decommissioning work begins. Westinghouse is responsible for the remediation of contamination, and therefore it is important that the initial site conditions be established. This Initial Radiological Site Characterization Program will also be used to identify sampling points, establish the data base for the comprehensive survey of the site for release, and assist in the development of the decontamination program. The information will be provided to the NRC with the final report for site release. This information will be available for review at any time during the decommissioning program. In the meeting between PSC and the NRC held on February 27, 1991, PSC committed to provide the NRC with a copy of the Initial Radiological Site Characterization Program. This program is currently scheduled to be submitted by separate correspondence by May 15, 1991. l
Attachment I to P-91118 April 26, 1991 Page 74-NRC Ouestion No. 31 (Section 3.2 Radiation Protection Procram):-
"A. Will PSC effect an early implementation of the revised 10 CFR Part 20 prior to the required date of January 1, 19937 This section should discuss how the transition will be accomplished and the schedule for implementation. The decommissioning schedule impact and cost of implementation should be discussed.
B. This section does not provide sufficient detail to evaluate the administrative organization of the radiatton protectfon program in accordance with the criteria suggested in Draft Regulatory Guide DG-1005 or Regulatory GJide 8.8. Draft NUREG-0761,
" Radiation Protection Plans for Nuclear Power Reactor Licensees," provides the format and guidance for a radiation protection program description. The Radiation Protectton Program should address the specific radiation protection aspects of the dismantling operations that involve highly radioactive components."
PSC Response: A. The decommissioning of Fort St. Vrain will be accomplished during transition to the new 10 CFR 20 regulations and continue beyond that time. For this reason, PSC plans to perform the decommissioning under the new rule. The general approach to this activity is:
1) Perform an in-depth assessment on the impact of the 10 CFR 20 revision on Fort St. Vrain decommissioning.

2) Develop, prepare and implement a program that integrates-the revised -sections of 10 CFR 20 into the Radiation Protection Program that will be used by PSC/W for the decommissioning project.
The proposed work will be approached through a multi-phase task basis. This will assist in identifying specific areas . within the Radiation Protection Program and associated programs which will be impacted by the forthcoming revision. The assessment-will differentiate between areas where modifications are-necessary and those that are not in need of revision. Specific information compiled during the initial phase will provide the basis for the most cost effective and timely revision of the
~
Radiation Protection Program in subsequent phases with respect to the new 10 CFR 20-requirements. An overview of the proposed phases of the program to implement the new requirements of 10 CFR 20 is provided in the following paragraphs:
Attachment I to P 91118 April 26; 1991 Page 75 Assessment Phase This phase addresses the identification of changes necessary to revise the Radiation Protection Program in accordance with the new 10 CFR 20 provisions. In this initial phase, an assessment will be performed to determine the impact on areas where revisions are necessary. The major changes will primarily occur in the aress of internal dose assessment, exposure limits, exposures to the public, and records and reports. The impact analysis will address the following: Magnitude of program changes; Capital equipment / engineering controls necessary to implement the changes;
- Extent of procedure revisions and/or new procedure development; Training essential to implement and maintain the revised programs; - Staffing support necessary to implement and maintain the revised program; - Sof tware revisions necessary for dosimetry, bioassay and air monitoring programs; - Cost estimates to implement and maintain the revised program.
The Assessment Pha+e results will form the basis for planning and deve? ....g; the subsequent phases. Plan and Schedule Precav ation Phase This phase will use the Assessment Phase results to create a plan to integrate the identified revisions into a schedule. This schedule will be consistent with the implementation goal of the actual start date for the onsite decommissioning activities. Revised Proaram Develooment Phase During this phase, the elements identified for change, upgrade, or new development will be prepared. Any capital equipment needed for program maintenance for engineering controls will be identified for procurement. Trainina Scoce Preoaration Phase This phase will develop, revise and implement a formal training program addressing the Radiation Protection Program requirements for the Program Implementation Phase. A schedule to conduct such training will be included. This phase will al so incorporate any changes necessitated by USNRC Regulatory Guidance issuance subsequent to and that is different from previously prepared elements.
Attachment I to P 91118 April 26, 1991 Page 76 Prooram implementition Phast Thii phase implements the revised program and properly traint. the staff on the implementation and ongoing maintenance requirements for the revised program. Post 1molementation Assessment Phase As regulatory guidance and transition documents become available, the Radiation Protection Program will be i assessed agahet these documents for compliance and possible program adjustment. When the issuance of the l regulatory documents is near completion, PSC will review i the new program to ensure quality of the final program. l The Assessment Phase report will detail all pertinent ! information accumulated during the initial assessment. The Plan and Schedule Preparation Phasa begins after the final issuance of the report. The Assessment Phase findings will be used to estimate the Plan and Schedule Preparation Phase costs. This will be the mechanism used to proceed in each sub:equent phase. The Assessment Phase evaluation requirements are fairly predictable with respect to costs. However, the total impact and implementation of a newly mandated program that does not have regulatory guidance issued at this point (aad may not for another year) is not predictable as to cost at this time. The new procedures will be prepared bned on the most current rerjulatory guidance available and implemented prior to the commencement of actual decommissioning work, or not later than January 1, 1992. Following implementation, the Radiation Protection Program will be assessed against any newly issued . regulatory guidance and modified, if necessary. B. A revised Radiation Protection Program will be provided in the proposed revision to the Proposed Decommissioning Plan, to be submitted in accordance with the schedule provided in the cover letter. Commitments made in response to the NRC questions contained in this RAI will be incorpo.ated in the revision to Section 3.2 of the Proposed Decommissioning Plan. The Radiation Protection Program will control the dismantling operations that involve highly radioactive components through a variety of methods, including:
1. ALARA Reviews of Work Packages

2. Shielding and Engineering Controls

3. Radiation Work Permits

4. Pre-job briefings and mock-ups

5. Controlling access to high-exposure areas

6. Monitoring by Radiation Protection Technicians

7. Iracking of exposure accumulation
.. *I. ... Attachment I to P-91118 April 26, 1991 Page 77 1he PSC Program Manager for Decommissioning and the Westinghouse Project Director have ultimate responsibility for assuring that an effective Radiation Protection Program is implemented during the fort St. Vrain decommissioning. Reporting directly to the PSC Program Manager for Decommissioning will be the Facility Support Manager. This individual will be responsible for the oversight of the Radiation Protection Program and will function as the fort St. Vrain Radiation Protection Manager. The facility Support Manager represents the formal line of communication and authority between fort St. Vrain management and the Westinghouse organization for radiation protection matters related to the def.ommissioning. The f acility Support Manager will be responsible for ensuring that the Radiation Protection Program and procedures meet the goals and standards established by fort St. Vrain management and the governing regulatory agencies. This individual will meet the qualifications contained in Regulatory Guide 1.8,
" Personnel Selection and Training."
Reporting directly to the Westinghouse Project Director will be the Westinghouse Radiation Protection Manager. This individual has responsibilities for the administration of the fort St. Vrain Decommissioning Radiation Protection Program. The Westinghouse Radiation Protection Manager will meet the qualifications contained in Regulatory Guide 1.8. The f acility Support Manager and the Westinghouse Radiation Protection Manager will formally interface via the Project
ALARA Committee which they will co chair. The facility Support Manager will also be a member of the Decommissioning Safety Review Committee. The facility Support Manager will be responsibb fnr the approval of the Project Radiation Protection Program and the content of radiation protection training programs. The Facility Support Manager will also be directly responsible for the Radiological Environmental Monitoring Program.
The responsibilities of the Westinghouse Radiation Protection Manager include:
1. The implementation and maintenance of an effective radiation protection program as required f or the fort St. Vrain decommissioning.

