ML20198S138

BNL Technical Review Rept of Insp on 851008-25 at Zion Station

ML20198S138
Person / Time
Site:	Zion  File:ZionSolutions icon.png
Issue date:	06/05/1986
From:	Higgins J, Keck A, Siskind B BROOKHAVEN NATIONAL LABORATORY
To:	NRC
Shared Package
ML20198S116	List: ML20198S113 ML20198S138
References
NUDOCS 8606100338
Download: ML20198S138 (20)

Text

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l B OOKEAVEN RATIONAL LABORATORY TECHNICAL RIVIN REPORT BATES OF INSFECTION: OCTOBEE 8 AND OCTOBER 22-25, 1985 SITE: ZION STATION LICENSEE: 00BeORfEALTD DISON CD.

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CONTENTS P,,,a, gg 1

sUnnAmT..............................................................

2 J 1. PERSONS CONTACTED...............................................

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2. DOCUMENTS REVIEWED..............................................

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3. DESIGN CHANGE EVALUAT10N........................................

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4. QU ALITY AS SURANCE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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5. FACILITY DESIGN.................................................

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6. OBSERVATION OF ONGOING W0RK.....................................

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7. AS-BUILT CONSTRUCTION REVIEW....................................

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! 8. TEST 1NG.........................................................

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9. PROCEDURAL CONTR0LS.............................................

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10. OFFSITE D0SE....................................................

18 TABLE 1 - ESTIMATED DOES RATES AT FIVE SELECTED LOCATIONS. ......

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SmMARY On October 8,1985 a preliminary trip was made to Eton for site and facility f amiliarisation and document collection. The full review of the In-terim Radweste Storage Facility (IRSF) was conducted October 22-25, 1985. The review was performed by three representatives of Brookhaven National Labora-tory (BNL) for NRC, Region III. A representative of Region III attended both the entrance and the exit meetings. The review was conducted both at the Zion Station and at the Architect-Engineer's of fices (Stearns-Catalytic in Oak Brook, Ill.).

Commonwealth Edison Co. (CECO) has decided to construct Low Level Rad-waste Storage Facilities at each of its operating plants due to the potential interruption of offsite radwaste disposal. A common design is being built at LaSalle, Quad Cities, and Zion. This facility will be used only to store solidified radwaste already prepared for over the road shipment. The Zion IRSF will be located within the site owner-controlled area but outside of the protected area and will serve both units. The facility was built under the design change provisions of 10CFR50.59. Areas reviewed by BNL included:

design change evaluation, quality assurance, facility design, ongoing work, completed construction, procedures, and of f site dose / shielding calculations.

As a result of the review the following three overall conclusions have been drawn:

1. No significant discrepancies were identified and the IRSF should meet its design objectives.

2. No formal Quality Assurance program has been implemented for the facility (4).

3. Specifications and procedures for design and construction were generally followed but not rigorously (4).

The following open items, needing further action or information, were identified:

1. Concrete compressive test failure (7).

2. Structural bolting discrepancies (7).

3. No test procedures available for IRSF (8).

f4. Backup analyses for 50.59 evaluation not . completed (3).

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5. Frocedural and administrative guidelines for operation of IRSP not developed (9).

6. Y streaming from vent duct holes not analyzed (10).

7. Near well radiation dose due to shadowing effect in question (10).

8. Actual offsite dose to maximally exposed number of the public in question due to measurement locations not determined (10).

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The number in parent.heses indicates the paragraph of this report where the details of the open ites may be found.

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DETAILS

1. PERSONS CONTACTED

• R. Basso SCC D. Caus SCC D. Dahlen CECO P. LeBlond CECO J. Marx SCC

• H. Massin CECO N. Nicholson - NRC

+P. Reichert SCC

• J. Reiss CECO

• J. Suermann NRC R. Thompson CECO R. Welch SC" C. Willis NkC

• Present at the exit interview on 10/25/85 SCC - Stearna Catalytic Co.

Ceco - Commonwealth Edison Co.

NRC - Nuclear Regulatory Commission The reviewers also held discussions with other engineering, construction, and '

operations personnel.

2. DOCUMENTS REVIEWED Engineering Design Change Notice (EDCN) Log and File of EDCNs for the Zion IRSF.

SCC QA Surveillance, letter dtd 10/15/85 from Hyland to Basso.

Concrete test packages.

Soil test packages.

Letter P. LeBlond to J. Higgins dtd 10/4/85.

ANS1/ANS - 6.6.1 - 1979, " Calculation and Measurement of Direct and Scattered Gamma Radiation from LWR Nuclear Power Plants."

SCC offsite dose calculation package for CECO IRSF (Calc. No. 70160-13).

Eton Station Modification Package #M22-0-84-35.

< NRC Region III Inspection Report No. 50-295/85034 dtd 10/8/85.

" Storage of Low-Level Radactive Wastes at Power Reactor Sites" A report 4

to the Utility Nuclear Weste Management (UNWMG). R. Lawroski and Assoc.

and Tankee Atomic, 5/1/83.

