ML111100645

Gpu Nuclear, TMI 2 Special Orders and Agreements

ML111100645
Person / Time
Site:	Three Mile Island
Issue date:	04/20/2011
From:	Office of Nuclear Reactor Regulation
To:
Bamford, Peter J., NRR/DORL 415-2833
Shared Package
ML111100594	List: E910-11-007, Gpu Nuclear, TMI 2 Defueling Completion Report ML111100576 ML111100595 ML111100645
References
Download: ML111100645 (451)

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2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 INDEX SPECIAL ORDERS AND AGREEMENT MANUAL SUBJECT 10 CFR 61.55 Exemption Exemption from 10 CFR 50.36(3) TMI-2 compliance with 10 CFR 50.49 Exemption from 10 CFR 50.55a

• Code Safety Valves

• lSI Testing Relief from Mini Decay Heat Removal System Surveillance Requirements March 24, 1987 Exemption from 10 CFR 50, Appendix A, Criterion 2, 50, 51, 57 and Approval of Alternate Design to Criterion 55 (NOTE: This exemption has been superceded by the Seismic Design SER, <<4430-7322-85-1)

Exemption from 10 CFR 50, Appendix J Exemption from 10 CFR 50, Appendix R Exemption from 10 CFR 50.71(e) Variance request from 10 CFR 61 Exemption from 10 CFR 61 10 CFR 71 requirements for the SN-l shipping cask Exemption from 10 CFR 100, Appendix A Reactor Coolant Purification System Ion Exchanger Hastes Partial Exemption from the Requirement of 10 CFR 50.54(a) Exemption from 10 CFR 50, Appendix A General Design Criteria 34 and 37 Core Accountability Exemption 0432d/0186d 18 19 20 21 22 23 INDEX SPECIAL ORDERS AND AGREEMENT MANUAL SUBJECT Exemption from 10 CFR 50.61 Preservation of Records City of Lancaster Agreement March 24, 1987 Order -Submerged Demineralizer System (SDS) Exemption from 10 CFR 50, Appendix A, GDC 17 & 19 Exemption from 10 CFR 171, Annual Fee Requirements 0432d/0186d

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\. \. UNITED STATU NUCLEAR REGULATORY COMMISSION

.... INGTON.D.

c._ Mr. R. C. Arnold 6PU Nuclear Corporation P. O. Box 480 Middletown.

PA 17057

Dear Mr. Arnold:

IOVI1-.. This letter is in response to your request dated October 27, 1983 for an interpretation of 10 CFR Part 20.311(d)(1).

You indicated that 6PU Nuclear Corporation has in the past received variances from burial site states for the disposal of radioactive wastes and would expect that other variances may be approved in the future. You asked if the Nuclear Regulatory Commission (NRC) would allow the variances approved by individual disposal site states without formal NRC review and approval.

In the low-level waste management regulation, Section 10 CFR 61.58 allows the consideration of exemptions to the waste classification and waste form requirements provided that the performance objectives in the rule are satisfied.

In the case where an Agreement State has the regulatory authority over a disposal site and applies the waste classification and form system of 10 CFR 61 on a compatibility basis, an authorization would be needed from the Agreement State to dispose of wastes in other concentrations or waste form characteristics.

In addition.

since you are an NRC licensee, approval would also be required from the NRC since the action would result in a change to license commitments to adhere to 10 CFR Part 20 which includes, after December 27. 1983, the requirements of Section 20.311. Specific variances from 20.311(d)(1), (2) and (3) would appear to be needed since waste would not be either Class A. B, or C. In addition, a way of tdentifying the wastes on the .. nifest should be provided that is acceptable to NRC. the states and the disposal site licensee.

If you intend to request such variances to Section 20.311, a copy of your proposal to the state and justification for the variance requested should be provided.

In addition, you should provide docUleJtation of state approval.

C'. --"Sce, _

$ PAGE 2. OF 4--2 -,. , 1f 1°U have allY further questions "garcllng this attar. please contact Dr. Bernard J. SlIY der of the Th"e Mile lSland ,rogrlll Office at 301-492-7761.

Sincerely, cJ.fb.

Office of Nuclear Material Safety and Safeguards

,." r r r f ( 3 GPU Nucte. CorporatIon Post Office Box 480 Rout ..... 1South -]Nuclear \ Middletown.

Pemaylvania 17057-0191 717844*7821

.... ... " . . .... Office of the Executive Director for Operations Attn: Mr. William J. Dirckl Executive Director US Nuclear Regulatory Commilsion Washington, DC 20555

Dear Sir:

October 27. 1983 4410-83-L-0246 TELEX 84*2388 Writer'. Direct Dial Number: Three Mile Island Nuclear Station, Unit 2 (TMI-2) Operating License No. DPR-73 Docket No. 50-320 Request for Interpretation of 10 CPR 20.311(d)(1) of this letter is to request an interpretation of 1 R 20.311(d)(1) for shipment of waste from THI-2. Section (\ .1) requires that a licensee "prepare all wastes so that the .. ste is classified according to Section 61.55 and meets the waste characteristic requirementl of Section 61.56 of this chapter" prior to shipment to a land disposal facility.

Currently, there are three (3) commercial land disposal facilities licensed and controlled via agreement state licenses as opposed to NRC licenses.

Per the Policy Statement 48 Fa 33376, these states have the authority to exceed the Itm1ts of 10 CPR 61 for dilpolal of radioactive materiall providtng a apecific exemption 11 liven by the state. Since these states have this authority and due to the unique nature of the wastes tenerated at TMI-2, GlUNC bereby requests an interpretation of 0 CPR 20.311(d)(1) in order to allow TMI-2 to ahip in accordance with the apecific exempt ion I acquired from the atatel involved in waste disposal.

CPUNC interprets 10 CPR 20.311(d)(1) to require paCkaltn&

classification of waites in accordance with 10 CPR 61.55 and 1.56 or as otherwise clallifiea by the agreement atate. QPUNC hal, in the past, received variance I from burial atates and 1a currently prepartng a submittal to the State of Walhtngton requestinl a variance from the burial requirement I being prepared to comply with the 'Oewly impoled 10 CPR 61. This variance will ';>>f t the burial of EPICOR liners currently atored at THI. \..-GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation

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i Mr. William J. Dircks' ... ,.. " .. 4410-83-L-0246 PAGE 4-OF 4 I l \ I ..... Additional variances may be requested from the appropriate authorities in the future 1f conditions require them. If you have any questions or desire further information, please contact Mr. J,. J. Byrne of my staff. -.. . Sincerely, lsI R. C. Arnold R. C. Arnold President RCA/JJB:RBS/jep CC: Mr. L. H. Barrett, Deputy Program Director -TMI Program Office Mr. H. R. Denton, Director -Office of Nuclear Reactor Regulation Dr. B. J. Snyder, Program Director -TMI Program Office UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. c. 20555 Docket No. 50-320 Mr. F. R. Standerfer Vice President/Director Three Mile Island Unit 2 GPU Nuclear Corporation P.O. Box 480 Middletown, PA 17057

Dear Mr. Standerfer:

October 24, 1985

Subject:

Approval of Exemption from 10 CFR 61.55 a::a &SfiUiJii5 II. "\.4\0 "l[. plS , .Mml .0 So, .5$ .: b§rl!YC! ...............

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• We have reviewed your request. dated June 25, 1985, for an exemption from the requirements of 10 CFR regarding the waste classification of TMI-2 EPICOR 11 resin liners. We have determined that an exemption from the requirements of 10 CFR 20.311 is not necessary but that an exemption from the waste classification requirements of 10 CFR 61.55 is appropriate.

Accordingly, we have granted an exemption from the requirements of 10 CFR 61.55.as described in the attached Exemption issued by the Director of Nuclear Reactor Regulation.

The grtnting of this exemption includes supplemental requirements for the waste shipment manifests required by 10 CFR 20.311. A Federal Register notice for this issuance is also enclosed.

Enclosures:

1. Exemption Sincerely, tA't.-l O. A---Bernar,d J.

Program Director Three Mile Island Program Office Office of Nuclear Reactor Regulation

2. Environmental Assessment and Notice of Finding of No Significant Environmental Impact 3. Federal Register Notices cc: T. F. Demmitt R. E. Rogan S. Levin W. H. Linton J. J. Byrne A. W. Miller Service Distribution list (see attached)

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• Mest lin. Strtlt LlftClUer,'A 17602 UNITED STATES NUCLEAR REGULATORY COMMISSION In the Hatter of Enclosure 1 GENERAL PUBLIC UTILITIES NUCLEAR CORPORATION Docket No. 50-320 (Three Mile Island Nuclear Station Unit 2) EXEMPTION

1. GPU Nuclear Corporation, Metropolitan Edison Company, Jersey Central Power and Light Company and Pennsylvania Electric Company (collectively, the licensee) are the holders of Facility Operating License No. DPR-73, which has authorized operation of the Three Mile Island Nuclear Station, Unit 2 (TMI-2) at power levels up to 2772 megawatts thenmal. The facility, which is located in Londonderry Township, Dauphin County, Pennsylvania, is a pressurized water reactor previously used for the commercial generation of electricity.

By Order for Modification of License, dated July 20, 1979. the licensee's authority to operate the facility was suspended and the licensee's authority was limited to maintenance of the facility in the present down cooling mode (44 Fed. Reg. 45271). By further Order of the Director, Office of Nuclear Reactor Regulation, dated February 11. 1980, a new set of formal license requirements was imposed to reflect the post-accident condition of the facility and to assure the continued maintenance of the current safe, stable, long-term cooling condition of the facility (45 Fed. Reg. 11292). This license provides, among other things. that it is subject to all rules, regulations and Orders of the Commission now or hereafter in effect.

• II. On October 26, 1983, General Public Utilities Nuclear Corporation (GPUNC) submitted a letter to the State of Washington requesting a variance to 10 CFR 61.55 regarding the classification of TMI-2 EPICOR II solid waste liners. This letter proposed that the EPICOR II liners be as Class A waste and, therefore, be burfed in an unsolidified and dewatered . condition.