2. Establishing and maintaining an effective ALARA Program.
Attachment I to P-91118 April 26, 1991 Page 78
3. Establishing and maintaining a program that minimizes the volume of radioactive waste and that ensures safe transportation of radioactive material.
The Westinghouse Radiation Protection Manager will be responsible for the approval of the content of the radiation protection training programs, for the selection and approval of all radiation protection staff members, and for the review and approval of all radiation protection procedures. The fort St. Vrain Decommissioning Radiation Protection Organization will be managed by the Westinghouse Radiation Protection Manager and will be comprised of the following seven functional areas:
1. Radioactive Waste

2. Dosimetry

3. ALARA

4. Respiratory Protection

5. Health Physics Surveillance

6. Radiological Engineering

7. Instrumentation These functional areas will be assigned to one or more supervisors who will report to the Westinghouse Radiation Protection Manager. Personnel assigned to these positions will meet the qualifications specified in ANSI N3.1-1981. The supervisors will be responsible for implementing the program elements and supervising techniciais who are assigned to each area.
.-_--__._-___--__mm_ _ _ , , , _
Attachment I to P 91113 April 26, 1994 Page 79 NRC question No. 32 (Sectlan 3.2.1.1 Responsibility):
"This sectton should describe the qualificatfons for the PSC Radiation Protection Manager (RPM) consistent with the applicable guidance in NRC Regulatory Guide 1.8 and ANSI /ANS 3.1 1981. ;
functional responsibilities should be specified, as appropriate, consistent with the suggested guidance in NRC Regulatory Guide 8.8 , This section should also establish that the and Draft NUREG 0761. RPM will have direct access to upper management to ensure radiation protection implementation problems can be escalated if necessary to upper management attention independent of the project organization elements responsible for cost and schedule decisions " PSC Response: Th3 Proposed Decommissioning Plan will be revised to specify that the PSC and Westinghouse Team Radiation Protection Managers will be required to meet (as a minimum) the qualification requirements of NRC Reg. Guide 1.8 and ANSI /ANS 3.1 (1981). The facility Support Manager will be designated as the PSC Radiation Protection Manager reporting directly to the Program Manager for Decommissioning, separate from the operational entities of the organization. This will provide sufficient access to upper management attention independent of the project organizational elements responsible for cost and schedule compliance. The PSC Radiation Protection Manager will be responsible for the Radiation Protection Program. The PSC Radiation Protection Manager will interface directly with the Westinghouse Radiation Protection Manager, who will assist in the implementation and maintenance of the Radiation Protection Program. This communication channel will ensure that radiation protection implementation problems are brought to the attention of upper management so that timely and appropriate corrective actions can be taken.
Attachment I to P 91116 April 26,' 1991 Page 80 NRC Ouestion No. 33 (Section 3.2.2.4 Contamination Control)
'PSC should provide specific detaiis on how contaminstfon wi11 be controlled, when and how-enclosures will be used, and what methods will be expected to require special contamination controls."
EEGjesconset , Contamination will be controlled by employing a variety of engineering controls including HEPA ventilation, enclosures, strippable . paint, and area / component decontamination. Examples of contamination control methods that have been incorporated into the decommissioning plan include: (1) The PCRV will be filled with water -to control radioactive particulates that would normally be released when handled , in air. (2) Containment or enclosures of appropriate sire, equipped with HEPA ventilation, will be used as necessary to prevent the spread of contamination while contaminated graphite blocks and other components are being removed from the PCRV or otherwise handled. (3) A work platform will be installed on the PCRV after the DCRV head has been removed. The platform will be equipped with a HEPA-filtered ventilation system that will exhaust air from beneath the work platform. This airflow will minimize the spread of contamination.
. (4) A debris collection system will be used in concrete cutting operations to minimize the spread of contamination.
(5) Strippable paint or other suitable enclosures will be applied to some radiologically clean components or areas to prevent cross contamination. Additional contamination control methods will be considered during job planning and work package review. Isolation containments may be used if the surrounding work area i s uncontaminated or is much
- cleaner than the work area to minimize the spread of contamination.
Radiological surveys wil1 routinely be conducted to identify _ and measure contamination levels. These data will be used to identify if additional controls are warranted. 1 w m y *-
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Attachment I to P-91118 April 26, 1991 Page 81 KRC _0uestion No. 34 (Siction 3.2.3.1 ALARA Pro. gram Elemenin_3 Trainina and the Use of Modypih "The traf."ing requirements for the radiation protection staff have not been identif fed. This sectton should include a description of the training requirements." E1C_Et122niti The training requirements for the radiation protection staff are outilned in the response to Question 19 above. Radiation protection personnel will participate in mock up training for complicated or potentially high exposure jobs. This will ensure that radiation protection personnel are knowledgeable in the work activity and provide an opportunity to identify ALARA measures to reduce exposure.
i Attachment 1 to P 91118 April 26, 1991 Page 82 REQ _ question Ng. 35 (Section 3.2.5 R m iratory Control):
"What is the criteria for use of respiratory control devtces such as masks or air paks?"
PSC k qqgn1u Respiratory protection equipment will only be used whon engineering controls, including process modification, containment or ventilation controls are not practicable. When necessary, respirators will be used to maintain intake of radioactive material and other contaminants by personnel to levels "As Lew As Reasonably Achievable" and consistent with the goal of maintlining the total effective dose ALARA, The selection of respiratory protection equipment will be based on work area survey data or expected airborne cont amination levels. All work tasks in contaminated areas will 'oe evaluated for respiratory protection. Special attention will b1 given to the need for respiratory protection when the task involves any of the following operations:
1. Thermal Cutting,