3. DESIGN CHANGE EVALUATION

NRC Cuidance issued in Generic Letter 81-38 stated that utilities could increase onsite radweste storage under the design change provisions of 10CFR-50.59 provided the guidelines of the Generic Letter were followed. The design of the Zion IRSF was reviewed against the Generic Letter guidelines in para-graph 5 of this report. The Generic Letter also referred to I&E Circular 80-19 which provided information on preparing 50.59 evaluations for radweste 1 facilities. The 50.59 saf ety evaluation for IRSF was reviewed by BKL and is I described below. The Zion IRSF was designed and built essentially as new

. construction. The only parts of the facility that were handled under the Zion station modification procedures were the tie-ins to existing systems, namely service water, service air, and alaras. Electric power for the IRSF was brought from offsite. The station modification paperwork for the tie-ins was also reviewed.

The 50.59 safety evaluation was completed and approved using the appro-priate CECO procedures, and determined that

1. The probability of occurrence or consequence of an accident per the FSAR was not increased.

2. The possibility of an accident dif ferent from the FSAR was not  !

created.

3. The margin of safety was not reduced.

The evaluation used a check sheet format to ensure a large number of areas were considered for the above. A conclusion was reached that 10CFR100 limits would not be exceeded in the event of building collapse. However, no of ficial backup calculations and evaluations were available to justify this statement. One unofficial calculation was produced to show acceptability.

The reviewers also requested information to document that, in the event of an accident such as a radwaste cask drop from the crane, no offsite limits would be exceeded. The licensee said that they had evaluated these types of occur- ,

rences based on engineering judgement but that no written evaluations or calculations were available. Overall it was thus concluded that the 50.59 safety evaluation did not have suf ficient documentation to show that during design basis events or anticipated operational occurences, no un-reviewed safety question existed. This documentation or evaluation is called for by 10CFR50.59, Generic Letter 81-38, page 2 and Item V(c), CDC-60, and IE

,f Circular 80-18 Item 4. The licensee stated that such evaluations would be formalized, reviewed, approved and included as part of the 50.59 safety evaluation for the IRSF. This is open item No. 4.

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1 The 50.59 anfety evaluation was siso reviewed against Circular 80-18, which provided guidance for such safety evaluations for changes to radioactive wtste treatment sy st ems. The licensee's 50.59 safety evaluation on Worksheet

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i E, item 9, Issue ST6/R9, stated that the intent of Circular 80-18 is met. In addition to reviewing 50.59, Circular 80-18 identified four specific areas for review. These areas were not specifically addressed in the licensee's safety evaluation, but the licensee stated verbally that they had been considered as follows:

1. Modifications should be evaluated against the quality assurance criteria in R.G. 1.143.: The licensee stated that since the rad-waste would not be processed in this f acility Quality Assurance per R.G.1.143 Ites C.6 was not required.

2. Radiological Controls should be evaluated against R.G. 1.21 and Standard Review Plan (SRP) 11.5.: Again this was deemed not applicable since there would be no radwaste processing.

3. Potentially explosive mixtures should be evaluated against SRP 11.3.: This was not originally considered applicable but will now be reviewed per NRC open Item 295/85034-04 ,

4. System design and operation should be evaluated to assure that the radiological consequences of unexpected and uncontrolled releases of radioactivity that is stored or transferred in a waste system are a small fraction of the 10CFR100 guidelines.: This was not originally formalized and is considered open item f 4 as described in the above paragraph.

4. QUALITY ASSURANCE NRC has a number of guidelines recommending appropriate quality assurance for structures, systems, and components of a nuclear power plant that are not strictly safety related. 10CFR50 Appendix A GDC-1 also calls for quality l

i standards commensurate with the importance of the safety functions to be per-i formed. Regarding radwaste specifically:

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1. IE Circular 80-18 states that prior to modification of radwaste systems the modifications should be evaluated against the seismic, quality group and quality assurance criteria in Regulatory Guide 1.143.

2. 'I&E Inspection Procedure 65051 directs NRC inspectors to ascertain whether QA has been established per position C.6 of R.C. 1.143.

3. R.G.1.143, " Design Guidance for Radioactive Waste Management Sys-

, tess, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants," Position C 6 " Quality assurance for Rad-waste Management Systems' providea a limited scope QA program for these systems which includes: quality standards, independent de-

1. sign and procurement verification, survey of suppliers, storage control, inspection, states control, nonconformance control and record keeping.

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Discussions with Ceco and their Architect-Engineer, Stearne Catalytic Co. (SCC) indicated that there was no formal quality assurance program estat-l 11shed for the design and construction of the IRSF. The utility nuclear waste management group, of which CECO is a member, determined that the IRSF is not a safety class building; theref ore, no Quality Assurance was deemed necessary. ,

They further stated that, since all material was prepared for over the road I shipment and so further processing of radweste would take place in the IRSF, that R.C. 1.143 did not apply. As a result an evaluation was made by RNL as to what controle were actually utilised for design and construction. For de-j sign SCC has generally utilized the Project Procedures Manual and the Nuclear 1

Project Design Procedures (NPDP) Manual. In areas where they determined that compliance with these manuals was not appropriate they have made individual judgements not to follow them.