Accordingly, SPUN proposed to increase the upper Class A limit for Sr-90 from 0.04 uCi/cc to 1.0 uCi/cc for the EPICOR II liners. On July 17. 1985, SPUN received a letter from the State of Washington 9ranting the variance provided that the following restrictive conditions are met: (1) Sr-90 concentrations are not to exceed 1 uCi/cc. (2) Wastes will comply with Class A waste requirements specified in 10 CFR 61.56. (3) Wastes are disposed of at the bottom of the trench and segregated from stable Class Band C wastes. and (4) Wastes do not contain. other nuclides listed in Tables 1 and 2 of 10 CFR 61.55 which exceed the Class A limits by themselves or giving consideration to the partial fractions rule. In order to implement this variance from 10 CFR 61.55, SPUN submitted a letter to the NRC, on June 25, 1985, requesting exemption from certain requirements of 10 CFR 20.311(b) and 20.311(d)(1), (2) (3) for classifying the TMI-2 EPICOR II liners. However, we have determined that an exemption from the requirements of 10 CFR 20.311 is not necessary but that an exemption from the waste classification requirements

• of 10 CFR 61.55 is appropriate.

• Ill. 10 CFR 20.311(b) in part states: -Wastes classified as Class A, Class B, or Class C in Section 61.55 of this chapter must be clearly identified as such in the manifest.-

10 CFR 20.311(d)(1) states: -Prepare all wastes so that the waste is classified according to Section 61.55 and meets the waste characteristics requirements in Section 61.56 of this chapter.-10 CFR 20.311(d)(2) states: -Label each package of waste to identify whether is Class A waste, Class B waste, or Class C waste in accordance with Section 61.55 of this chapter.-10 CFR 20.311(d)(3) states: -Conduct a quality control program to assure compliance with Sections 61.55 and 61.56 of this chapter; the program must include evaluation of audits.M The above regulations require the licensee to comply with the waste classification requirements of 10 CFR 61.55. Under 10 CFR 61.55, the TMI-2 liners (approximately 100 line!s total, each with 170 ft.3 of spent resin) would be classified as Class B waste. If the licensee proposes to reprocess the EPICOR liner waste to meet Class A classification under 10 CFR 61.55, there would be an increase in waste volume to be disposed of by about 6001. Compliance with the Class B conditions of 10 CFR 61.55 would require stabilization of the waste form. This would also result in substantial increases in the volume of EPICOR liner wastes to be disposed and the occupational exposure due to required increased handling of waste. We estimate that the stabilization requirements for Class B wastes would result in a volume increase of 201 to 50S for the EPICOR liners to be disposed.

Additionally, we estimate that occupational exposure resulting . -. from either the stabilization requirement of Class B form or reprocessing to meet the Class A classification condition would increase by at least a factor of two over the exposure which would result from the handling of the EPICOR liners as Class A waste. Accordingly, an exemption from the waste classification requirements of 10 CFR 61.55, which would otherw,se require the EPICOR wastes to be classified as Class B and stabilized, is appropriate as required stabilization would result in an adverse impact and 6PUN has proposed alternatives -for the handling and disposal of the EPICOR wastes. In lieu of the waste classification requirements of 10 CFR 61.55. 6PUN proposed to classify the TMI-2 EPICOR II liners in accordance with a letter submitted by GPUNC to the State of Washington on October 26. 1983, ing a variance to the requirements of 10 CFR 61.55 to allow a 1 uCi/cc limit on Sr-90 as the upper Class A limit for TMI-2 EPICOR II liners. In response to a September 11, 1981 request, the NRC staff performed an evaluation (Letter from B. Snyder, NRC, to J. Barton, Metropolitan Edison Company, dated October 22, 1981) to determine the Sr-90 concentration limit that would be acceptable for burial of an unstabilized EPICOR II liner. The staff's evaluation concluded that dewatered resin wastes with a concentration limit of 24 uCi/cc of Sr-90 would be acceptable for burial at an arid disposal site such as the Hanford site in the State of Washington provided certain restrictions on disposal were met. The acceptability of the disposal was based on pathway analyses that demonstrated that the

• perfonmance objectives in proposed 10 CFR Part 61 would be met. Disposal as provided in the State variance would meet the performance objectives in final Part 61 and all other aspects of the staff's earlier October 22, 1981 evaluation were reviewed and determined to remain valid for this current exemption request. The staff, therefore, concludes that the ltcensee's proposal for an upper Class A limit of 1.0 uCi/cc for Sr-90 is acceptable in this instant act10n and an exemption to the waste classification requirements of 10 CFR 61.55 is appropriate.

Alternatively, without the exemption, the licensee would not be able to implement the State variance from 10 CFR 61.55 resulting in a substantial increase of waste volume to be handled and transported for disposal.

Such an increase would be detrimental to the public health and safety and would both increase unnecessary exposure to radiation and consumption of burial site capacity without providing any benefit to public health and safety at the burial site. IY. Accordingly, the Commission has that, pursuant to 10 CFR 61.6. an exemption is authorized by law and will not result in undue hazard to life or property.

The Commission hereby grants an exemption the requirements of 10 CFR 61.55 as discussed in Section 111. The exemption is to the Sr-90 concentration limit of 0.04 curies per cubic meter curies per cubic centimeter) in Column 1 of Table 2 in 10 CFR 61.55 for the specific EPICOR II wastes. The wastes must be labeled and identified as Class A. Further, in order to assure that the site operator can identify * 'the special tase EPICOR II Class A wastes and meet the prescribed disposal requirements.

the licensee is hereby directed to add the following language or equivalent to the manifest required by 10 CFR 20.311: -Class A EPICOR II waste packages must be disposed of as prescribed in the attached variance.-(The requirement to attach a copy of the variance '9 the shipping papers is included in the State approval.)

It is further determined that the exemption does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. In light of this and as reflected in the Environmental Assessment and Notice of Finding of No Significant Environmental Impact prepared pursuant to 10 CFR 51.21 and 51.30 through 51.32, issued on October 3. 1985, it was concluded that the instant action will not have a significant impact on the environment and thus, an environmental impact statement need not be prepared.

Effective Date: October 24. 1985 Dated at 8ethesda, Maryland Issuance Date: October 24. 1985 FOR THE NUCLEAR REGULATORY COMMISSION Harold R. Denton. Director Office of Nuclear Reactor Regulation Enclosure 2 UNITED STATES NUCLEAR REGULATORY COt"11SSION GENERAL PUBLIC UTILITIES rmCLEAR CORPORATION*

DOCKET NO. 50-320 REVISION TO ENVIRONnENTAL ASSEssnENT AND NOTICE OF FINDING OF NO SIGNIFICANT ENVIRONMENTAL IMPACT

• On September 20, 1985, the U.S. Nuclear Regulatory Commission (the Commission) provided notice (50*F.R. 38234) of a planned issuance of an Exemption relative to the Facility Operat1ng License No. DPR-73, issued to General Pub11c Utilities Nuclear Corporation (the licensee), for operat10n of the Three M11e Island Nuclear Stat10n, Unit 2 (TMI-2), located in Londonderry Township, Dauph1n County, Pennsylvania.

Specifically, the not1ce stated that the Commission was cons1dering an exemption from certain requirements of 10 CFR 20.311(b) and 20.311(d)(1), (2) and (3) for classifying TMI-2 EPICOR 11 solid waste liners. Since the issuance of the aforementioned notice (50 F.R. 38234), the Commiss10n has determ1ned that exemption from certain requirements of 10 CFR 20.311 is unnecessary but that exempt10n from certain requirements of 10 CFR 61.55 is appropriate.

While the environmental impacts associated with the cons1dered exemption from 10 CFR 61.55 are no different from the impacts previously described (50 F.R. 38234) for exemption from 10 CFR 20.311, the Comm1ssion is nonetheless providing the follow1ng revised Environmental Assessment to correctly describe the action being considered (i.e., exemption from certain requirements of 10 eFR 61.55). -"-!,,'-'.'

....

• ENVIRONMENTAL ASSESSMENT Identification of Proposed Action: The action being considered by the Commission is an *exemption from certain requirements of 10 CFR 61.55 for classifying TMI-2 EPICOR II 'solid waste liners. Specifically 10 CFR 61.55 requires, in part, that the classification of waste for surface disposal be in accordance with the radionuclide concentration limits provided in Tables 1 and 2 of 161.55(a)(3) and (4). For Sr-gO, the concentration limit for Class A waste is 0.04 curies per cubic meter. The licensee has received a variance from the State of Washington to permit the burial, as Class A waste, of EPICOR II resin liners containing Sr-gO centrations up to 1.0 curtes per cubic meter. In order to implement this variance, the licensee requires an exemption from the requirements of 10 CFR 61.55 for classifying EPICOR II resin liners. This action do,s not involve any other exemptions and the EPICOR II resin liners will be packaged and transported tn accordance with applicable Commission and Department of Transportation regulations.

The Need for the Action: The licensee has received from the State of Washington a variance to the Class A waste criteria of 10 CFR 61.55 regarding the TMI-2 EPICOR II solid waste liners to increase the upper Class A limit for Sr-gO from 0.04 uCi/cc to 1.0 uti/cc. In order to implement this variance.

the licensee requires an exemption from 10 CFR 61.55 as discussed above. Without the variance, the waste volume for disposal would significantly increase and there would be corresponding increases in occupational exposure resulting from additional waste handling without any benefit to public health and safety at the burial site * . -----..*.. -... -.. -._-.. -..

.. .. --....

• Environmental Impacts of the Proposed Actions: The staff has evaluated the subject exemption and concluded that it will not result in significant increases in airborne radioactivity inside facility buildings or in corresponding releases to the environment.

There are also no non-radiological impacts to the environment as a result of this actton. Alternative to this Action: Since we have concluded that the environmental effects of the proposed action and exemption are negligible, any tives with equal or greater environmental impacts need not be evaluated.

Denial of this exemption would not reduce environmental impacts of plant operations and would result in the application of overly restrictive regulatory requirements when considering the unique conditions of TMI-2 * . Agencies and Persons Consulted:

The NRC staff reviewed the licensee's request and consulted with the Department of Social and Health Services, State of Washington.