2. Concrete Scabbling,

3. Welding,

4. Grinding, or

5. Concrete Demolition.
The respiratory protection equipment will be selected so that it provides a protection factor greater than th9 multiple by which peak concentration of airborne radioactive materials in the working area are expected to exceed. Assigned protection factors will not exceed those specified in 10 CFR 20 Appendix A. If the selection of a respiratory protection device with a protection factor greater than the peak concentration is inconsistent with the goal of keeping total effective dose ALARA, equipment with a lower protection factor will be used if it results in keeping total effective dose equivalent ALARA.
1 i l Attachment 1 to P 91118 April 26, 1991 Page 83 NRC Ouestion No. 36 (Section 3.3 Radioactive Waste Manac?n.ent):
"A. Describe the use, if any, of existing plant area, process, and effluent radiological monitors during the dismantlement and decontamination phases. If some or all of the existing plant monitors will be taken out of service, or supplemental monitors provided, describe the locatfon, purpose, and detectton l capabilities of these monitors.
B. Compliance with NRC and State disposal requirements should be added to the list of consideration for a radioactive waste management program." PSC Response: A. The use of the plant area, process and effluent monitors is described in the proposed Decommissioning Technical Specifications. The use of supplemental monitors will not be required. Monitors proposed to be used by the decommissioning technical specifications are described as follows:
1. Area Monitors; The purpose of the area radiation monitors is to alert personnel of unexpected changes in exposure rates. lhe area radiation monitors are capable of detecting gamma radiation.
The two (2) area radiation monitors listed below may be utilized to meet the requirement of the decommissioning technical specifications. If it becomes necessary to replace or substitute monitors for the area radiation monitors, the replacement monitors will have local alarms. The alarm set points will be established according to the expected exposure rates and established procedures. Area Monitor locations: RT-93252-1 Refueling Floor Northeast - Reactor RT-93251-6 Reactor Building Truck Bay Reactor
2. Process Monitors:
None of the existing plant process monitors will be maintained as operable. The PCRV Water Cleanup and Clarification System will have a process monitor to identify the need for filter / demineralizer replacement. This monitor will be capable of detecting gamma radiation and the alarm set point will be established by procedure.
3. Effluent Monitoru The plant liquid and gaseous effluent monitors will be maintained and operated in accordance with che Offsite Dose
Attachment I to P 91118 April 26, 1991 Page 84 Calculation Manual (00CM). Supplemental effluent air monitoring in the form of air samples for areas or operations remote from the Reactor Building with air discharge capabilities will be maintained. Monitoring capabilities include beta / gamma radiation measurement of samples. D. Section 3.3.3.1 of the Proposed Decommissioning Plan specifically states that "the radioactive waste disposal program will follow the regulations established in 10 CFR 20 and 10 CFR 61, the disposal site criteria and other applicable federal and State regulations."
. . , . , . _ - - , - , . . . , . . . . , _ _ - - _ , , . , , . . - - -- .w -. *v--,-=-=e----- - - - - - -
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Attachment 1 to P 91118 April 26,.1991 Page 85 NRC Ouestion No. 37 (Section 3.3.2 Waste Processinal:
"PSC should discuss how 1iquid and solid wastes wi11 be processed using existing systems or by the use of service contractors.
Decontaminatfun chemical wastes should be spectfical1y identiffed and special processing requirements for them should be discussed in detail." PSC Resoonse: Liquid and solid wastes will be processed and disposed of in accordance with written approved procedures. Typical liquid waste expected includes oils from plant systeas, water from PCRV bleed and feed operations, and sludges from diamond wire cutting and sump clean-outs. Disposal of contaminated oils is expected to be accomplished by transfer of the oil to a licensed vendor for incineration. The chemical / hazardous nature of oils will be known prior to transfer for incineration. Water from the PCRV will be processed through the PCRV Water Cleanup and Clarification System and discharged through normal plant effluent systems. Sludges will be dewatered and dried, or solidified / absorbed using disposal site approved solidification / absorbent media. Solidification, if required, will be controlled by an approved Process Control Program (PCP). Solid wastes will be processed in accordance with written procedures. A general plan for solid waste processing is as follows:
1. The waste will be initially identified at the point of generation as to the type 'sf material and exposure rate.

2. The material will be segregated to allow for decontamination on site, shipped to an off-site vendor for volume reduction or packaged for disposal at an approved disposal site. -
Chemicals used for decontamination will be evaluated for hazardous constituents using 40 CFR and the Material Safety Data Sheets (MSDS). Decontamination chemical wastes expected include acids, caustics, detergents and non hazardous solvents. The specific chemical for a particular application will depend on the material to be decontaminated. Acids or bases will be neutralized and solidified or used for water chemistry control in the PCRV water clean up system. Detergents and other water based solvents will generally be associated with damp rags or wipers. The wiping material will be dried prior to packaging for disposal or volume reduction. I
Attachment I to P 91118 April 26, 1991 , Page 86 , NRC Ouestion No. 38 (Section 3.3.2.2 Onsite Processino of Liouid Wastes):
"A. This section states that it is anticipated that large volumes of tritfum will be released from the graphite and picked up by the water. PSC has indicated that eventually the total tritium concentrations are expected to reach an equilibrium ifnit below the unrestricted release Ilmit for water discharge. What are the total concentrations and total curie quantities of tritium expected to be released from the graphite blocks?
l B. How will potential exposure of workers to tritium be handled 1 with respect to ALARA principles?" PSC ResDonse: A. In assessing levels of tritium in the graphite reflectors and side spacer blocks, the Fort St. Vrain Activation Analysis (Attachment 2 of this letter) conservatively assumed upper limit concentrations of impurities in the graphite, and also assumed that none of the tritium formed by neutron activation of impurities migrated out of the graphite. Based on these assumptions, the amount of tritium residing in the graphite was computed to be approximately 100,000 curies. However, it is expected .that the actual amount of tritium will be significantly less than this. Based on data on measured tritium release rates from graphite and assuming the conservative estimate of 100,000 curies, levels of trittum in the water system were estimated as a function of time for various purge rates from the system. The tritium will be removed by a feed and bleed method and released to ensure concentrations are below. 10 CFR 20 limits. The maximum . concentration in the PCRV water system is expected to be about 0.3 microcuries/cc and a maximum total water inventory of 310 curies. During .the initial purging operation, the PCRV water tritium concentration is expected to drop to -below 0.1 microcurle/cc within 40 days, and continue to decrease thereafter. The integrated total tritium released from- the grapl.ite blocks into the PCRV water system is predicted to be slightly above C00 curies. A plot of total tritium predicted to be in the PCRV water system vs. time is shown in the attached figure. B. The exposure to personnel will be controlled and monitored in accordance with 10 CFR 20, " Standards for Protection Against Radiation", Regulatory Guide 8.32, " Criteria for Establishing a Tritium Bioassay Program", and approved written procedures. In addition, portable ventilation will be utilized on the PCRV work platform to exhaust the local area above the flooded PCRV. This system will move the tritiated water vapor away from the work area. The need for other protective measures, such as