It was also noted that paragraph 5.03.2 of the SCC Project Proceduren Manual for Zion states that design control and document approval for non-safety related documents will be in accordance with the Nuclear Project Design Procedures Manual except for seven listed areas. As mentioned sove for the IRSF, additional areas of the NPDPs beyond those seven listed, were not followed. The judgements and the areas not followed are not documented, however, the various design controls actually utilized appeared appropriate. Areas reviewed by BNL included standards utilized, surveillances performed, drawing revision and control, change control, and calculations.

At the site actual construction was reviewed, as discussed in Sections 6

& 7 of this report, to determine the degree of conformance to specifications.

Areas reviewed included concrete work, structural steel and bolting, electri-cal, overhead crane, piping, soil, fire detection, and HVAC. Overall it was in found that construction was completed professionally and for the most part, accordance with specifications despite the lack of a QA program. However, it was noted both in the design area and the construction area that specifications and procedures were not rigorously followed floor as would be bolting in the required by a QA program. Some examples are given here:

control room and auxiliary equipment room not per drawings, crane festoon rail bolting not per drawing, holes burned in structural steel, dimensions en crane rail not per drawing, concrete HVAC foundation dimensionr, not per EDCN, concrete slump tests out of specification, concrete compressive test low, concrete tests not always done at 3 days as required, EDCNs not handled per procedures, different routing sheets used for vendor prints, not all calculations were controlled, concrete surf ace voids not filled in per specification and then painted over, the latest version of the Uniform Building Code was not used per specification, and incorrect electrical breaker with incorrect rating installed. While these items are examples of program inadequacies none appeared significant and therefore, should not compromise the ability of the IRSF to perform its design function.

5. FACILITY DESIGN In Generic Letter 81-38 the NRC has provided guidance regarding low-level radioactive waste (LLRW) storage facilities at power reactors. It is noted in ,

81-38 that the safety of a proposal for an increase in the LLRW storage capa-city must be evaluated by the licensee under the provisions of 10 CFR 50.59.

The NRC will judge the adequacy of the 50.59 evaluation based on the compli-ance with the 81-38 guidance, which is to be used in the design, construction, and operation of the storage facility. The following review of the Interim Radwaste Storage Facility (IRSF) design at the Commonwealth Edison Company I- _ _. __ _ . _ _ _ _ _ _ _ _ _ __. _ _ _ _ _ . _ _.

6-(Ceco), Eton Nuclear Plant is based on the applicable sections of Generic Letter 81-38. Procedural controle necessary to implement this guidance are discussed in section 9. An asterith (*) is used in the paragraphs below to indicate further discussion in Section 9.

5.1 Generally Applicable Guidance The following discussion is organised in the same manner asSection III of Generic Letter 81-38 but with different numbering.

5.1.1 Quantity of Radioactive Material This paragraph of the Generic Letter calls for limiting the quantity of material stored in the f acility to ensure of fsite dose within 40CFR190 limits

(= 1 area / year), onsite dose limits within 10CFR20 limits and also includes the ALARA principle. Zion and SCC performed dose calculations to ensure that onsite and offsite limits would not be exceeded. The dose calculations were reviewed in detail by BNL as discussed in paragraph 10 of this report. In order to ensure that the material actually stored in the IRSF does not go be-yond that assumed in the analyses, procedural controls will, be required *.

The design and construction of the IRSF was reviewed and it was determined that ALARA was appropriately considered.

5.1.2 Waste Container and Waste Form Compatibility of the container materials with the waste forms and with environmental conditions external to the containers in accordance with the 8138 guidance is necessary to prevent significant container corrosion.

According to J. Reiss of CECO, cement-solidified LLRW in steel liners will be stored in the Zion IRSF. Compatibility of the steel container material with the cement solidified waste form may be optimized by maintaining alkaline conditions in the cement binder material and by avoiding free liquid in the final waste form; control of free liquid and pH are part of the Process Control Program (PCP) established for the power plant's solid radioactive waste management system. Compatibility of the container material with the IRSF environment may be optimized by maintaining the relative humidity in the IRSF as low as possible throughout the year. According to the Project Plan, the storage area / truck bay ventilation system shall maintain the sp+ ace temperature at +50'F when it is 10*F outside, and at a maximum of 110*F when it is +95'F outside. It shall consist of a supply fan, exhaust relief opening, distribution system, electric duct heaters, and control system. The storage area / truck bay ventilation system described in the Project Plan should be capable of heating and drying incoming air as needed to prevent condensation and in general, maintain a low relative humidity. Procedural controls for the PCP and for operation of the ventilation system to optimize compatibility of the container material with the waste form and with the storage f acility environment, respectively, are needed*.