Alternate Use of Resources:

This action does not involve the use of resources not previously considered in connection with the Final matic Environmental Impact Statement for THI-2 dated Harch 1981. Finding of No Significant Impact: The Commission has determined not to prepare an environmental impact statement for the subject Exemption.

Based upon the foregoing environmental assessment, we conclude that this action will not have a significant effect on the quality of the human environment.

• For further details with respect to this action see; (1) Letter to J. J. Barton, Metropolitan Edison Co., from B. J. Snyder. USNRC, Evaluation of EPICOR II liner disposal conditions.

dated October 22, 1981; (2) Letter to L. Gronemyer.

State of Washington, from B. K. Kanga, GPUNC, 10 CFR 61 Exemption.

dated October 26, 1983; (3) Letter to B. J.

from F. R. Standerfer.

GPUNC, 10 CFR 20.311 Exemption Request, dated June 25, 1985; and (4) Letter to B. K. langa, GPUNC, from J. Stohr and M. J. Elsen. State of Washington, dated July 17. 1985. The above documents are available for 1nspection at the Commission's Public Local Document Room. 1717 H Street, N.W., Washington, DC, and at the Commission's Local Public Document Room at the State Library of Pennsylvania, Government Publications Section, Education Building, wealth and Walnut Streets, Harrisburg, Pennsylvania 17126. FOR THE NUCLEAR REGULATORY COMtUSSION

-

Bernard J. Snyder, Director Three Mile Island Program Office Office of Nuclear Reactor Regulation

• Enclosure 3

I I I I ... , Docket No. 50-320 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C.1055S October 14. 1185 Docketing and Service Section Office of the Secretary of the Commission

SUBJECT:

nree IIfl. IIlaad Batt z Approval of Exemption from 10 CFR

• * * . . . Two signed originals of the Federal Register Notice identified below are enclosed for your transmittal to the Office of the Federal Register for publication.

Additional conformed copies ( ) of the Notice are enclosed for your use. o Notice of Receipt of Application for Construction Permlt(s) and Operating Ucense(s).

o Notice of Receipt of Partial Application for ConstructIon Permlt(s}

and Facility Ueense(s):

Time for Submission of Views on Antitrust Matters. . o Notice of Availability of Applicant's Environmental Report. o Notice of Proposed Issuance of Amendment to Facility Operating Ueense. o Notice of Receipt of Application for Facility Ueense(s);

Notice of Availability of Applicant's Environmental Report; and Notice of Consideration of Issuance of Facility License(s) and Notice of Opportunity for Hearing. o Notice of Availability of NRC Oraft/Final Environmental Statement.

o Notice of Umited Wort< Authorization.

o Notice of Availability of Safety Evaluation Report. o Notice of Issuance of Construction Permit(s).

o Notice of Issuance of Facility Operating License(s) or Amendment(s).

II Other: Exempt 1 on

Enclosure:

As Stated

• Ierul"'d Suder. ProgrlS1 DIrector Office of Nuclear Regulation Docket No. 50-320 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 105$5 September 27, 1985 Docketing and Service Section Office of the Secretary of the Commission "

SUBJECT:

Three Mile Island Unit 2

• Environmental Assessment and Notice of Finding of No . Significant Environmental Impact Two signed originals of the Federal Reg!!!!!:.

Notice identified below are enclosed for your transmittal to the Office of the Federal Register for publication.

Additional conformed copies ( ) of the Notice are enclosed for your use. o Notice of Receipt of Application for Construction Pennit(s) and Operating Ucense(s).

o Notice of Receipt of Partial Application for Construction Pennit(s) and Facility License(s):

Tme for Submission Of VIeWS on Antitrust Matters. D Notice of Availability of Applicant's Environmental Report. D Notice of Proposed Issuance of Amendment to Facility Operating license. D Notice of Receipt of Application for Facility License(s):

Notice of Availability of Applicant's Environmental Report; and Notice of Consideration of Issuance of Facility Ucense(s) and Notice of Opportunity for Hearing. . D Notice of Availability of NRC OraftlFinaI Environmental Statement D Notice of Urntted Work AuthorIzation.

o Notice of Availability of Safety EvaJuation Report. o Notice of lasuance of Construction Permit(s).

D Notice of Issuance of Facility Operating Llcense(s) or Amendment(s).

l!I Other: Environmental Assessment and Notice of Finding of No Significant Environmental Impact

Enclosure:

.8ernard .J. Snyder. ogram Director Office of Nuclear Reaelo Regulation As Stated .. --_._----.. _--...... ,_. -.--..... " ...... . -."

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Pennlylvanla

'7057'()1D1 178<<*7821 , "' ....... . TMI Program Office Attn: Dr. B. J. Snyder Program Director US Nuclear Regulatory Commission Washington, DC 20SSS

Dear Dr. Snyder:

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.... 11M t", . nfl'l ftIJI So &

ELEX ... *2388

• riter', Direct Dial Number: .' (717) '48-8461 441D-8.5-l-oU8 DocLlllent 10 Ju'le 25, DSS Three Mile Island Nuclear Station, lklit 2 (TMI-2)

License No. DPR-n Docket No. 50-320 10 a:R 20.311 Exemption Nuclear letter dated October 26, D83, requested from the State of Washington a variance to the Class -A" waste criteria of 10 a=R 61.SS regarding the TMI-2 EPlCOR II liners. This variance recp!St proposed an increase in the Class -A" limit for Sr-9O from 0.04 L£i/cc to 1.0 L£i/cc. The State of' WaShington forwarded this request to the tft: for their tectn1cal review. On May 8, D8S, the tft: transmitted a meroorandLrn to the State of Washington reconmending appro,val of GPU Nuclear's variance request. This recorrmendation was based on an tft: safety evaluation performed in October of D81. However, in order to this variance request, GPU Nuclear requires an exemption from the requirements of 10 aR 2O.311(b) and 20.311(d)(l), (2), and (3). SpeCifically, exemption is required from those portions of the above regulations which require the classification of wastes 1n accordance with 10 aR 61.S5. Therefore, based on the attached evaluation, GPU Nuclear 1s requesting exemption from these requirements.

GPU Nuclear Corporation

'1 a lubsldlary of the General Public Utilities Corporation

" Dr. B. J. Snyder -. Jl.ne 25. 1985 441D-85-L.-Dl38 In addit1on, enclosed for your information 1s a copy of GPU Nuclear's variance-z:equest tQ 10 CFR 61.55 which, based on discussions with the State of Nashington, GPU Nuclear expects to be approved.

Upon* receipt of the State cf.*Washington'*s approval, a copy will be forwarded to your office. .. .. FRS/RDW/eml Attact1nents Sincerely, lsi F. R. Standerfer F. R. Standerfer Vice President/Director, TMI-2 cc: Deputy Program Director -TMI Program Office. Dr. W. D. Travers ATIACtf.ENT (441D-85-L-D138)

INTRODlCTICI'!

GPU Nuclear letter 441o-83-L-0259 dated October 26, 1983, requested from the State of Washington a variance to 10 CFR 61.55 regarding the TMl-2 EPlCCR II resin liners '/ This letter proposed that the EPICOR II resin liners be categorized as Class "A" waste and, therefore, be buried in a dewatered condition.

Accordingly, GPU Nuclear proposed increaSing the upper Class "A" limit for Sr-90 from .04 uCi/cc to 1.0 uCi/cc. In order to iJlt>lement this variance request from 10 CFR 61.55, GPU Nuclear is requesting exemption from the following regulatory requirements:

o 10 CFR 20.311(b)

GPU Nuclear is requesting exemption from the portion of 20.311(b) which states, "wastes classified as Class A, Class at or Class C in Section 61.55 of this chapter must be clearly identified as such in the manifest".

o 10 CFR 20.311(d)(1)

This section states, "Prepare all wastes so that the waste is classified according to Section 61.55 and meets the waste characteristics requirements in Section 61.56 of this chapter".

GPU Nuclear is requesting exemption from this requirement for the EPICCR II liners. o 10 CFR 2O.311(d)(2)

This section states, "Label each package of waste to identify whether it is Class A waste, Class a waste, or Class C waste in accordance with Section 61.55 of this chapter".

GPU Nuclear is requesting exemption from this requirement for the EPlCCR II liners. o 10 CFR 2O.311(d)(3)

This section states, "Conduct a quality control program to assure compliance with Sections 61.55 and 61.56 of this chapter; the program must include management evaluation of audits". GPU Nuclear is requesting specific exemption from the requirement to comply with 10 CFR 61.55. The TMl-2 EPICOR II liners will comply with the requirements of 10 CFR 61.56. Reason for Exemption The above regulations, from which exemption is requested, require the licensee to comply with the waste classification requirements of 10 CFR 61.55.

ATIACI+£NT ( 441o-85-l.-Q138) lklder 10 CFR 61 * .5.5, the TMI-2 EPICCR liners would be classified as Class "8" waste and, therefore, would require stabilization, via solidification, in accordance with 10 CFR 61 * .56. However, cOl'J1)liance with the Class "8" conditions of 10 CFR 61 * .5.5 would result in an increase of burial volune and AL.ARA CORCerns because: o n,

used for miscellaneous processing and for polishing the effluent of our Sut:merged Demineralizer System (50S). These liners are sodilJ1l limited rather than curie limited. As a result, the present curie loadings on these resins cannot be increased above their current 1 UCi/cc level because these resins become chemically depleted.

Therefore, stabilization via solidification of resins at this level would result in a 30 to 40 percent increase in volune due to solidification efficiency.

o The EPICCR II resin liners have no insitu solidification capability; the resins would have to be sluiced from the EPICOR liner into another container.

The sluicing activity and volune increase from solidification would cause additional handling of the EPICOR liners. This additional handling would increase personnel exposure at both TM! and the burial site, and would increase the potential of a radioactive release accident.