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Attachment I to P 91118 April 26, 1991 Page 87
" wet suits", will be evaluated based on the radiological conditions on the work platform.
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Attachment I to P 91118 April 26,'1991 Page 88 FIGURE 1 ESTIMATED TRITIUM INVENTORY IN PCRV WATER SYSTEM Curies 350 - 300 - - - " - - - - - -- " - - " - - - - - - - ~ ~ 250 -- - - - - - - - l 200 -- - - - 150 - " . - - - - - - - - - - - - - - - - 100 - - - - - - - - - - - -- - - . - - ----- 50 -- - - - - - - - - " -- " "- - - 0 O 10 20 30 40 50 60 70 80 Days After Flooding PCRV Curies s
Attachment I to P 91118 April 26, 1991 Page 89 NRC Question No. 39 (Section 3.3.2.3 Release of Airborne Contamination):
"Specify the efficiency, capacity, and replacement frequency for the venti 1atton ff1ters and equipment."
PSC- Resoqtui.gl . As stated in FSAR Section 6.2.3.2.3, the HEPA filters in the existing Reactor Building ventilation system conform to the requirements of the following documents:
- Military Specification MIL-F-51068 (D) - Underwriters Laboratory Standard UL-568 1977 - USNRC Regulatory Guide 1.140 - USNRC Regulatory Gutde 1.52 Revision 2 - ANS! Standard N509 1976 The design specifications require that the maximum penetration is 0.03% when tested with thermally-generated mono dispersed dioctylphthalate (00P) smoke having a light scattering geometric mean droplet diameter of 0.3 microns. Each HEPA filter is flow ,
leak- and penetration tested by both the manufacturer and by a Department of- Energy (DOE) Filter Test Facility to ensure conformance to the design. specifications and overall operating efficiency of not less than 99.97%. However, as discussed in proposed Decommissioning Technical Specification (DTS) 322, the filter penetration and bypass acceptance limits in- the surveillances are applicable based on a HEPA filter efficiency of 95%. The HEPA filter bank will be tested using the test procedure guidance in Regulatory Position C.S.a and C.S.c of Regulatory Guide 1.52, Rev. 2, March 1978, with a flow rate of at least -17,100 cfm to verify that the filter penetration and bypass leakage test . acceptance criteria of 1% is met. The replacement frequency of the HEPA filters in the existing Reactor Building ventilation exhaust system is also identified in Decommissioning Technical Specification 3.2, and is based upon either high exhaust radiation readings (or alarm) in the ventilation exhaust duct, or upon exceeding the maximum allowable )ressure differential which I.1dicates that the filters are filled wit 1 dust, t
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Attachment I to P 91118 April 26; 1991 Page 90 NRC_0uestion No. 40 (Section 3.3.2.4 Decontamination Techniaues11 "The decommissioning plan states that mixed waste may be converted to a non hazardous product. PSC should provide detailed information on the treatment processes to be used for this conversion." PSC Response: Certain types of mixed waste may be treated through various procedures (e.g., solidification) to render them non hazardout. If a mixed waste stream is identified, a treatability study will be performed to determine if it can be made non hazardous. At this time, there have been no mixed waste areas identified, furthermore, there will be no processes used during the course of the decommissioning that will create a non treatable mixed waste.
Attachment I to P 91118 April 26, 1991 Page 91 HRC Ouestion No. 41 (Table 3.1-1 Waste Volume Reduction):
"This tabic indicates that incineration and melting are waste volume reduction options. Are these processing options going to be carried out on site or off site? If they will be performed on site, PSC should provide detailed information on the wastes to be treated, the system designs, effluents, and waste product characteristics. If incineration is to be performed on site, specific NRC approval will need to be obtained.
PSC Response: Incineration and melting for volume reduction will be accomplished offsite at a f acility licensed for these specific waste volume reduction techniques.
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Attachment I to P 91118 April 26, 1991 Page 92 NRC Ouestion No. 42 (Section 3.3.2.5 Volume Reduction): "PSC should provide detailed information on waste volume reduction equipment that will be used on site, how this equipment is consistent with Regulatory Guide 1.143, how volume reduction operations will be monitored, how effluent discharges will be controlled, and how occupational safety will be ensured." PSC Respons11 Some solid materials may be volume reduced onsite using the existing Fort St. Vrain compactor or a mobile compactor supplied by Westinghouse. If a compactor is used, its effluent discharge will be filtered by HEPA ventilation and airborne contamination surveys will be conducted during operation. Compactor operation will be controlled by a specific radioactive waste procedure to ensure safe operation. Compacted waste will be placed in 52 or 55 gallon drums, compacted, overpacked, and shipped directly to the disposal facility. All activities will be performed in accordance with applicable radiation protection and radioactive waste procedures. USNRC Regulatory Guide 1.143, " Design Guidance for Radioactive Waste Management Systems, Structures and Components Installed in Light-Water Cooled Nuclear Power Plants," addresses liquid waste processing systems and does not appear to be applicable to waste compaction systems. However, this document will be considered to ensure the compactor is operated and controlled in a manner commensurate with the need to protect the health and safety of the public and plant personnel. . l
, Attachment 1 to P-91118 i April 26, 1991 Page 93 NRC Ouestion No. 43 (Section 3.3.2.6 Waste Packaainal:
"What types of waste containers will be used (e.g., SS gal. drums, 4x4x8 metal boxes, 100 ft3 HIC's, etc.), what are their capacities and weight limitations, and what materials will be packaged in them?"
PSC Response: Examples of the waste containers that may be used are drums ($2 gallon, 55 gallon), boxes (2'x4'x6', 4'x4'x6'), liners, HIC's, sea / land containers, casks, and soecialty containers. The capacity and weight limitations are governed by the activity levels and classification of the enclosed materials as specified in 10 CFR Part
61. Examples of materials include piping, valves, graphite blocks, concrete rubt>1e, etc. The waste container will be determined by the size, weight, classification, and activity level of the different types of materials. Guidance for this activity will be contained in radioactive waste procedures.
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Attachment I to P 91118 April 26, 1991 Page 94 tLRC Ouestion No. 44 (Section 3.3.3.1 Waste Discosal Proarami:
"What waste storage facilitles will be used and how are these facilities consistent with the storage recomendations in Generic Letter 8138?"
PSC Responsn Waste storage facilities planned for use during decommissioning activities include:
1. The Independent Spent Fuel Storage Installation (ISFSI) will be used for greater than Class C wtstes (GTCC), if any, pending approval of an appropriate disposal site. (No GTCC wastes are currently expected.)