Gas generation from stored LLRW is an open item (295/85034-04; 304/85036-

04) from the routine safety inspection conducted by Mr. P. C. Lovendale and Ms. N. A. Nicholson of the Region III office on September 17-19 and 26,1985.

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_7 i According to the IRSF Project Plan, the overhead crane will include TV i

cameras, lighting, and a position indexing system to facilitate a program of periodic visual inspection of container integrity. Procedural controls are 1

I also required *. i 5.1.3 Security According to the Project Plan, the f acility shall have a lock on all ac-cess doors. Since the IRSF st Zion is located outside the security fence, it shall also be surrounded with a locked, chain-link fence. Security procedures are required *.

5.1.4 Low-Level Dry Waste and Solidified Waste Storage It was noted by J. Reiss of CECO that no low-level dry waste will be stored in the IRSF.

5.1.4.1 Release Pathways, Surveillance, and Decontamination of Containers Potential radionuclide release pathways shall be monitored, according to the 81-38 guidance. Joe Reiss of CECO noted that isokinetic sampling of the air in the storage area was planned by using aspirating collection devices mounted at an angle in holes through the wall. See Section 9, below.

Surveillance programs for detecting failure of container integrity and measuring releases to the environment are required *.

Procedures for decontamination of containers before storage are needed*.

5.1.4.2 Collection and Sampling of Liquid Drainage According to the IRSF Project Plan, floor drains shall direct incidental excess water or condensation to sumps servicing the truck bay and storage areas. The rectangular stainless steel construction of the sumps is intended to prevent leakage of radioactive fluids into the soil. The sumps shall be designed to allow for pumping to portable tanks for transportation to a pro-cessing location. See Section 9, below for procedural considerations *.

5.1.4.3 Waste Stored in Outside Areas Not applicable.

5.1.4.4 Corrosion From External Environment The only applicable part of this 81-38 guidance is the following: Stor-age containers should be raised off storage pads where water accumulation can be expected to cause external corrosion and possible degradation of container integrity. This is also an open item from the routine inspection of September 17-19 and 26,1985 (295/85034-06; 304/85036-06).

5.1.4.5 Total Curie Limits for Storage According to the IRSF General Design Critera, shielding thicknesses for the IRSF were determined based on a solid 5 Rem /hr array of containers.

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- B-However, it is also~ noted in the General Design Criterie that some administra-tive control over placement and specific higher activity containers um7 be re-quired to achieve the radiation sone criteria *.

5.1.4.6 Inventory Records Use of a peg-board for the purpose of inventory records was mentioned by Joe Reiss of CECO. (See Section 9, below*.)

5.2 Wet Radioactive Waste Storage According to CECO, wet radioactive waste will not be stored in the IRSF.

Therefore,Section IV of the 81-38 guidance document is not applicable.

5.3 Solidified Radioactive Waste Storage The following discussion is organized in the same order ac Section V of Generic Letter 81-38.

5.3.1 Definition of Solidified Radwaste Cement-solidified LLRW meets the definition of solidified radwaste as given in Paragraph V (a) of the 81-38 guidance document.

5.3.2 Container Protection and Reprocessing Requirements See Section 9, below*.

5.3.3 Uncontrolled Released Due to Handling Transportation on Storage Since the containers intended to be used for storage in the IRSF are also intended for use in transporting LLRW to the disposal site, they should meet Certifica-the various NRC and DOT handling and transportation requirements.

these re-tion vendor should be obtained from the container to insure that l

quirements are met.

l 5.3.4' Applicable Design Objectives and Criteria i

j Restricted areas were discussed in Section 5.1.3, above and accountabil-ity was discussed in Section 5.1.4.6, above. According to CECO, dewatered re-Corrosion of the container by sins or sludges will not be stored in the IRSF.

the waste was considered in Section 5.1.2, above.

There is no provision for additional reprocessing or repackaging due to container f ailure. CECO maintains that such provision is unnecessary.

5.4 Uniform Building Code Section 1.1.2 of the Zion IRSF General Design Criteria states that the l

latest version of all design documents shall be used. Among the design docu-ments listed is the Uniform Building Code. Upon review it was found that i

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design was done to the 1979 Revision vice the latest which is 1982. Stearns-Catalytic (SCC) agreed to review the 1982 Revision to determine any differ-ences as they apply to the IRSF. This review was completed during the week of the onsite review. SCC determined that for the IRSF the 1979 and the 1982 re-visions were equivalent.

6. OBSERVATION OF ONGOING WORK At the time of the IRSF review the f acility was approximately 85% com-plete. There was not extensive work in progress but the work being performed was observed. Work actually observed included concrete finishing, painting, and roof insulation work. Work techniques, methods, and quality appeared ap-propriate for the facility. Surface finishing work being performed on the ex-terior walls of the main building was being performed in accordance with the requirements of the Reinforced Concrete Work specification 70160-2700-05.