Compliance with the current Class "A" conditions of 10 CFR 61 * .55 would also result in an increase of burial volll1le and ALARA concerns because: o Oue to the accident at TMI-2, there is a higher concentration of Sr-90 in the waste stream than normal. Therefore, implementation of the 10 CFR 61 * .55 Class "An limit for Sr-90, i.e., 0.04 uCi/ml, would result in apprOximately ten (10) times more waste as COI'J1)ared to the proposed Sr-90 limit of 1 UCi/cc. Alternative Methods o 10 CFR 2O.3l1(b) and 20.3l1(d)(1), (2) In lieu of the waste classification requirements of 10 CFR 61 * .55, GPU Nuclear will classify the TMI-2 EPICOR II liners in accordance with GPU Nuclear letter 441o-83-L-02.59 dated October 26, 1983. This classification will be annotated on the shipment manifest.

GPU Nuclear will comply with all other requirements of these regulations.

o 10 CFR 2O.311(d)(3)

GPU Nuclear has and will continue to conduct a quality control program for the TMI-2 EPICOR II liners. In lieu of assuring compliance with 10 CFR 61.5.5, our quality control program will assure the compliance of the subject liners with the criterion of the requested variance to 10 CFR 61.5.5.

ATIACtKNT (441Q-S5-L-oDa)

Safety Evaluation Justifying Change The NRC staff performed an evaluation in October of 1981, at the request of GPU Nuclear, to determine the Sr-90 concentration limit for an LIlstabllzed EPICCR liner that would be acceptable for burial at the Hanford site. The results of.. the tflCis evaluation show that a concentration limit of 24 t.Cilcc of Sr-90 would be "acceptable for the TMI-2 EPICCR II liners. (plJ Nuclear's variance request to 10 CFR 61.55, which proposed an upper Class -A" limit of 1.0 t.Ci/cc for Sr-90, is very conservative in catparison to the results of the NRC's safety evaluation.

Therefore, an to the waste classification requirements of 10 CFR 2O.311(b), and 20.311(d)(1), (2), and (3) will not jeopardize the health and safety of the public.

", . . .

• ENCLOSURE (4410-85-l-0138)

OPU Nucle., Corporetlon Post Office Box 480 Route441$outh Middletown.

PeMsylvlnil17057..()181 7178 .... -7621 -, ... October 26. 1983 4410-83-L-02S9 State of Washington Department of Social and Health Services Attn: Mr. Lee Gronemyer Radiation Control Section Mail Stop LF-13 Olympia. WA 98S04

Dear Sir:

TELEX Writer'. Direct Dill Number: Three Mile Island Nuclear Station. Unit 2 (TMI-2) Operating License No. DPR-73 Docket No. SO-320 10 cn 61 Exemption 1 on recent conversations between you and members of my staff. Nuclear has been informed of the State of Washington's intentio to change the license of the Hanford Disposal Site to tmplement appropriate requirements contained in 10 cn Part 61. It was also learned that this license change is intended to become effective

_*a .at. ",j ** ,

.. ,

-1pe1OJ = .. Iiiiii liUoI

.-t.no14*'ard, . liard-1'\>

r. n IoIdd.I-.d

.... lark-,..d . Vi ... -.d. a ** U1Dov-M. a.* 1ft .ook-.... ........ -M. ati. vnd-' ... Io. lld.b .. "d-,ld. .AI "rroe-Ad.

a *. u."-'ard . .... ,,-M. a *. v1l>< ***** ConDO. -** Ul . mth-T. r. ,

• tKl'.-... te

• llaoo, .-, ... 10. &lU.S* * . a *. I>>( C-,. .. te.

• 10 t ..... h POIC SIC-Ad. a *. .L..a.. , * '-'vI IV Iv 1,./ If' i ... 1,11 ... I .... by the end of this year. Although GPU Nuclear has Dot, as of yet, had the opportunity to study this change, we understand that the new license will require shipments to the disposal site to be classified in accordance with the requirements in 10 cn 61.SS and **** -'1M----a *** neet the waste characteristics requirements of 10 CFR 61. ",5 I" rhe purpose of this letter is to request, from the State of iashington, a variance to this change so that EPICOR II resin liner :ould be classified as Class "A" "aste and. therefore, be buried in I dewatered condition as is the current case. Under the current License these EPICOR II resin liners comply with condition 27j in :hat the specific activity of materials with half-lift greater than :ive (5) years is less than one (1) uCi/cc. Under 10 cn Part { ;1.S5, however. these liners would be classified as Class "B" waste ind require stability in accordance with 10 CFR Part 61.S6. :

2. Column 1 of 10 CFR Part 61.SS lists the maximum concentration

or Class "A" waste. Isotopes of interest to the TMI-2 Recovery 'rogram are Sr 90 and Cs 137. The 1tmits for these isotopes are 1.04 uCi/cc and 1 uCi/cc respectively.

For the rest of the nuclear these values are a relaxation of the current license lition 27j. However. due to THI's unusuallI high Sr 90 and Cs 137 aeio, these values are more restrictive.

Imp ementation of the more estrictive Sr 90 criteria for Utlstabilized waste (Class ",A") at TMI GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation 4 Mr. tee Gronemyer ENCLOSURE (4410-85-L-0138) 4410-83-L-02S9 would resuit .in the generation of approximately ten (10) times more waste than would be generated

\mder the current limit. '. Compliance.:with the proposed license Class "B" conditions would also result in an increase of burial volume and ALARA concerns.

EPICOR 11 liners are processing and for polishing the effluent of our Submerged Demineralizer System (SDS) and they are sodium limited rather than curie limited. As a result. the present curie limits cannot be increased above their current 1 uCi/cc level because these resins chemically deplete at this l.vel. Stabilization via solidification of resins at this level would result in a 30 to 40 percent increase in volume due to solidification efficiency.

Because the EPICOR 11 resin liners have no insitu solidification capability, the resins would have to be sluiced from the EPICOR liner into another container.

The sluicing activity and volume increase from solidification would cause additional handling and, therefore, personnel exposure at both TNI and the burial site leading to ALARA concerns along with the possibility of a radioactive release. The NRC staff performed an evaluation in October of 1981, at the request of GPU Nuclear, to determine the Sr 90 concentration limit for an \mstabilized EPICOR liner that would be acceptable for burial at the Hanford site. The results of the NRC's evaluation show that a concentration Itmit of 24 uCi/cc of Sr 90 would be acceptable for waste to be considered Class "A" waste \mder the criteria used to develop the limits in 10 CFR Part 61. A copy of the NRC's evaluation is enclosed for your information. . The limits expressed in 10 CFR 61 are for the burial of Class "A" waste at a humid site and at normal burial depths, less than three (3) meters. Provisions for exemptions from specific limits are provided for within 10 CFR Part 61 if the performance objectives can be met by consideration of options such as burial at an arid site and at a depth greater than five (5) meters. Based on the NRC's analysis, GPU Nuclear is requesting a variance to allow a 1 uCi/cc limit on Sr 90 as the upper Class "A" limit for TNI EPICOR II waste. All other Table 2, Column 1 limits would remain the same. In addition, the liners would be requested to be buried at the bottom of the disposal trench. It is our belief that this variance would be granted without any adverse effect on the health and safety of the public. GPU Nuclear believes that this variance would be in compliance with the full intent of 10 CFR Part 61. If you have any questions, please contact Mr. J. J. Byrne of my staff. BKK/JJB/jep Enclosure Sincerely, lsI J. J. Barton for B. K. Kanga Director, TNI-2 CC: Mr. L. H. Barrett, Deputy Program Director -TM1 Program Office Dr. B. J. Snyder, Program Director -TNI Program Office UNITED STATES ENCLOSURE (4410-85-l-0138)

NUCLEAR REGULATORY COMMISSION "" .. :"

Mr. .}ohn Barton kting Director of 1MI-2 Metropolitan Edison COmpa"1 P. O. 80x 480 Mi ddl etown. PA 17057

Dear Mr. Barton:

WASHINGTON.

D. C. ass October 22. ',81 This is in response to Mr. Hovey's 'etter LL2-8l-D2l4 of September

11. 1981. concerning the use of EPleOR-II"for SDS effluent polishing.

which included Metropolitan Edison's plans fOr tPICOR-II liner radioisotope loading and disposal.

In that letter. Met-Ed proposed to load the EPICOR-II liners to a maximum concentration of 1 ut-/cc of isotopes with half lhes greater than five years and dispose of the liners (with resins in a dewatered, but fied form) at the bottom of a disposal trench (approximately 10 meters deep). Even though not specifically stated. we understand that Met-Ed is proposing to dispose of the EPlCOR-Illiners at an arid disposal facility.

Prior to final promulgation of Part 6T,your proposal would be allowable under current NRC regulations.

Subsequent to final promulgation of 10 CFR 61, the remaining waste covered by your proposal would require an exception to the Sr 90 concentration limit (0.04 uc/cc) in Table 1 for Class A waste if the regulation is approved as proposed by the staff *

• The NRC staff has performed an evaluation of the waste and disposal conditions proposed by Met-Ed. The evaluation indicates that the proposed conditions would be acceptable fOr the waste to be considered a Class A unstabilized waste under 10 CFR 61. provided all other requirements of the proposed 10 CFR 61 for Class A wastes were met (e.g ** the waste is segregated from Class 8 and C stabilized wastes and disposed of in a separate trench). Since the existing commercial disposal sites are re,ulated by the individual States. acceptability of the waste form and disposa conditions would rest with them. However. it is our position that we would recommend acceptance of your proposal.

--" ENCLOSURE . (4410-85-L-0138) . . ... . . .

• Mr. John J. Barton . . It is requested that you continue JOur cireful Inalytical program to determine the content of these. isotopes in the various waste containers to .nsure conformance with the criteria discussed above. -:-_. .. .. :,-... -:.-:

• cc:* See Service Distribution List Sinc.rely

• Bernard J.

1MI Program Office . Office of Nuclear Reactor ReQulation . . a-"

-.-* < . . .. . ' . . -** 1 .... ENCLOSURE (4410-85-l-0138) for Dilpolal of UDltabilized tHI-2 Devatered le.in .. . .

SrlO Concentratioi.