2. The New Fuel Storagc Building will be used as a processing and storage area for dry low level wastes.

3. The I ACM Building (Compactc Building) will be used as a processing and storage area for dry and dewatered low level wastes.

4. The Reactor Building will be used for the storage of liquid and solid wastes.

5. Trailers and sea / land containers will be stored and used onsite to house dry and solidified low level waste.

6. Selected yard areas will be used for short term storage of packaged waste staged for transport.
The activity levels of wastes stored in these areas will be controlled to levels as evaluated in the accident analysis, Section 3.4 of the Proposed Decommissioning Plan. Comoliance with Generic letter 81 38 Generic Letter 81-38 authorized nuclear power reactor licensees to store their generated low level radioactive waste onsite on an interim basis until it could be transported to a waste disposal facility. The interim period was defined as not longer than 5 years, and the letter directed licensees to submit any needed Technical Specification amendments to allow onsite possession and storage of byproduct material. Storage for greater than five years requires a 10 CFR 30 license. Safety evaluations have been performed that assess and permit storage of low level radioactive waste on the Fort St. Vrain site consistent with the guidelines of Generic Letter 81-38 and the Standard Review Plan (NUREG-0800), Appendix ll.4-A. The Fort St. Vrain Technical Specifications permit possession and use of
Attachment I to P 91118 April 26, 1991 Page 95 byproduct, source, and special nuclear material in quantities as I required pursuant to 10 CFR 30, 40 and 70. Due to the building seismicity and other drainage and collection requirements for the storage of wet radwaste, PSC does not intend to store wet / liquid radwaste outside the Reactor Building. The Reactor Building was designed and built with drainage systems that route , spillage to collection points / sumps that are monitored for raaicactivity and properly processed. Other forms of radwiste may also be stored in the Nietor Building without significant concern, due to the building's iJditional features relative to fire detection and suppression, and its filtered ventilation system. The Compactor Building is a steel building constructed on a concrete foundation, with its own " wet pipe" fire suppression and fire detection systems. This building has two concrete basins that may be used -to store barrels of dewatered wastes, consistent with the recommendation of Generic Letter 8138. Other dry and solidified - wastes may be stored in this building in amounts consistent with limitations of the decommissioning accident analyses. A radwaste compactor, with a self-contained HEPA filtered ventilation system, is also housed in this building. , The New Fuel Storage Building will also be used to store packaged dry and solidified low level radwaste. A safety evaluation has determined that no increase in an accident probability will result from radwaste storage in this location. As stated in the decommissioning accident analysis, fire suppression and fire detection systems will be provided before combustible radwaste can be stored in the New Fuel Storage Building. ,
- Trailers- and sea / land containers have been evaluated to house dry and solidified radwaste. Accident scenarios have been postulated and the . total allowable activity levels for storage are controlled accordingly. Yard fire hydrants are available for use if necessary, Certain large radioactive components (such as helium circulators packaged for shipment) may be stored outside within the protected <
area while awaiting shipment offsite. Tie-down systems will be considered for components stored outside, and will be installed when needed. Steps will be taken to protect containers from external corrosion as required. All interim low-level radwaste storage locations described above exist within the plant's protected area. Radiation protection procedures will be implemented that will specify requirements for processing and packaging radwaste. Procedures will also be implemented to provide suitable instructions to establish radiologically controlled areas to ensure occupational and public , exposure are kept within the requirements of 10 CFR 20, 10 CFR 100, and 40 CFR. Procedures will be implemented to monitor storage areas periodically and to check waste container integrity. s ,-
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Attachment I to P 91118 April 26, 1991 Page 96 tdQ_Qvestion No. 45 (Section 3.3.3.2 Proiected Radioactive Waste Generation)1 "A. This section indicates that the BOP radivactive contaminated waste af ter processing and volume reduction will be 853 cubic feet. Table 3.3 4 BOP Waste Volume Estimates indicates a volume of 12577.5 cubic feet exists before reduction anc a quantity of approximately 6500 cubic feet RFP. What does RFP mean? holain the differences in the volumes of waste for the BOP. Also include an estimate of the volume of contaminated asbestos. B. The staff concludes that there are significant uncertainties in the volumes of low level wastes generated because the activation analysis is based on generic studies using average values instead of actual samples. What assurances does PSC have that the disposal facility will accept increases in waste volumes? C. Table 3.3 3 indicates that there will be 401 cubit feet of GTCC wastes from the metal control rod reflectors. Won't the boronated pins and hastelloy cans also be GTCC wastes? The discussion in this section should specifically identify these materials as GTCC wastes." PSC Resoonse: A. The radwaste volumes listed in the "RFP Volume" column were the initial estimates provided in the Request for Proposal (RFP) and no longer have any significance; this table will be revised to delete this column. The pre volume reduced volume of
radioactive waste is identified in response to NRC Question No.
8. A significant volume reduction can be realized during the dismantlement process, dependent upon the methods (o.g.,
sorting techniques) used and developed during removal operations and voluma reduction of the materials prior to disposal. No contaminated asbestos was identified during the initial assessment of plant radiological conditions as summarized in the RFD. Necessary precautions will be taken during asbestos removal to avoid cross contamination and the generation of additional radioactive waste, if contaminated asbestos is identified, it will be packaged and disposed of in accordance with 10 CFR 61, the disposal facility license and applicable radioactive waste procedures. B. There is no assurance that the disposal facility will accept increases in waste volumes beyond that which may be allocated to the generator by the proposed contract between the Rocky Mountain Compact and the Northwest Compact. However, PSC is
Attachment I to P-91118 April 26,1991 Page 97 confident that sufficient disposal capacity exists for the Fort . St. Vrain decommissioning project for the following reasons: I
1. PSC has access to the Beatty, NV, disposal site until December 31, 1992. The Beatty, NV, disposal site does not have a maximum allocation for PSC waste.