7. AS-BUILT CONSTRUCTION REVIEW 7.1 Discussion Various completed portions of the IRSF were reviewed and compared with specifications, procedures and drawings to determine their acceptability.

Areas reviewed included: concrete work, structural steel and bolting, the electrical overhead crane, service air and water piping, soil, HVAC ducts, and the grounding system.

7.2 Structural Steel and Mechanical Equipment A number of structural steel and mechanical areas were reviewed including the overhead crane, crane rails, festoon rails, steel columns and girders, HVAC ductwork, and base plates. The main documents used for guidance were SCC Specification for Fabrication and Erection of Structural Steel (No. 701601800-7), structural drawings, equipment drawings, and the IFSF Specification (No.

El-70160-2000-01 ) . Installed steel and equipment was checked for conformance with the various documents and for overall good construction techniques. With the exception of the below items no discrepancies were identified.

l 1. Fage 6 of the Structural Steel Spec. item 4.02.7 states that no holes l

will be made or enlarged by burning. Contrary to this a number of bolt holes were noted to have been burned in the crane festoon rail, which carries the crane cabling.

2. The structural bolting of the festoon rail to the overhead girders was I not per drawing in that: shims were added; bolt heads, shims and nuts were not flush or level with steel; and all bolts were not double-nutted. 1

3. The one inch dimension on the drawing for the supports for the crane retrieval rail was only 3/8" installed. This did not appear to affect the function of the rail or support. ,

4. Inspection of the control room / equipment room wing revealed that the four f roof support columns were not properly bolted to the floor slab as

required by drawing 5-1136. Two columne had nuts missing due to the embedded bolts being too short. The rest of the embedded bolte had less than full thread engagement with their respective outs.

The bolting items above (1, 2, and 4) constitute Open Item No. 2.

7.3 Soil Excavation Actual excavation and ccupaction was completed some time ago. SCC Speci-fication No. 70160-2100-06, Excavation, Filling and Backfilling, item 7.013 )

requires compaction of backfill to a minimum of 95% relative density. Test records were examined to determine if actual compaction had achieved the 55%

, density. All test results showed acceptable compaction.

7.4 Concrete Actual concrete work and concrete test records were examined for com-pliance with Specification No. 70160-1700-05, " Reinforced Concrete Work" and with drawings. The following areas were reviewed: concrete slump tests; concrete compressive tests at 3, 7 and 28 days; proper rebar installed; con-crete surface finish; and concrete thicknesses in wall and ceiling. The con-crete thicknesses were independently measured since they were an important input to the shielding calculations. The lower portion of the wall was at least 30" thick, the upper portion of the wall was 15", and the ceiling alter-nated between 12" and 15" thick. These were all as specified and as used in the calculations. All other areas reviewed were acceptable, except for the following discrepancies:

1. Concrete slump test results exceeded the maximum spec in a number of cases. Generally, however, the required strength was still achieved.

2. Concrete tests at 3 days were not always met, but the 7 day and 28 day tests were generally performed acceptably.

3. For two test cylinders (both in Test #56) the 28 day concrete compressive

! strengths came out below the 3500 psi minimum. This is open iten No. 1.

4. Concrete surface finish was not always repaired to remove any voids greater than 1/4" as required. A number of these were then painted over without repair.

5. The concrete base mat for the HVAC unit was originally specified on Draw-ing B-1129 to be 10'6" x 10' . EDCN No. 09 changed this to 10'6" x 11' .

The actually installed base mat was measured in place and found to be 10'6" x 10'6": and no further change notices were issued. This discre-pancy should not impact any function of the equipment or structure.

7.5 Electrical Because electrical panels and equipment were not installed yet, inspection of the grounding system consisted of verifying the requirements of the IRSF General Design Criteria, section 3.4.3 to determine if grounding cables had been installed to the equipment areas as shown on drawing 22E-0-6502. The grounding cable installation was satisfactory.

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The project plan requires a 480 volt, 70 amp welding receptacle to be provided, whereas electrical drawing 22E-0-650 shows a 400 volt, 60 amp bresker supplying the welding receptacle. Either should suffice.

7.6 Sumps and Piping Inspection of the embedded drain and sump system of the main building to drawing M805 indicated a satisf actory installation.

l A review of the pneumatic test package of the service air piping to the IRSF indicated that the installation was satisf actorily inspected by the plant ,

OC group. Test gauge calibration check sheets indicated that the test gauge had received a calibration check prior to and af ter the test.

8. TESTING ,

A number of the systems, equipments, and instrumentation designed for and installed in the IRSF requires final post-installation testing to ensure that they are functionally ready to operate. Although the facility was 85%

construction complete during the review, no test procedures were available for

• review. Purther, there was no outline or list of precisely which test procedures would be written, approved, and implemented. At the minimum there should be documented test procedures for the HVAC system, the overhead crane, the fire detection system, and the sump alarms. This is an open item (#10).