Creater than 0.04 uc/cc Pur,!.e: Tbe JlUrpoae of rul evaluation Sa to 'eterlll11le the acceptabi9btJ of dl.poliDa of aD.tabililed

!KI-2 'evatered re.iD "Ite. Sr concentratioDa ,r .. ter thaD 0.04 uc/cc,-the upper limit for Sr concentratioDa for Claaa A _Itel lpecified til the Fopoled 10 cn. 61. leferencel:

1. Propo.ed rule, 10 en. 61, Licenlhl lequire.entl for Land Dilpol81 of ladioactive WIIte, Federal leauter, Vol. 46, aD. 142, July 24, 1981, pp. 38081 -38105. 2. Draft lAv1rODllental x.pact Statement OIl 10 en 61 "LiceDahl _ntl for I.aDd D1Ipoial of ladioactive Vaate," atmEG-0782, Append1z C. 3. DVDSI co'e nD, .JUDe 12, 1981. lelulu: D1Ipol81 of 1H1-2 'evatered reliD ** t .. havina Sr lO concentrat10Da le.1 than 24 uc/cc WDuld .. acceptable for dlapoaal in an unatabilized CODdition at 'eptha ,reater than 5 .. tera at aD arid di.poa81 aite. If other ilotopea liated in T.ble 1 of the propoaed 10 CFR 61 are alao preaent, thele ilOtOpel would alao Deed to be accounted for uahl the concentration ratio factor identified in Table 1. Ivaluation:

Tbe propoaed rule for 10w-le.el waate .. nllement, 10 CF.R 61, includes a _ate ela ** ification

.,.atem (Reference 1). The upper cOJ6entration limit for the d1lpol81 of unatabilized

... tea (Clasa A) for Sr 11 ,i.en u 0.04 uc/cc. Thu l1JD1t .a determined by evaluetinl the effecta of intruder pathvaya at a reference diapoa81 fac1l1ty.

Tbe intruder pathwaya included CODatruction and -aricultural cuea. Tbe * . '.

. . . . . '. . . . .. 2 ENCLOSURE (4410-85-l-0138)

'raft environmental impact atat .. ent for 10 era 61 (leferanc.

2) provlde. a'aetalled

'e.crlption of theae

• • . . , .. a * :. "'., . "': * * ._ !be cODcentratloaa for the intruder pathway .. aluatloaa in the ... te cal ** lflcatlon ay.t .. are '-eed on a performance obj.ctlve that the ricelve.

• azmual do.e to the whol. t.ody of lea. thaD 500 area. !be va.te cl ... iflcation

.,.t .. in aeference 1

that ... te. burled at Dermal .. pthe (tDclude.

d1tao.al at Ie.. than , .etera) at .lther" humi. or arid.alt .. 'aviDa Sr concentratlon.

ar .. ter tban 0.04 . 'Dc/cc" atabUb.d.

aovev.r, 10 en 61 'oe. provlde for exeaptlon.

if the .pacific dll,o.al condltion.

provlde ... urnce that the perfol"lllnce objectlve.

are .. t. ID evaluatlna certaln optlon. Whlch could provlde the a.surance tbat the ,.rforunce objectlve.

are .. t, aeveral 90 alterutlv

.. could .. con.lderad for autabUbed

• • t ** witb Sr concentratloaa areater than 0.04 'Dc/cc. !be.e alternatlv

• • tDclude: burlal at 'epthe ar .. t.r than 5 .et.r. (tbat 1., v1.th an lntruder krrler), burial at

• arid alt., or a comblDat101l of th .... lecau.e tbe propo.ed ... t. would" 'Dut.bUleed, the ** t ** would be .1I,o.ed of :lD a trench CIa ** A ** t... Ca ** A ** te. would be aeareaat.d from the atabUleed Cl ..... d C .. te.. fte '-elc a.8\1mptlou

lD the Cla ** A ** te .ceuno. for Dermal 'epthe and .eeper 'epth. (ar .. tar than 5 .. tar.) are .. follov8: 1. !be raferance dlapo.al alt. 11 locat.d in a humid South ... tern alt ** 2. IDadvertent intrualO11 11 .. de after tDatltutloul control 11 lo.t follov1na an actlve control ,.nod of 100 year ** 3. At the t1me of intru.lon the ... t ** bav. 'earaded to the .xtent that they are UIlr.cop,luble

.. va.t. and Ulull1tlnaulahable from IOU. 4. !be ** t. 'earadatlon tate. place at a rate independent of alte location.

!bat 1.,' the .earadatlon 11 the aame for an arld and a humid alt ** 5. Aarlcultural actlvitl ** occur only in ** te. located la ** than , _t.r. below arade. 1.'bla 1. '-eed on the coutructlon of a ra.ldenee v.l.th a ..... ent excavat.d to 3 .eter.. fte aoU.

\ .. . .. ... .. . -. .. . .'

• ENCLOSURE " (4410-85-L-0138)

.,." .

for lbe are Iraded about the re.tdence and '-:0:;. ,.. .f0048 are arow ill the excavated a011ll. ,._ aU. I COutructtoll

.. ante DOnall1 tab place at .. ptha le ** thaD 3 .-.. tera. "

• 7. When deep di.po.al s.. "'U1Ded, it s.. judged le ** l:f.bl, that aipUicant coutruCtioll w111 tab ,lace at tbe.e deptb. (blab ri **

coutructio'D, for uample). ror ** te. thu. dt.po.ed, it s.. ... umed that 0111, 10 percent of the ... t .. are cODtacted and become .vallable for di.per.ion Into the air and aubtequ.nt 1Dhalatlon

'" human.. Further, potential direct lammA expoeur ** from .orkilll 011 bomo&eneouel, contaminated IroUDd are ... U1Ild to .. reduced " a factor aqual to one _ter of 80il abielcU.q (1/1200). . With the.e '-elc ... umptiou the allowable Srto concentrati011e for the atated option. vere computed u111& the DVER.SI code which ** al.o a.ed to determ1De the limit111&

radionuclide cOllcentrati01Ul for the 10 CFl 61 ** te et ... Uleation .,.tem .(leferellce 3). !be n.ult. are provided in Table 1. Table 1 'Allowable Br to Concentratiou for U11Itablized Va.te. Opti011 un.tabilized

... te, di.po.al (110l'1D&l depth.) un.tabiltzed

... te, at depth. ar .. ter thaD 5 .. ten

• Allowable Concentratioll, Allowable Concentratioll, Coutruction Scaario, qricultural Scenario, ac/cc ac/cc 2.0 0.04 24 qrleultural activiU.e.

are DOt ... uMd to take place for ... te. di.po.ed at depth. areater thall 5 .. ter ** '" \

.*. ' . . . . *z-* .... " . I .. , ENCLOSURE (4410-85-L-0138)

.;" .:-.' -... . Ilace the dl.po.al alfeeu for u ari4 ucl a lluld.d alte are ... umed to .. tbe the allowable collcelltr.tlou would" theta....

Bowver, t. above, n.a1utlO1l haa eODalclere4 oral7 laoto,. Ir ud Iaaa 1lot 14 '--evlgu.te4 eff.ct. of otb.r lia1tlDa loq-llved laoto,.. aucb .. C * 'Ie

• or 1 whlcb 1Il,"t .. pre ** llt !D a ... te ofthll .. mre. !be ** laoto,.. hav. blab -l&r.tloll pot.lltlw at IwBlld alt ** INt ar. lell.ra11 7 , Dot apeclflea117

.... ur.d at JOftr pluta 'ue .to low C01lc.lltr.tlou ud au17tle complent7. .ll.lovllll dllpo.al of hlah.r aetlvit7 ** t.bUlaed ** t .. at bull1d dl.po.al alt ** could nault ill iIler **** d Iro\lZulvat.r alar.tlO1l of nch ltm1tllll 101ll-11ved aobUe laotopea .. w11 a. iDcr .... d .,.t operatl01l81

_tAteuIlC.

co.u. liDee lt la po ** lble tbat na-2 ** t .. al,"t &lao COllt.iD .ome of th .. e 101l,er-l1vu laotopu ill cOllcelltratlona

... r thalr Cl *** A ltmlu. it la jud,.d to .. prudellt to dl.po ** of aueb bl&".r aetlvltJ uatabUl.ed w.t ** at u arld .lte where it, call .... lUMd that qr.tlOD la IIOt a af.p1ficat patbva7. t'b1a evaluatloll.

iIl.refore.

CODcl8d .. that .l.po.al of .. tabUl ** d !KI-2 'ew.tered re.ln ... t ** bav11la Ir CODe.lltratlona up to 24 ac/ec .ould be accept.ble provlded the ... t.. wr. INried at '.ptha Ir .. t.r thall 5 _t.r ** t aD arid dupo.al alte. Oth.r laotope. llat.d ill T.ble 1 of Ieferellc.

1. of, cour.e, would ** &4 to .. aeeoWlted for -11la the cOllcelltr.tloll zatlo factor i4elltlfled 1D !able 1. lvaluatloll parforMd '7 Date ,..,,-e I

__ ..

Approved '7: l\ .' *

• Date __ _ *

.I *-ii*'" I '" ..... Docket No. 50-320 ---.noITATII NUCUAR REGULATOR" COMMISSION WASHINGTON, Do Co _ Declltber

19. 1984 Mr. F. R. Standerfer, Director Three Mile Island Unit 2 SPU Nuclear Corporation P.O. Box 480 Mfddletown, PA 17057 OIar Mr. Standerfer:

Subject:

Three Mile Island Nuclear Statfon, Unit 2 Operating License No. DPR-73 Docket No. 50-320 ,. J Technical Specification Change Requests 39, "41, 43 Recovery Operatfons Plan Change Requests 19, 20, 22 Exemption Request from 10 CFR 50.5Sa (Code Safety Val"es) Exemption Request fran 10 CFR 100, _pendix A and CFR 50.36(3) (Sefsmic Instrumentation)

_ .-; I The Nuclear Regulatory Commission has issued the enclosed Amendment of Order; Recoyery Operations Plan Change Approval of fran the requirements of 10 CFR 50.551 for e Safety Valves; and Approval of Exemption from the seismfc instrumentation requf .... nts of 10 CFR 100, Appendix A. and 10 CFR 50.36(3).