2. A proposed agreement has been prepared between the Rocky Mountain and Northwest Compacts that will allocate 140,000 ft3 of disposal space for PSC and access to the Hanford, WA, site from January 1, 1993 until the completion of decommissioning. To date, this proposed agreement has not been finalized between the two compacts. As previously committed. PSC will keep the NRC apprised of final changes in this status.

3. The total estimated volume of radioactive waste to be disposed LPRIOR TO VOLUME REDUCTION) is approximately 100,000 ft3 The total estimated volume of radioactive waste to be disposed (AFTER VOLUME REDUCTION) is approximately 80,000 to 90,000 ft3 .

4. Approximately 5000 ft3 of waste will be disposed of at the Beatty, NV, facility prior to closure. This conservatively assumes a 6 month delay in defueling activities.

5. Therefore, .approximately one-half of the Hanford, WA, allocation is projected to be needed. With a factor of two margin, PSC can safely project that sufficient disposal capacity exists for completion of the decommissioning project. ,
C. PSC does not anticipate that the waste form containing the boronated pins and Hastelloy cans will be classified as GTCC waste. Current 10 CFR 61 characterization data from MCRB and RCD sample analyses indicate that these waste streams are not GTCC wastes. The license conditions at both the Beatty NV and Hanford, WA, disposal facilities provide for activity-averaging over the waste form being buried. Utilizing this approach, the disposal of the boronated pins and Hastelloy cans with associated graphite blocks will not be GTCC waste,
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Attachment 1 to P-91118 April 26, 1991 Page 98 NRC Ouestion No. 46 (Section 3.3.3.5 Mixed Wastel:
"The decommissioning plan implies that mixed wastes will be " declassified." We are unaware of any process to " declassify" mixed wastes. Consequently, PSC should plan to handle any mixed wastes that might be generated in accordance with both the applicable NRC and EPA requirements."
PSC Rescotigi All hazardous chemicals and materials will be subjected to a chemical control review to determine if a non hazardous or a less toxic chemical can be substituted to prevent generation of mixed wastes, in the event that hazardous chemicals or materials must be used, procedures will ensure that all waste minimization techniques will be applied during usage. Steps will be taken to ensure that if hazardous materials must be used, the necessary controls will be in place so these materials will not inadvertently become radioactively contaminated. if some hazardous material does inadvertently become radioactively contaminated, it will be considered as mixed waste and subject to applicable regulations. If mixed wastes are generated, they will be managed according to Subtitle C of RCRA to the extent it is not inconsistent with NRC handling, storage, and transportation regulations, if technology, resources and epproved processes are available, PSC and the Westinghouse Team will evaluate the processes for rendering mixed waste "non-hazardous' to determine its adaptability to Fort St. Vrain decommissioning activities. PSC does not intend to petition the EPA to delist any mixed waste. However, if PSC determines it is necessary to delist any mixed wasto, the procedures outlined in 40 CFR 260.20 and 260.22 will be used to exclude that waste form from regulations.
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Attachment I to P 91118 April 26, 1991 Page 99 NRC Ouestion No. 47 (Table 3.3 1 PCRV Wastes): "How will stabilization be accomplished? Table indicates that defueling elements and core support posts would be incinerated. How would this be accomplished?" PSC Responga Some stabilization will be accomplished through the use of solidification media approved by the disposal site to which the waste will be shipped and will be performed in accordance with the criteria established in 10 CFR 61. Some solid waste forms may meet the stabilization requirements by virtue of their physical and chemical form per 10 CFR 61, if the graphite defueling blocks and the graphite core support posts / blocks meet the acceptance criteria of an offsite licensed incinerator facility, they will be transported to that facility for incineration. Facilities are available that have successfully demonstrated the ability to incinerate graphite.
i Attachment I to P 91118 April 26 1991 Page 100 NRC.0uestion No. 48 (Tables 3.3-3 and 3.3 4 Waste Volume Estimates):
"What does the heading "Cf/ load" mean? Why are the columns in both tables incomplete? The activity of each of the weste items listed should be provided."
PSC Resoonse: Any load of radioactive waste that exceeds the maximum activity 1 allowed without the assessment of a surcharge by the disposal i facility has been indicated in the column entitled "Ci/ load." The j surcharge for the "overlimit" amount is indicated in the " Curie surcharge" column. The "Ci/ Load" column is left blank for loads that do not exceed the maximum quantity silowed without the assessment of a surcharge by the disposal facility. It is impossible at this point to identify the activity of each package because the activity level will vary with the size of the section/ component, the number of pieces that can be placed in the selected container, the location in the PCRV or B0P from which it was removed, and the type of container needed to meet 49 CFR or 10 CFR 71 requirements. The activity will be determined for each package prior to shipment offsite. _ , . . . . - . - - . . . , , , _ _ , _ . _ , .m. , - . ,, ..
Attachment I to P 91118 April 26, 1991 Page 101 NRC Ouestion No. 49 (Section L4 Accident Analysish "Due to the complex procedures for removal, the PCRV head and core support floor structures, the possible collapse of the PCRV head or core support structure should be analyzed in the accident analysis." PSC Responst.;. An objective of the Proposed Decomissioning Plan accident analyses was to describe the consequences of the worst case scenario within different accident categories. Pnstulated accident scenarios involving the PCRV top head and the core support floor (CSF) were considered as heavy load drops, ar.d are addressed in Section 3.4.5 of the Proposed Decomissioning rian. The radiological impact of pastulated heavy load drops is governed by the radioactive source tarm of tho object dropped, and not by its weight. PCRV top head sections contain very little radioactivity in comparison with graphite reflector blocks. As discussed in Preposed Decomissioning Plan Section 2.3.3.7, the PCRV top head concrete will be cut into sections and then removed, leaving approximately 6 inches of concrete above the PCRV top head liner. This f,-inch concrete layer has the highest activity concentratir.n of PCRV concrete, since it had been exposed to the highest integrated neutron flux. Section 3.4.3 of the Proposed Decomissioning Plan analyzes the drop of 10% of this 6 inch thick concrete wafer (wetgbing approximately 3.8 tons). This accident is considered to produce bounding consequences for credible activated concrete arop accidents. The collapse of the entire PCRV top head is not considered credible, . since the top head will be supported by the rCRV structure tnd will be removed in Individual sections. Section 3.4.5 of the Proposed Decomissioning Plan addrasses the possible collapse or drop of the CSF. Ihis section identifies the possible alternative methods for removal of the 270 ton concrete CSF from the PCRV, which include (1) sectioning tl'e CSF within the PCRV and removing the pieces by means of the Reactor Building crane, or (2) raising the entire CSF above tha PCRV with specially installed high capteity jacks. The activated graphite blocks will be removed from the PCRV prior to removal of the CSF. The CSF concrete is predicted to contain six curies of activity. Therefore, a heavy ioad drop during this operat ton does not have the potential for release of significant quantities of radioactivity, if the entire CSF is raised by high capacity jacks, a drop of the CSF is not considered credible since such an accident would require multiple jack failures. Therefore, any impact from a PCRV top head or the CSF fall / collapse is enveloped by the drop of a side reflector block or the loss of PCRV Shield Water.
Attachment I to P 91118 April 26, 1991 Page 102 NRC Ouestion No. 50 (Section 3.4.3.2 Accident Descriotions):
"Why is the Co 60 and Fe SS in the concrete all assumed to be in the ,
rebar? Cobalt and fron are trace constituents in cements and ' aggregates." PSC Resoonse: Cobalt and iron are confirmed to be trace constituents in cements and aggregates, and the radiological impact from these impurities has been factored into the accident scenario. The activity loadings of Co-60 and Fe 55 in the 6 inch thick concrete wafer above the PCRV top head liner have been calculated to be: 1.43 Ci and 32.83 Ci i respectively, in the concrete. The concrete rubble drop accident is ' analyzed in Section 3.4.3 of the Proposed Decommissioning Plan. The assumption is made that -10% of the concrete in the 6-inch wafer, ' adjacent to the top head liner, is involved in the accident. Only 1% of the activity is assumed to become airborne as a result of the : impact. Thus, the Co 60 and Fe 55 contribute 0.00143 Ci and 0.0328 Ci respectively to the accident source term. The doses from the revised source term at the 100-meter Exclusion Area Boundary have also been calculated. The revised whole body - dose is 4.92 mrem, a 4.9% increase over the original dose analysis (4.69 mrem). The largest organ dose is still to the bone with a 54.7 mrem exposure -(2.3% increase over 53.5 mrem)). These dose increases are small. However, since they are more conservative, they will replace the existing dose values in the concrete rubble accident that have been reported in Section 3.4.3 of the Proposed 2 Decommissioning Plan. 1
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Attachment I to P 91118 April 26, 1991 Page 103 NRC Ouestion No. 51 (Section 3.4.6.1 Identification of Causel:
"This section states that procedures and controls will be developed to ensure sufficient spatial separation to preclude ftre propagatfan to an adjacent group of LSA containers. Will these procedures be part of Fire Protection Program Plan?
Why does the fire analysis use the single spacer blocks as the b source term for the evaluatfon? Wouldn't the large side reflector .~ blocks be more conservative?"
'5 p'
Ejit Response:
.y A. The Decomissioning fire Protection Pl an will contain a ',, 4 specific Fire Protection Operability Requirement (FPOR) to ~