9. PROCEDURAL CONTROLS 4 In this section, a review is given of the procedural controls which may be necessary to implement the guidance in Generic Letter 81-38 at the IRSF at CECO's Zion Power Plant. This section is organized in the same manner as the review of facility design in Section 5, above. The needed procedural and administrative controls are collectively designated open item f 5.

9.1 Generally Applicable Guidance 9.1.1 Quantity of Radioactive Material The dose analysis assumes containers of radwaste which are 5 Rem /hr on contact. Some containers are periodically generated for storage in the IRSF which exceed this 5 Rem /hr. In this case certain steps must be taken (shield-ing, placement, etc.) to ensure that the dose calculations are not invali-dated. Procedural controls are needed to ensure that the needed steps are determined and implemented.

9.1.2 Waste Container and Waste Form .

The procedural controls for the Process Control Program (PCP) and for operation of the ventilation system to optimize compatibility of the container material with the waste form and with the IRSF environment have not been made available. Procedures for operation of the ventilation system should assure that incoming and/or recirculating air will be heated and dried as needed to avoid condensation and high humidity. This is of particular concern since the

. . l automatic temperature control setpoint will not prevent condensation during all seasons and onsite tours during this inspection noted heavy condensation inside the IRSF.

The PCP available from the vendor of the radwaste solidificaton process and reportedly includes provision for control of pH in a narrow basic range as well as for elimination of free liquid from the final solidified product.  ;

This was not available for review. At present, procedural controls for the 1 PCP and for the operation of the ventilation system remain open items.

As noted in Section 5.1.2, above, gas generation from stored LLRW is an open item (295/85034-04; 304/85036-04) from the routine safety inspection of September 17-19 and 26,1985.

The development of a container inspection program to assure container in-tegrity is also an open item from the routine safety inspection of September 17-19 and 26, 1985. (Open item 295/85034-05; 304/85036-05).

9.1.3 Security The development of security procedures remains an open ites.

9.1.4 Low-Level Dry Waste and Solidified Waste Storage 9.1.4.1 Release Pathways. Surveillance, and Decontamination of Containers The isokinetic sampling of the storage area air mentioned above in See-tion 5.1.4.1 responds in part to an open item (295/85034-02; 304/85036-02) from the September 17-19 and 26,1985, safety inspection. Regarding this open item, CECO agreed to establish an air and direct radiation monitoring program for the storage area and occupied areas'.

A surveillance program incorporating adequate methods for detecting fail-f ure of container integrity is an open item from the routine safety inspection of September 17-19 and 26,1985. (open Item 295/85034-05; 304/85036-05).

Since containers of LLRW are presently being shipped from the Zion Plant for disposal, the containers are presumably decontaminated to below the limits specified in 10 CFR 20.202, 20.205, and 49 CFR 173.397 and thus are in accord with the 81-38 guidance. The specific procedural controls for such decontani-nation prior to storage are not available, however, and thus remain open.

9.1.4.2 Collection and Sampling of Liquid Drainage Monitoring of water runoff collected in the sumps is an open ites from the routine safety inspection of September 17-19 and 26,1985 (open Item 295/

85034-03; 304/85036-03).

9.1.4.3 Monte Stored in outs ;a Areas Not Applicable.

9.1.4.4 Corrosion from External Environment The procedure as well as the design for raising the storage containers off the floor is an open item. This is also an open ites from the routine safety inspection of Septester 17-19 and 26,1985, in response to which the licensee agreed to consider placing the first layer of containers on a metal  !

grating to minimize condensate collection. (0 pen Item 295/85034-06; 304/ I 85036-06) 9.1.4.5 Total curie Limits for Storage I

As mentioned in Section 5.1.4.5, above, the shielding thicknesses for the IRSF are based on a solid array of 5R/hr containers and, therefore, adminis-trative controls on >5R/hr containers will be necessary. The description of administrative controls on >5R/hr containers remains an open ites.

9.1.4.6 Inventory Records Procedures for the the maintenance of inventory records remains an open item.

9.2 Wet Radioactive Waste Storage Not applicable.

9.3 Solidified Radioactive Waste Storage 9.3.1 Definition of Solidified Radwaste See Section 9.3.2, which follows.

9.3.2 Container Protection and Reprocessing Requirements The specification that only cement-solidified radwaste not requiring further processing will be stored in the IRSF does not appear to be in the General Design Criteria. J. Reiss of CECO stated that such a specification vill be issued by Station Nuclear Engineering, but until then, the specifica-tion of the type of stored waste remains an open item.

9.3.3 Uncontrolled Releases due to Handling, Transportation on Storage Procedures should be developed and implemented for early detection, pre-vention, and mitigation of accidents such as fire, cask drops, etc.

9.3.4 Applicable Design Objectives and Criteria Restricted areas and accountability were considered above in Section 9.1.3 and 9.1.4.6, respectively.