The Amendment of Order tltfch lIOdifies .ny .ections of the Proposed Technfcal

$pecificatfons ePTS) .. s requested by leneral Public Utilfties Nucl .. r Corporation (SPUNC) in letters dated .January 12. 1983, Sept8lber 12 *. 1983 and Septenber 3D, 19a3. Other docllllents related to thfs request Include: lecovery Operatfons Plan (ROP) Changes tltich flere requested In separate letters al so dated January 12, Septllftber 12, .nd Sept.ber 30. 1983; ..... quest for ex.ption fran the require-.. nts of 10 CFR 50.551 with respect to Code Safety 111fef Valves In

• letter elated April 18, 1984;.nd a ... quest for .n ex.ptton fran. the .ei.fc Iionitortng

... guir.ents of 10 CFR SO.36(3) .nd 10 CFR 100, Appendix A. Paragraph Yl(a)"(3) in

• letter dated April 18. 1184. .. As previously explafned fn a letter t,sued by the .taff on .JUly 17, "1184, your _ PTS .nd lOP change ... quests .... dfyfded tnto tw separate fl ** ncll. The first fl.ance .s _de on ",-, y 17 f 1984 .nd .. s l-.dfately effectfve.

the ltaff . Ills ...." .. JOUr* IIflty IVa uatfons for the .bove doc-.nts and concludes that . ,"ur requests addressed by this fSlu.nce ..... cclptable with .s discussed With your ltaff. PTS changes that .re the subject of this letter wfll become effectfve on January 7. 1185. The £I...,tfons to 10 CFt SO. 551 , 10 CFt SO.36(3) and 10 cn 100. Appendix A. Par.graph Yl(a)(3.)

.... affectj,y_

upon fl.lnce. " . *

( ( P!\GE 2-OF 7 Mr. F. Since the February 11, 1980 Order faposing the Proposed Technical cations is currently pending before the At_ic Safety and Lie_Sin, Board, the staff will be advising the Licensing Board of this Amendlent 0 Order through a Notice of Issuance of Amendment of Order and a Motion to Proposed Technical Specifications in Accordance Therewith.

-Federal Register Notices for the discussed is:suances are enclosed.

Coptes of the related Safety Evaluation and revised pages for the Proposed Technical Specifications and the Recovery Operations Plan are also enclosed.

Enclosures:

1.

of Order 2. Safety Evaluation

3. Proposed Technical Specification Page Changes Sincerely, Bernard J. SIllier, Program Director Three Mile Island Program Office Office of Nuclear .. actor Regulation

4. Recovery Operations Plan Change Pages 5. Exemption fram 10 CFR 50.55a I. Exemption from 10 CFR 100, Appendix A, Paragraph VI(a)(3) and 10 CFR 50.36(a) 7. Notice of Environmental Assessment and Finding of No Significant I_pact 8. Federal Register Notices cc: J. Sarton R. Rogan S. Levin R. Fr.-ennan J. Byrne Service Distribution List (see

.( UNITED STATES NUCLEAR REGULATORY COMMISSION In the Matter of GENERAL PUBLIC UTILITIES NUCLEAR CORPORATION

-Mile Island Nuclear Station. Unit 2) Docket 110. 50-320 EXEMPTION I. Enclosure 6 6PU Nuclear Corporation.

Metropolitan Edison Cc.pany. Jersey Central Powr and Light Company and Pennsylvania Electric Callpany (collectively.

the ltcensee) are the holders of Facil1ty Operating License No". DPR-73 ** ich had authorized operation of the Three Mile Island Nuclear Station. 2 (lMI-2) at powr levels up to 2772 Mga.tts thenaal. The facility ** ich 15 located in Londonderry Township.

Dauphin County. Pennsylvania.

15 a pressurized wter . l reactor previously used for the c..-rcial generation of electricity

• ( .. By Order for ModificaUon of License. dated .Ju11 20. 1979. the ltcensee's authority to operate the faci11t,y ws suspended and the Hcensee's authority

.. s 1111ftllf to .intenance of the facn it,y in the present sllltdown cooling .ode (44 Fed. Reg. 45271). I.Y further Order of the Director.

Office of "'clear .. actor. Regulation.

dated February 11, 1180. a new set of formal license requir_nts ws .posed to reflect the post-accident conciition of . the faci1it,y and to assure the continued .intenance of the current life. sUblet 10ng-teYII cool1ng condition of the facl1itl (45 Fed. leg; 11292). Th1l Heense provides, IIIOng other things. that ft fs. subject to a11 ",'es, regulations and Orders of the C ... 1Ision now or hereafter in effect.

r ( ( I f PAGE 4-OF -II. In I letter dated April 18, 1184, the Hcensee requested an ex..,Uon fran -the

... nts of 10 CFR 50 relative Seis.ic Monitoring Instrumentation.

10 CFR 50 50.36(c)(3) requires surveillance

... nts -*.** to assure that the necessary qua' ity of systans and cOlltponents is .. intlined, that flcil ity operltion

.111 be within the safety l1_its, Ind that the conditions of operltion will be _t.* 10 CFR 100, Appendix A, Section VI (1)(3) states that: *Suitable instrUientltion shill be provided so that the seismic response of nude.r powr plant feltures i_portant to sifety Cln be determined promptly to permit comparison of such response with that used IS the deSign basis. Such a canparison is needed to decide .ther the pllnt Cln be operated safely Ind. to permit such ti.,y Iction IS .. y be approprilte.

These criteril do not address the need for instrumentation_that .ould .. tica"y shut down I nuclear powr pllnt an earthqulke occurs exceeds I predetermined intensity

• Presently, Section 4.3.3.3.1 of the 1MI-2 PTS requires that Transaxill Time -History Acc"ographs be operlbl. tor the Relctor Building Ring Girder Ind the "actor Building Mit; that Triaxill .. at Acc"ographs be operable for the .. actor Service Structure, ... Core Flood Tank Piping and 2-1£ .. ar; that Triaxial Sets_ic Switches be operable for the Rlactor Building lase and that Triax1al Rlsponse -$pect .... Rlcorders be oper.bl. tor *the .. actor Building Mit. <t '. ... ..

i JU. The 1M1-2 core is cooled via loss of heat to reactor building -envJ.rornent. , This is I passive mode does not require any ..chanical equipment to be operating to .. fnta1n In to cool tile core. As stated in 10 CFR 100, Appendb A, SecUon Yl(a)(3), one of the reasons for seismic instrumentation is to decide Whether or not the plant can be operated safely. In the July 20, 1979 Order for Modification of License, the authority to operate the

.. s suspended and the licensee's authority was limited to the .-fntenance of the in the present shutdown ing .ode. Therefore this basis for Sectton YI(a)(3) does not Ipply to TMI-2. In reference to the seismic instrumentation providing 1nfonaat1on for timely actions by plant personnel and the NRC, it is the stiff's op1raion that if a seismic event wre to occur at 1MI, the status of the core lIIOuld not be affected because of the passive cooling .ade and therefore no tm.ed1ate actions lIIOul d hive to be talten to _inta1n the heal th and safety of pubHc. It is also the staff's opinion that wn considering the above discussion, .. intenance and surveillince requi,...nts for se111111c .ntation is a'so not justified and is an unnecessary burelen on the licensee.

Because of the of the Heens .. 's luthority to operate the facfl tty 1n other than the present recovery lIOde as defined 1n the Proposed technical speciftcations, of the regulations, w.ich are fntended tq apply to ltona1 operating plants. are st.p1y inappropriate Ind, .,re significantly, are unnecessary to protect the public health and safety. liven ** tque .....

{ . . ". -status of the plant in tenu of pr1 * ..., s1St. t_peratur.

and pr.ssu .... avan-able ftsston product inventory.

the .bl1tty to cool the .... ctor wtthout forced -cin:1l1atfon (loss-to-IIDbfent).

and the low decay .... t rate ** fntenance of the fac11ity with the .... pUons granted her.by wi:" provfde .n adequate level of safety. IV. Accordfngly,'

the eClllllission has deter'llined that. pursuant to 10 eFR 50.12. an ",., .xempUon is authorized by law and will aot endanger 11fe or property or the CCllr.lOn defense and security and is otherwise in the pubHc tnterest.

lased on the discussions above. the Cor.IIIission

"'reby grants .n l .. pUon to the ( requirements of 10 eFR 50.36(c)(3) .nd 10 eFR 100. Appendix Ai l Section VI(a)(3) relative to sefsmic instruaentatton.

It is further detenntned that the .x..,Uon does not authorize a change' in effluent types or total lIIOunts nor an fnc .... se fn po_r level and will not result fn any significant Inviro_ntal fllpect. In light of this atnation and as reflected in the Environ.ental Assessment and Notice of Finding of 10 Significant Environaent.'

llpact prepared pursuant to 10 eFR 51.21 and 51.30 through 51.32, fssued concurrently herewith,*

it "5. OF 't

( ( ( -! concluded that the instant action is insignificant fran the stindpoint of .nviro ..... tal _pect and an envirornental i_pect statement need not be * "y of ,*z.*,e prepared.

Effective Date: Dec_ber 19. 1984 Dated at Bethesda, Maryland Issuance Date: December 19. 1984 FOR THE NUCLEAR REGULATORY COMMISSION

"/p aL.. Harold R. Denton. Director Office of Nuclear .. actor Regulation

. -. -1t!1-2 . ... lDllTll.::flO11 Al'. -

PAGE <6 OF cr-I .... t.: . ,lar'-Trlr L.li1!INuclear V IV GPU Hue ... Corporation Post Office Box 480 Route 441 South Middletown, Pennsytvania 17057-0191 717944-7621 , 1MI Program Office Attn: Dr. B. J. Snyder Program Director US Nuclear Regulatory Commission DC 20555

Dear Dr. Sny der :

I ! I I -aT -u -1:---war :-n-u . " -.. :t -I.". -T 1>.