ensure adequate spatial separation to preclude fire propagation to an adjacent group of LSA containers. Additionally, the FPORs will also address required fire suppression and detection for the facility. B. As shown in Table 3.4-5, " Waste Volume / Activities Estimates for the PCRV", the activated graphite blocks have the highest radioactive inventory. Moreover, the large side reflector blocks have a higher activity concentration of Fe-55 and Co-60 than the side spacer blocks. However, as stated in Proposed Decomissioning Plan Section 3.4.6.2, it is assumed that only 0.015 percent of these particles are released during a fire. This particle release fraction is consistent with the release fraction that was used for the analysis of postulated fires of radioactive materials in NUREG/CR 0130, " Technology, Safety and Costs of Decommissioning a Reference PWR Power Station and NUREG/CR 0672 " Technology, Safety and Costs of Decommissioning a Reference BWR Power Station." Due to the low fractional release of particulate activated products, the consequences from a fire involving activated graphite blocks are dictated by the tritium source term. As discussed in Proposed Decommissioning Plan Section 3.4.1, a tritium concentration of 10 microcuries/ gram was assigned to both the large side reflector blocks and the side spacer blocks, based on an evaiuation performed by GA Technologies, Inc. (Reference 15 of Proposed Decommissioning Plan Section 3.4). The side spacer blocks have a surface area to volume ratio that is approximately eight times larger than that of the large side reflector blocks due to wells in the side spacer blocks for inclusion of baron poison pins. Since the graphite oxidation rate is directly proportional to the surface area to volume ratio, a fire involving side spacer blocks has a higher potential for graphite oxidation than a fire involving large side reflector blocks. Therefore, a fire involving the side spacer blocks would have the highest tritium release source term.
Attachment I to P-91118 April 26, 1991 Page 104 Based on the current waste processing plans, the largest inventory of activated graphite for a fire scenario is a waste transport truck loaded with 230 graphite side spacer blocks. This amounts to a source term of approximately 3700 curies. To preclude the possibility of a fire with consequences greater than that analyzed in Section 3.4.6 of the Proposed Decommissioning Pl an, fire detection capabilities will be installed in the LSA container storage area and controls governing the spatial separation of the LSA containers will be implemented. Therefore, the postulated fire analyzad in Ocetion 3.4.6 of the Proposed Decommissioning Plan represents the bounding case for release of radioactive nterials from fire initiators.
Attachment I to P-91118 April 26, 1991 Page 105 [fRC Ouestion No. 52 (Section 3.4.7.1 (PCRV Shieldina Water Accident):
"The analysis assumes the PCRV shielding water is released by a rupture of the water cleanup system. Would the rupture of the PCRV penetration seals result in a more conservative accident?"
PSC Resoonse: g Although it is true that a large leak / break location can be postulated, the potential rupture of one or more of the PCRV penetration seals would not result in a more conservative accident. The Loss of PCRV Shielding Water accident assumes that the entire water inventory of the PCRV is emptied into the Reactor Building sump / keyway. Additionally, the dose analysis assumes the entire PCRV water volume is in the lower level of the Reactor Building instantaneously at the initiation of the accident rather than leaking or draining into the sump over a long time period.
Attachment I to P-91118 April 26, 1991 Page 106 NRC Ouestion No. 53 (Section 4.5.1 Survey Docume_n_tation):
"The Ifst of information that will be included on survey maps or forms should include the instrument serial number or equivalent identification informatlon."
PSC ResDantti The surveillance and survey forms to be used for the decommissioning will include the instrument serial number and will include other information that will ensure traceability of instrumentation, E h is
-_ to P 91118 April 26, 1991 Page 107 NRC Ouestion No. 54 (Section 4.6 Ouality Assurance):
"This section should provide a description of the ma,ior quality assurance program elements for radiological measurements which will be documented prior to inttiatton of the dismant1ement and decomissioning phase. The quality assurance elements shoul.1 be consistent in requirements wIth the importance of the activity and be consistent in requirements with the importance of the activity and be consistent with the quality assurance criteria in Appendix B to 10 CFR Part 50. The format and content of the program documentation may be based on applicable parts of NRC Regulatory Guide 4.15 recomendat ions. This program should also address measurements associated with radioactive effluents released from the site during dismantlement and decomissioning." PSC Resoonsei The NRC-approved PSC 10 CFR 50 Appendix B program will continue to apply for QA oversight activities until approval and implementation of Section 7 of the Proposed Decomissioning Plan. Specific activities performed prior to the initiation of the dismantlement and decommissioning phase (e.g., radiological measurements) will be controlled using specific QA programs or plans. Verification will be made that these QA plans or programs are NRC-approved plans, or that they meet the requirements of the Fort St. Vrain 10 CFR 50 Appendix B QA Program. Examples of these QA plans for specific activities include the following: Routine activities -WestinghouseSEGgAPlan Project management activities - Westinghouse NATO QA Plan Activities performed by WSEG will be accomplished in accordance with the WSEG Quality Assurance Plan (QAP-107). 'his plan encompasses all eighteen criteria specified in 10 CFR 50 Appendix B. The specific application of these criteria is ir accordance with the WSEG QA Program, which is an eighteen criteria program and is approved by the NRC in accordance with 10 CFR 71, Subpart H. Specific radiological measurements will be perfermed in accordance with the WSEG QA Program (QA-100) and implementing procedures. The WSEG QA Plan (QAP-107) describes project specific QA activities and controls, The proposed Decomissioning Technical Specifications, Section 5.4.4(a) and (b) speci fy that the Radioactive Effluent Controls Program and Radiological Environmental Monitoring Programs shall be contained in the Offsite Dose Calculation Manual. This manual is one of the implementing manuals for the Radiation Protection Program, and will be performed under the Quality Assurance Plan in Section 7 of the Proposed Decomissioning Plan.
- Nuclear and Advanced Technology Division
l Attachment 1 to P 91118 April 26, 1991 Page 108 NRC Ouestion No. 55 (Section 5.1 Decommissionina Fixed Price):
"10 CFR 50.82(b)(4) requires PSC to provide a cost estimate. The submittal of a fixed price quote from Westinghouse does not allow the staff to make the required conclusions on cost. PSC has been provided RAls on the information necessary to be included in the cost estimate. PSC must provide a cost analysis as requested by the NRC in previous RAls and as recommended in DG-1003."
PSC Resoonse: See response to Question 3.
Attachment 1 to P 91118 April 26, 1991 Page 109 HRC Ouestion No. 56 (Section 5.3 Ma.ior Assumotions. Bases and Scope of Fixed Price Contract):
"The section states some of the assumptions made in the fixed price.
Assumption 2 states that radionuclide inventories, activation analysis and dose rates are based on information in Section 3. Most of this information is based on generic studies with average values for input and no actual sampling has been taken. This raises serious question about the validity of the cost estimate and the volume of waste produced. Also, this section states that no cost allowance was factored into the estimate to scrommodate changes once actual samples are taken. Since several of PSC's decommissioning discussions are preliminary in nature and are subject to major changes as stated in the Proposed Decommissioning Plan, this uncertainty raises additional concern about the validity of the cost estimate. For example, if the tendon tubes are not sufficient for the diamond cutting wires, core drilled holes will be required. This will result in a significant impact in time, exposure, and schedule extension. However, Assumption 7 states that schedule delays were not included in the cost estimate. Therefore, a detailed cost estimate must be provided before the NRC can make a finding that it is in a reasonable range." PSC Resoonse: The fixed price contract for the Decommissioning of Fort St. Vrain is written so that the Contractor is responsible for dismantlement, removal and decontamination of the PCRV and its components to the extent it has been specified to be contaminated. For example, the CSF is required to be removed and disposed of by the Contractor within the scope of work and within the agreed fixed price. The Contractor originally proposed a technically feasible method and now is considering alternate methods that could be used. If the Contractor chooses to utilize a different method that is technically satisfactory, it will be done at no additional cost to PSC. The same is true for the use of tendon tubes for diamond wire cutting. The Contractor is responsible for removal of the activated concrete, and if use of the tendon tubes is not satisfactory, then the drilling of core holes or the use of an alternate technique is the responsibility of the contractor within his fixed price. To determine the amount .of contaminated side wall concrete to be removed, PSC performed a conservative activation analysis; the verification of the activation analysis results have been forwarded to the NRC in PSC letter (Crawford to Weiss) dated April 15, 1991 (P-91137). Based on this verification, the volume required to be removed, even in a worst case scenario is contained within the planned volume of concrete to be removed. In the unlikely event that a greater depth of concrete must be removed, a change to the contract will be negotiated. in this case, PSC will not pay for the removal of additional concrete unless it actually is found to be
m j Attachment I to P-91118 April 26, 1991 Page 110 required. PSC is convinced from results of the activation analysis that there is a very small likelihood that this will occur. The Contractor has accounted for the possibility of difficulties or the need for alternate methods within the fixed price. PSC has a: counted for the possibility of limited scope changes with funds as part of their reserve. PSC does plan to analyze core drills taken from the PCRV sidewall concrete after the sptnt fuel has been removed to further characterize the concrete. As discussed in PSC's response to NRC Question No. 4, PSC is preparing a detailed decommissioning cost estimate that will be submitted to the NRC before the end of May 1991.