Procedural controls, such as the PCP, for avoiding corrosion of the con-tainer by the waste were considered in Section 9.1.2, above.

As noted in Section 5.3.4, above, there is no provision for additional reprocessing or replacement due to container failure. CECO maintains that >

such provision is unnecessary.

Finally, procedures must be developed to address normal operation of the IRSF to include both placement of radweste in the facility and removal from the facility.

10. OFFSITE AND ONSITE DOSE ASSESSMENT 10.1 Introduction Stearns Catalytic has performed onsite and offsite dose calculations of direct and skyshine gasma radiation from an interim radwaste storage facility at the Zion nuclear plant. The f3 acility is expected to accumulate twenty four 170 ft3 liners and twelve 85 ft liners per year for up to five years. The nominal contact dose of the liners is 5 Rem /hr. The facility is shielded with vertical concrete walls 30 inches thick to a height of 36 feet. Above this height are 15 inch thick concrete walls reaching to a concrete roof at a height of 49 feet. The roof thickness varies between 12 and 15 inches. The south vertical wall has penetrations for five intake ducts and one exhaust duct centered approximately 2 1/2 feet below the roof level. The intake pene-trations are each 16 inches high x 40 inches wide, while the exhaust penetra-tion is 30 inches high x 60 inches wide. The skyshine dose calculations have been made wigh both the Monte Carlo code, SKYSHINE II, and the single scatter code code, G , which uses the Klein-Nishina formula and corrects for multiple scatter by applying buildup factors. Both direct and skyshine dose rates have been calculated as a function of the distance from the facility wall.

10.2 Calculational Method and Results The direct gamma dose rate has been calculated in a straight forward manner using source buildup factors and mass attenuation coefficients. Fct of fsite dose calculations a 17-group source spectrum was reduced to an equiva-lent 4-group spectrum. The source isotopic mix ' chosen was identical to that used in the UNWMG study and is a representative composition of wastes from both PERs and BWRs. The skyshine calculations were performed with programs installed and controgled by the service bureau of the University Computing Company. Both the G and SKYSHINE II codes have been used in the past for reference calculations of scattered radiation in an ANSI /ANS standard.

Several simplifications were made in the calculations regarding effec-tive source strengths, combination of buildup factors for different media, angle of incidence of the radiation on the ceiling, etc. The effect of these simplifications on the expected accuracy of the results has been discussed in Section III of calculation 70160-13 and Section IV of Calculation 70160-15.

The effect of the duct penetrations on the vertical wall of the facility has not been calculated, although rough estimaten were provided in telephone conversations with representatives of Consonwealth Edison Company and Sterns Catalytic. The largest uncertainties in the calculations that have been

identified by the licensee are (1) a factor of two in the non-conservative direction associated with the representation of the source as a slab source of 5 Rem /hr dose rate, and (11) a 50% -200% conservatism, depending on samma energy, due to the multiplicative contination of buildup f actors for the source self-shielding and the roof shielding. In the opinion of the licensee the overall impact of the conservatisms and non-conservatisms is a net reasonably conservative result for the skyshine calculation; and a reasonable estimate of the direct dose provided a margin of about 50% of the calculated value is maintained below the acceptance criterf s.

10.3 Discussion 10.3.1 Uncertainties,in Theoretical Modeling A major source of uncertainty that was not identified nor accounted for in the licensee dose calculations is that associated with the mathemat-ical and physical modeling of the complex skyshine phenomenon. The significance of this uncertainty is readily discernable in Figure 3 of cal-culation7g160-13,whichshowsthatevenwithconsistentinputandsource modeling G and SKYSHINE II predictions differ by as much as a factor of three. A review of the results of the reference calculations presented in ANSI /ANS Standard 6.6.1-1979 for simple problems with well defined inputs also indicates a spread of approximately a factor of two for the various codes used. We, therefore, regard the results of the licensee calculations i as "best estimates," and conclude that the doses calculated by the licensee need to be multiplied by a factor of two to account for uncertainties in the theoretical modeling.

10.3.2 Effect of Penetrations for Inlet and Exhaust Ducts Dose calculations by the licensee do not account for unshielded gamma rays escaping through the penetrations for inlet and exhaust ducts. In re-sponse to our concern regarding this effect, the licensee indicated that the ef fects would be small based on rough estimates made using extremely simplistic models. We believe that a proper estimation of this effect can only be made using a much more detailed calculation that acccunts for, among other things, the directionality of the unshielded gamma rays. We also believe that the contributions of these unshielded gamma rays to the dose at vulnerable locations (e.g., at locations in direct line of the l

ducts such as the Turbine Building) may be quite significant. In the ab-sence of detailed calculations we are unable to arrive at a quantitative evaluation of this effect.