",'on--on,,!,:-&ra:' :-" .. -I TI .-T -. ft. lLL -

., -...-. \':r':5 ....... -I : ." , I I Writer's Oirect Oial Number: (717) 948-8461 441D-8iK-0050 DoclJ'nent 10 0481U April 18, 1984 Three Mile Island Nuclear Station, lklit 2 (TMI-2) license No. DPR-73 Docket No. 50-320 Seismic MJnitoring Exemption Request Your letter of January 13, 1984, Which provided conments on various Technical Specification and Recovery Plan Change Requests required that GPUt'£ submit a specific relief request from the seismic requirements of 10 CFR Part SO. Based on the attached justification GPUNC requests an from the seismic requirements of 10 a=R Part so. As this request is sUbmitted in conjunction with Technical Specification Change Request No. 43, no additional fee is required.

Please call Mr. J. J. Byrne of my staff if you have any questions on this information.

E£KIJ::B/jep Attad'lnent Sincerely, lsI E. E. Kintner E. E. Kintner Executive Vice President cc: Deputy Program Director -1MI Program Office, Mr. L. H. Barrett GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation

• . . .... l .lJSTlFICATION EElETING SEISMIC MONITORING REQUIREMENTS f .,' C. r r r C; '1 Paragraph (c) of Section 50.36, "Technical Specification", of 10 CFR Part SO provides that the Technical Specifications will include surveillance requirements to assure that the necessary quality of systems and components is maintained, that facility operations ,will be within safety limits, and that the Umiting Conditions for will be _t. Appendix A, "Seismic and Geologic Siting Criteria for tU:lear Power Plants", to 10 (7R Part 100, "Reactor Site Criteria", requires in Paragraph VI(a)(3), a suitable program for this requirement with regard to seismic instrunentation needed to determine promptly the seismic response of nuclear power plant features Important to Safety to permit cc:Jq)8rison of such response with that used as the desi", basis. That is, the seismic lDDJlitoring instrunentation is used only to record and define actions after a seismic event. These actions consist primarily of engineering evaluations to determine damage caused by a seismic event and the repairs required prior to restart. As performing the surveillances on the seismic instrumentation would require 3 to 5 man-rem per year (Reference Gf\JNC Letter 4.410-83-L-OlSl dated July 20, 1983), and the data provided would, in general, not be needed unless a decision is made to restart TMI-2 continued performance of these surveillances is not a prudent man-rem expenditure.

Additionally, US Regulatory Guide 1.12 by reference to ANSI Standard N18.S provides guidance on seismic instrumentation required for multi-unit sites. Section 4.4 of ANSI N18.S states that, "Instrumentation in addition to that installed for a single unit will not be required if essentially the same seismic response is expected at the other units based on the seismic analysis used in the seismic desi", of the plant." Gi ven that both un! ts are located in close proximity and are both founded on bedrock, it is expected that the lhit 2 seismic response will closely approximate the Lnit 1 response.

Therefore, given the above guidance and the Recovery status of TMI-2, Technical Specification 3.3.3.3 which requires surveillance of seismic instrumentation in Unit 2 can be deleted without imposing a si",ificant risk to the health and safety of the public and would save an exposure fo 3 to 5 man-rem per year to TMI-2 workers. Document 10 0481U

( UNITED STATES /3-dI5g> NUCL.EAR REGULATORY COMMISSION WASHINGTON, D. c. 2OI5S Docket No. 50-320 Mr. B. K. ICInga, Director Three Mile Island Unit 2 6PU Nuclear Corporation P.O.Box 480 Middletown.

PA 17057

Dear Mr. Kanga:

July 22. 1983

Subject:

Three Mile Island Nuclear Station, Unit 2 (TNI-2) Operating License DPR-73 Docket No. 50-320 . PAGE / 10 CFR 50.49, -Environmental Qualification of £lectrical Equipment to for Nuclear Plants-The NRC has reviewed your letter dated April 11, 1983 requesting exemption from the requirements of 10 CFR 50.49, -Environmental Qualification of Electric Equipment I.portant to Safety for Nuclear Power Plants.-This rule requires that by May 20, 1983, each holder of an operating license issued prior to Februar,y 22, 1983, shall identify the the scope of the rule already qualified and submit a schedule for the I i I OF3 _-2 .-.u.c ... ---61'. .

aa"14 ti . , I ... 1p4Tol < \ .... tal 9' . I *** ,. na iii ,Ad. a., l . qualification or replacement of the .... ining equipment.

The rule also specifies a deadline by all the equipment lUst be qualified.

t., .. . ... , la, -IIIYI : .... The bases for your exemption request arei 1) campHance 10 CFR 50.49 .rtll not contribute to the overall safety of 1MI-2 due to its present post accident condition.

2) campliance would direct resources better utilized to accomplish cleanup activities, 3) item 3 of inspection report 50-289/ 50-320 which reflects acceptance of electrical at TNI-2 as required by Inspection and Enforcanent (IE) Circular 78-08. and 4) the R:;: .. , TNI-2 response to IE aulletin 79-01B which refers to it .. 3. J Section 50.49 of 10 CFR Part 50 MIS developed taking into consideration previous efforts by licensees.

As SUCh, the exemption bases cited, ularly 3) and 4), do not justify an exemption to this rule. It should be noted that a detailed review of the environmental of related electrical equipment in 1MI-l has been cClllpleted.

with .ny deficiencies identified.

These deficiencies are also applicable to identical equipment in 1MI-2. Therefore the staff concludes that you have not provided an acceptable bases for granting an exemption.

( Mr. B. K. ICanga However, w do believe that the current status of TMI-2, reflected in bases 1) and 2) justifies an extension to the deadlines specified in the rule by which each licensee lUst provide a response and by which all equipMent lUst be qualified.

We therefore conclude that the protection of the health and safety of the public not decrease by extending the cGlpliance deadlines. . . The licensee therefore shall d .. onstrate cCllpl1ance 10 CfR 50.49 . not less than six .nths prior to the anticipated re:turn to po __ r of lMI-Z or by the deadline in the rule which ever is later. cc: -1. Barton Byrne J. Larson Service Distribution List (see attached)

Sincerely, Bemaril J. $rW&er, Program Di rector Three Mile Island Program Office Office of Nuclear Reactor Regulation

... . . ---------------

PAGE 3 OF 3 (jNuclear GPU Nue"a, Corporation Post Office Box "80 Route .... , South Middletown.

Pennsylvania 17057 717 9"4-7621 TELEX 8 .. -2386 .... Writer"s Direct O,al Number:

• tMI Program Office . .u 11,1983 41+10-83-L-007S Dr. B. J. Snyder. Program Director J3 Nuclear Regulatory Camd ssial 1ash1ngtcm.

DC 20555 lear Sir: '!bree Mile Island Nuclear Statial. Ddt 2 ('00-2) Operating L:lcense No. 1JIR-73 Docket No. .50-320 10 CPR 50.49 "Envircnll!lltal of Elec:t:ric EquipDent liIportant To Safety for Nuclear Power Plants" J\., ,:lear (hzpcn:atial requests exaJptial fran the of 10 CPR .50.49 'f.rMrt:nJEntal Qual1f1catial of Electric EquipDent lIIportmt to Safety for b::l ear Power Plants". lue to the present post-accident cmditial of '00-2, ClClJl)l1ance with LO CFR 50.49 will not ccnt:rlbute to the overall ufety of '00-2 and J<<JUld tlvert resources better utilized to 8CCC11plish cleanuP activities.

In 1ddit1an, Item 3 of Jnspectial 50-289/.50-320 79-01 nflects -=eptance

>f e1ectri.ca1 ecpdpmnt at '00-2 as required by I:Dapec:t1.m

.-ad BlforcellB1t

trcular 78-08. '!be '00-2 mspcme to I:nlpeCtian md BlfOZ'CllDE!rlt Bulletin 79-01B refers to the above I:napectial.

!lerefore.

GPLK: believes that grant1ng fran 10 CPR .50.49 for '00-2 aild not decrease the protect1im of die haalth md aafety of the public Ind 1IOUld provide for mre effective utUizatial of ft8CUrCeS for the cleanup If 00-2. --rIO .-.'IS/jep Bmtoere1y, lsi I.. C. Amold I.. C. Amold Preaidmt L. ..,r. L. B. Barrett, DIputy rzosr-DinIctar -'JH[ Office GPU Nuclear Corporation is a subsidiary of the Genera' Public Utilities Corporation

_-2 Al'.,:-=-__ _

=t)15i':'i%:

-.. 5&: . ..::::-___ _

TMI Program Office Attn: Dr. B. J. Snyder Program Director US Nuclear Regulatory Commissior Washington, DC 20555

Dear Dr. Snyder:

1Jiiibite,_

!l IQn lla:d*'trlr, fed .... A . '1" ., ':"'1 . f\ "nt*A . nO\.* .. . In roclt .. aT 1'::1. ..... 'r'I"A I __ . /';,'1 'B .. rtO,.. ... A . 'U':":'I-1!"lltl. .rl '0. i..Ir,on-A I. I fV;TI*A. , I or.:icr .. 111:.--:.

r'::., .. ! r. )'

... , .... :-,:, . . I A ' ! ""IV" i I I I v'l v ,

! , I , : , I Irl '!). I I ! *.

., v : ,1/ : " i I ! t: 44\';£*5 ----.",-' GPU Nuel ** r Corporation Post Office Box 480 Route 441 South Middletown, Pennsylvania 17057-0191 717 944-7621 TELEX 84-2386 Writer's Direct Dial Number: (717) 948-8461 44l0-84-L-0058 Document ID 0788y Ap ril 9, 1984 Three Mile Island Nuclear Station, Unit 2 (TMI-2) Operating License No. DPR-73 Docket No. 50-320 Relief from ISI Test Requirements for Category B & C Valves . This letter is in response to the NRC letter from Dr. B. J. Snyder of April 27, 1981, granting relief from ASME Section XI, "Inservice Inspection Requirements", and requests relief on an individual valve basis. Your letter stated: We recognize that Category B and C valves in systems out-of-service need not pe tested during the shutdown period. However, we believe that all Category Band C valves in safety related systems in-service should be exercised at least once per 92 days where practical to determine their operational readiness.