i Attachment 1 to P-91118 April 26, 1991 Page 111 NRC Ouestion No. 57 (Section 5.5 Decommissionino P13n}.1 "10 CFR 50.82(b)(4) requires a funding plan. However, PSC has not submitted a funding plan at this time. PSC has indicated that a final decision has not been reached for funding the Fort St. Vrain decommissioning. PSC has also indicated that if their funding approach is not approved, they wi11 not proceed with the DECON option. PSC must address these areas before an SER can be developed." PSC Response: PSC recognizes that the Funding Plan must be submitted, reviewed and approved before the Proposed Decommissioning Plan can be approved and an SER issued. The status of PSC's Funding Plan is addressed in PSC's response to NRC Quastion No. 4 of this attachment.
Attachment 1 to P-91118 April 26, 1991 Page 112 NRC Ouestion No. 58 (Section 5.8 Uudates to the Decommissionina Fundina Plan):
"The possible inaccuracies associated with the base assumptions for the firm fixed price may result in large increases in the cost of decomissioning. What assurance can FSV provide to demonstrate sufficient additional funding will be available to complete the project?"
PSC Resoonse: PSC is preparing a detailed cost estimate to support its projected cost for decomissioning that will not rely on the firm fixed price contract as previously proposed by PSC. The detailed cost estimate will be submitted by the end of May 1991, and will be a stand-alone cost ostimate that will provide the full scope of assumptions, as well as a sufficient level of detail for the NRC to independently confirm cost estimates for decommissioning tasks. Contingencies will also be addressed.
l l Attachment I to P-91118 April 26, 1991 Page 113 NRC Ouestion No. 59 (Section 7.1 Policy Statement):
"A. Organizational charts and functional responsibility statements should be revised to clearly specify the lines of authority and areas of responsibility for each organizational unit, including the QA staff. These should show, in addition, how the various contractors and any subcontractors will interface with each other and PSC management, and how PSC will exert sufficient authority and control to ensure project success in carrying out its responsibility for all activities conducted under the PSC 1icense.
6. The QA plan should identify, describe, and define all the major organizational and work interfaces to ensure clear and effective communication between organizations and units, and proper coordination and control of all work activities. The QA plan needs to address these concerns.
C. The QA plan appears to assign responsibility to the PSC QA staff only for ensuring quality by performing surveillance and audit functions. What is the role of the QA staff in achieving quality: (1) by involvement in establishing work procedures to ensure inclusion of appropriate QA checkpoints and documentation; (2) by identifying, monitoring, and ensuring implementation of necessary corrective actions; and (3) by reviewing program status, problems, and needs, or by otherwise directly helping to achieve quality? D. Since the work efforts themselves are short in duration, how will surveillance and audit functions be scheduled to ensure timeliness, and what is the schedule for preparation of the necessary work procedures?" PSC Response: A. Organization charts and functional responsibility statements are included in Sections 2.4, L 5, and 7.3 of the Proposed Decommissioning Plan. Figure 2.4-1, PSC Decommissioning Organization, indicates the major interface between PSC and the Westinghouse Team. The PSC Program Manager for Oecommissioning interfaces with the Westinghouse Project Director for decommissioning activities. Paragraph 2.4.4 identifies that the Westinghouse Team Project Director will report to the PSC Program Manager for Decommissioning. Figure 2.5-2, Westinghouse Team Organizational Chart, also identifies the PSC-Westinghouse Team interface. Paragraph 2.5.1 identifies that the PSC and Westinghouse team interface will be structured to clearly demonstrate PSC compliance and control as required. Paragraph 2.5.2.3 explains that Westinghouse will be responsible for overall project management. Further, paragraph 2.5.3.1, the Fort St. Vrain ;
Attachment 1 to P-91118 April 26, 1991 Page 114 Westinghouse Team Project Director, identifies that (1) the Westinghouse Team Project Director will provide a single point of contact for PSC on the decommissioning effort, and (2) the Project Director reports to the PSC Program Manager for Decommissioning and is fully responsible for Westinghouse Team personnel, plant safety, prevention of environmental occurrences, quality assurance, project integrity, costs, schedule, efficiency, and technical output of the overall program. Revisions to organization charts and functional responsibility statements will be incorporated to clearly identify lines of authority and areas of responsibility for each organizational unit, including the PSC QA staff. The revised organizational charts will Ww the major interfaces between PSC and Westinghouse, ana between Westinghouse and the various team members, and how PSC will exert su f ficier,t authority and control to ensure project success in carrying out its responsibility for all activities conducted under the PSC license. Section 7.3.2 of the Proposed Decommissioning Plan will be revised to identify that PSC will interface directly with the Westinghouse Project Director who represents the focal point for decomissioning activities. Working interfaces will be defined in specific implementing procedures. B. Major organizational interfaces between PSC and Westinghouse, and Westinghouse and its team members are defined in organizational charts Figures 2.4-1 and 2.5-2, respectively. Functional responsibility statements are provided in Sections 2.4, 2.5 and 7.0. The various work interfaces will be dascribed in specific implementing procedures. C. The PSC QA staff will function primarily in an oversight role in verifying achievement of quality. (1) As part of this function, PSC QA will review select Westinghouse Team prepared work packages and identify appropriate PSC QA checkpoints. (2) The PSC QA staff will identi fy the need for corrective actions through the surveillance (monitoring) and audit functions, and will track corrective actions in accordance with Section 7, Part 18 " Corrective Action", described in the Proposed Decommissioning Plan. (3) The surveillance (monitoring) and audit processes will be utilized to review program status, problems and needs, and document outcomes and results. PSC QA personnel will be available for consultation with all organizations for all issues involving quality. QA personnel are also available to provide other management support to the project team that does not conflict with their independent status.
_ _ . -- . _ . ~ . _ . __ - - _ _ . . _ _ _ _ _ . _ . . _ - _ . . ._ _ . .m - . .-_ _ ___ Attachment 1 to P-91118 April 26; 1991 Page 115=
0. Surveillances_ (monitoring _ activities)- will be conducted commensurate -with the' work activities being performed. A specific schedule is not planned, but surveillances (monitoring activities) will be evaluated- relative to the types of activities occurring to _ assure appropriate activities are covered. Auditing will be . initiated as early as practicable, consistent with the schedule for accomplishing the activity, to assure timely implementation of the Quality Assurance Plan.
i 1 l
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Attachment.1-to P-91118 April 26, 1991 Page ll6.
~NRC Ouestion No. 60 (Section 7.7.2 Procedures): "A. Specify the ' structures and activities to be covered,'and the titles of al1 procedures whose implementation is subject to oversight in the QA plan.
B. Identify the specific procedures that govern the conduct of QA staff in executing the QA functions for which they are responsible." PSC Resoonse: A. The structures- and activities to be covered are identified in the- Proposed - Decommissioning Plan, Sections 1. 2.1, 2.3.1 and
l. 2.2.3. See response to Question No. 60B regarding procedure l
titles. B. The general categories of procedures -are presented in Section 7.7.2 of the Proposed Decommissioning Pl an. The specific procedures are being developed. In addition to identifying the specific procedures that govern the conduct of the .QA staff, the titles of all proceduras subject to-QA oversight will be identified in QA implementing procedures. l l-( ... l L L 1
Attachment I to P-91113 April 26, 1991 Page 117 Aibestos Removal Programt in PSC's letter dated January 14, 1991 (P-91001), PSC committed to provide an update on the details of the asbestos removal program, fort St. Vrain was completed in the early 1970's, and continuing maintenance and modifications have been occurring at the plant since that time. As a result, there is asbestos containing material in the plant. Planning for all decommissioning activities will recognize this potential, and appropriate precauHons will be taken to ensure the health and safety of the workers and the public. As part of the preparation for the initiation of actual decommissioning activities, a characterization survey was performed of those systems and areas that will be involved with the decontamination and dismantlement of the plant. The t5aracterization found that 60 of 155 samples taken contained astestos. The asbestos was associated almost entirely with the insu?ation on two plant systems (the steam generators and the helium circuotors), and contained 10 - 40% chrysotile asbestos. However, none of the samples were radiologically contaminated. Thus, asbestos is not exoected to present a major problem for decommissioning. All activities involving asbestos will be conducted in accordance with OSHA (29 CFR 1910 and 1926) and EPA regulations (40 CFR 61, Subpart H). An asbestos removal specification will be developed for the site consistent with the National Institute of Building Sciences format. All asbestos that is removed will be packaged for shipment and disposed of at an authorized disposal site. It is not expected that any of the asbestos will be contaminated , with radioactivity. However, if any contaminated asbestos is found, it will be disposed of at a commercial radioactive waste repository. It is PSC's understanding that contaminated asbestos is not regulated as a mixed waste. At the present time, the operators of the low level radwaste disposal facility near Richland, Washington, will accept asbestos for disposal. The expected costs of the asbestos removal and disposal activities associated with decommissioning will be provided in the detailed cost estimates that will be submitted in May 1991.
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ML20077B902

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ML20077B902

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