10.3.3 Effect of Wall Shadowing on the Near Building Dose The licensee estimates that immediately adjacent to the building the i skyshine dose rate will be reduced by a factor of two compared to that at 20 feet due to building shielding of the receiver from the skyg This esti-mate is based on the prediction trends at small distances of G , which mod-els vall shadowing, and SKYSHINE II, which does not (Fig. 3, Calculation 70160-13). Although the physical basis of the shgdowing effect is plausi-ble, due to the substantial differences between G and SKYSHINE II it is l

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3 not evident that the difference between C and BKYSHINE 11 predictions at small distances is due to the shadowing. In the absence of a demonstration of the magnitude of the wall shadowing ef fect, a conservative estimate of the near building dose rate should not take credit for this effect.

10.3.4 Location of the Maximally Exposed Individual EPA Regulations, 40CFR190.10, state that operations from the uranium fuel cycle shall be conducted so that the annual dose does not exceed 25 area whole body to any member of the public. NUREG-0133 discusses methods for calcula-tion of these and other offsite doses and refers to the utilities' Offsite Dose Calculation Manual (ODCM) as containing the methodology and parameters for these calculations. Generic Letter 81-38 states that offsite doses from onsite radweste storage must be sufficiently low to account for other uranium fuel cycle sources (e.g. it states that an additional dose of ;[ 1 arem per year is not likely to cause the limits of 40CFR190 to be exceeded). Review of various documents and discussion with NRC personnel lef t the method of deter-mination of the maximally exposed individual to be used in these calculations uncertain. It was not clear whether one should use the site fence, the site boundary, the Lake, the nearest business, or the nearest permanent residence.

It was also not clear if, or how residence time, such as number of business shif ts, should be taken into account. The approved Zion ODCM was not available for this inspection but ODCM Table 7.2-4, Rev. 2 entitled, " Critical Ranges" was given to the reviewers. This table tabulates the " nearest resi-dent" for each of the 16 cardinal compass directions from Zion. The closest value was 1609 meters in three directions: South, West, and West North West.

the reviewers requested documentation as to where onsite these distances were measured from and precisely to what " resident" offsite they were measured.

The licensee was unable to provide this documentation. This is a concern since this table is used for many routine offsite exposure calculations, not just a new design of the IRSF. In response to the questions, the licensee stated verbally that these distances were measured from the center of the two containments and to the following nearest residents or maximally exposed indi-viduals: Burgess-Anderson-Tate (WNW), to a bakery (W), and to the Illinois State Park (S). The licensee's calculation for dose from the IRSF also uses a

the same 1609 meters to determine offsite exposure. Independent measurements and tours of the Zion facility environs by BNL identified some discrepancies

' with this information. Namely, in the WNW direction, Burgess-Anderson-Take is

! on the order of 3500 feet from the center of containment and from the IRSF; and in the westerly direction the American Air Filter Company is closer than the bakery, has both a day and a night shif t, and is about 3000-3500 feet from the containments and the IRSF. Additionally, while there are no residences or permanent businesses situated in the following locations, the nearest land site boundary to the IRSF is 1421 feet and the Lake Michigan shoreline is only 240 feet from the IRSF. Fisherman were observed in the areas offshore of the facility.

In the light of the above facts, it is concluded that there is consider-able uncertainty as to where the dose to the maximally exposed individual off-site should be measured. As a result it can not be determined if the limits of 40CFR190 have been met for the IRSF. Additionally, there appears to be

! some uncertainty regarding offsite dose measurements for the site in general.

This is open item #8.

10.3.5 Dose mates for Selected Onsite and offsite Locaticne We have selected five representative locations (onsite and offsite) which address the concerne stated above. Table-1 presents the dose rates at the selected locations based on (1) the Stearns Catalytic analysis, and (2) the BNL estimate which is the SKYSHINE II plus direct dose rates (each component) increased by a factor of two for modeling uncertainties, as recommended in the above sections. Neither set of estimates accounts for the unshielded gamma rays escaping through the ducts.

10.3.6 Open Items

1. A detailed calculation of the effect of the unshielded gamma rays escaping through the inlet and exhaust ducts has not been made (Item f6). .

. 2. In order to take credit for the suggested conservatism in the SKYSHINE II near-building calculations, the effect of wall shadowing on the near-building dose should be quantified (Item #7).

3. The location of the maximally exposed of fsite individual has not been clearly determined either for IRSF calculations or other offsite dose calculations (Item #8).

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j TABLE-1 ESTIMATED DOSE RATES AT FIVE SELECTED 14 CATIONS Distance From Estimated Revised Location Facility Wall (ft) Dose Rate

• Dose Rate *t Adjacent to Facility 0 0.57 area /hr 2.06 area /hr Lake Michigan Shore Line 240 2.6x103 mrea/yr 5.2x103 area /yr Nearest Site 1421 50 aren/yr 100 arem/yr Boundary Nearest Offsite Business =3500 0.2 ares /yr 0.4 mrem /yr Nearest Resident 5280 -0.1 aren/yr -0.2 area /yr

• Contribution of gamma rays escaping through ducts not included.

t SKYSHINE II and direct dose rate each increased by a factor of two to secount for modeling uncertainties.

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