Relief from the test requirements for Category B and C valves in safety related systems specified above will have to be submitted on an individual valve basis. GPJNC performed a review of the lSI Valve Testing Program in order to establish the basis for relief from test requirements for specific valves. The review was conducted as follows: All valves which were listed in the TMI-2 lSI Program for Valves, dated December 7, 1977 and AP 1042 Rev. 1 (lSI Systems List and Retest Requirements) dated 10/02/79, were reviewed.

From this listing, certain valves were eliminated for the following reasons: (see Table I) A. The system is out of service for the Recovery period. GPU Nuclear Corporation is a subsIdiary of the General Public Utilities Corporation Dr. B. J. Snyder April 9 , 1984 4410-84-L-0058 B. They are category "A" valves and exerrpt frOOl testing as stated in NRC letter from Bernard J. Snyder to Gale K. Hovey, dated April 27, 1981. C. They are in an inservice system, but do not perform a safety related function.

D. E. Testing is impractical.

ALARA considerations

--As areas in TMI-2 are decontaminated sufficiently to allow entry on a routine baSiS, the testing of these valves will be reevaluated.

F. They are Mini-Decay Heat Removal System valves and are exempt from lSI Testing as stated in NRC letter from Bernard J. Snyder to Gale K. Hovey dated April 18, 1981. All remaining valves will be tested and are listed in Table II. Additionally, valve DH-V135 was erroneously included in the 1977 lSI Testing. This valve is not installed in the plant. GPUNC requests your approval of the revised lSI program shown in Table II and, upon approval, will begin the preparation or revision of the necessary procedures in order to implement this program. If you have any questions, please contact Mr. J. J. Byrne of my staff. BKK/jep .. Attachments Sincerely, /s/ B. K. Kanga-B. K. Kanga Director, TMI-2 cc: Deputy Program Director -TMI Program Office, Mr. L. H. Barrett *

• r

• TABLE I A. Valves in systems out of service during the recovery period. MS-Rl MB M5-R2 A&B M5-R3 A&B M5-R4 A&B M5-RS A&B M5-R6 A&B MS-V3 MB M5-V4 MB M5-V7 A&B MS-Vll A&B M5-Vl2 A&B MS-Vl4 M5-V207 FW-VlB A&B CO-VBl MB CO-V21S A&B EF-Vl A&B EF-V2 EF-Vll A&B EF-Vl3 A&B EF-V26 EF-V27 A&B EF-V32 A&B EF-V33 A&B RC-Rl MB RC-Vl RC-Vl49 IC-V2 IC-V3 IC-V4 IC-VS IC-VlOO IC-Vl47 rJ-V4 MB rF-VS A&B rF-VlOO A&B CF-Vll4 A&B rJ-Vl44 rF-Vl4S rF-Vl46 BS-Vl A&B BS-V4 A&B BS-VlOO MB BS-VlOS A&B BS-Vll3 BS-Vl30 A&B PP-VllO A&B EB-V6 EB-V7 EB-VB EB-V9 EB-VlO B. "Category A" valves which are exempt from testing DW-V2B SA-V 20 MJ-V2 A&B MJ-Vl6 A,B,C&D MJ-VlB MJ-V2S MJ-Vl6l A;B, C&D MJ-V376 MJ-V377 MJ-V37B MJ-V402 A,B,C&D g:'-VlDS WDL-Vl092 WDL-Vll2S WDG-Vl99 NS-V72 NS-VBl NS-V99 NS-VlOO CA-Vl CA-V3 CA-V6 CA-VlO DC-Vl03 OC-VllS t-tv1-VS2 AH-Vl A&B AH-V4 MB AH-VS AH-V7 AH-VS2 AH-V60 AH-V62 AH-V72 AH-VBl AH-V90 MB AH-VlOl AH-VlD2 AH-VlOS AH-Vl07 AH-Vl2D A&B WDL-Vll26 CA-VB CA-V9 SV-VlB SV-VSS PP-VllO A&B

,.. v. TABLE I Valves in systems which are in service but are

• recovery required to perform safety related functions dL J Valve No. MJ-VlO MJ-Vl2 HY-V55 MJ-V28 MJ-V36 MU-V37 MJ-V433 MJ-V434 MU-V439 MJ-Vl27 MJ-V325 MJ-V326 DH-Vl Q-i-V2 DH-V7A1B Explanation This valve was formerly used for boron control. It is in a line which ties in lines from the demineralized service water system, reactor coolant bleed tanks, the boric acid pumps, and the deborating demineralizers.

Since boron control is now performed by the Standby Pressure Control (SPC) system, MJ-VlO need not be verified operable.

This is an outlet valve for make-up tank lAo The make-up tank is no longer in use, and therefore this valve is not required to function.

This valve is part of the hydrogen supply manifold to the make-up tank, thus the justification is the same as for MU-Vl2 ** Justification is the same as HY-V55. These valves are in a line from the make-up (MJ) pump discharge header to the seal return coolers. The line is also a path for operating the MJ pumps on recircUlation.

The seal return coolers are not required to be operable during recovery.

Additionally, the breakers for the make-up puIl'Ps are racked-out, thus the make-up pumps will not be operated during the recovery mode. These valves are part of the discharge line from the MJ pumps to the seal injection line. Seal injection is not utilized during recovery, therefore these valves are not required to be operable.

These valves are in a line from the boric acid system to the MJ system. Since boron control is performed by the SPC system, these valves serve no safety function.

These valves remain open during the recovery period to allow Reactor Coolant System (RCS) pressure sensing and level indication.

Additionally, DH-V2 is a containment isolation valve, and, like category "A" valves, is maintained according to the Recovery Technical Specifications.

These valves may be opened to allow "piggy-back" operation of the MU &: Decay Heat Removal (OrR) pumps (Le., the DH system supplies water to the RCS through the MU system) when the plant is at operating pressure.

During the recovery period, such operation is not required, therefore, DH-V7A/B serve no safety related function.

v'alve No. DH-V8A/B DH-V146 rn-V228 DH-V171 rn-VI8SA&B DH-V190 NS-V67 NS-V83A/B NS-VS4AIB NS-V 21 5 NS-V216 NS-V32 NR-VSA/B NR-V9AIB NR-V27AIB TABLE I (contlnueO)

Explanation These are the sodium hydroxide (NaOH) tank discharge valves. Since the NaOH tank will not be used during recovery, there is no need for these valves to be operable.

These valves are the vacuum breakers for the NaOH tank. Justification is the same as above. This valve is a bypass around DH-VI. DH-VI is maintained open at all times to provide a means of level indication.

DH-V171 serves no safety function in that DH-VI provides the necessary flow path from the reactor. These valves are located in the auxiliary pressurizer spray line which is not required to be operable durirg the recovery period. Later in the recovery period these valves will be used occaSionally for flushing water out of the pressurizer into the ReS. However, these valves will not be required to fulfill a safety function.

This valve is in the supply line to the seal return coolers which are not operated during recovery, therefore, this valve is not required to be functiona

1. NS-V83A/B are the inlet valves for the nuclear services (NS) cooler. NS-V84A/B are the outlet valves for the NS coolers. NS-V2IS and NS-V216 are the bypass valves for the NS coolers. Due to the extremely low heat load the cycling of these valves is not required.

This valve is in an NS line which discharges to the reactor coolant evaporator.

Since this is currently not in use, NS-V32 is not required to be operable.

Later in the recovery period this valve will be used occasionally for operation the reactor coolant evaporator; however, it will not fulfill a safety function.

These valves are situated in a line between the NS coolers and the mechanical draft cooling towers. There is no emergency situation which would require these valves to operate durirg the recovery period. These valves are part of the Nuclear Services River Water (NR) supply to the Emergency Feedwater (EF) pump suction. The EF pumps are removed from serVice, thus, these valves are not required to be operable during recovery.

Valve No. f\R-V46A/B NR-V5l NR-V55 NR-V246 DC-V96A/B OC-VllBA/B WDS-Vlll RR-V1A/B/C/D RR-V2A/B/C/D RR-VllA/B/C/D RR-VSA/B/C RR-V6C/D/E I AI:jLt. 1 \COm;.l.llut::u/

Explanation These are the suction check valves to the reactor buildirg emergency coolirg pumps which are no lorger in service. Therefore, there is no need to test these val ve s. This valve is part of the NR supply to the intermediate coolers which are no lorger in service. Therefore, there is no need to test this valve. This is an isolation valve for the river water pump house de-icirg line. Since there is virtually no heat load in the plant, openirg the de-icing line is not required.

Therefore, there is no need to test this valve. This is the suction valve to the control buildirg area east fan coil units. Due to the extremely low heat loads during recovery, the operation of this valve is not required.

These valves are associated with the leakage coolers (WDL-C-1A/B) which are not in operation during recovery.

This valve is in a line between the reclaimed boric acid punps and the roric acid mix tank. Since the process of reclaiming boric acid is not done during recovery, there is no need to te st this valve. This valve, however, may be used later in the recovery period for transfer of miscellaneous waste to the boric acid mix tank at which time the testing of this valve will be re-evaluated.

These are discharge valves for the reactor building emergency cooling river water (RR) booster pump. The RR System is no lorger required to be operable during recovery and has been removed from the TMI-2 Technical Specification Additionally, RR-VllA/B/C/D are containment isolation valves and are maintained in accordance with the TMI-2 Technical Specification.

These valves are located upstream of the Reactor Building Normal Cooling Coils. This system operates to maintain a habitable environment in the Reactor Buildirg, but is not required by the Technical Specifications.

The original safety function of these valves was to open on an Engineered Safety Feature (ES) signal. Since the ES system is out of service during the recovery period, these valves no longer serve a safety function.

Valve No. HR-VL50 SW-V5A/B/C U. Testing is Impractical MU-Vl43 A,B,&C A,B,C,&O OH-Vl48 A&B UH-V3 uH-V6A & B TABLE I (co Jed) Explanation This valve is part of the outlet line from the "0" RB Cooling Coil to the Evaporator Cooler. This valve is presently red tagged closed as ,part of the isolation of the "0" cooling coil. Due to leakage of this cooling coil, testing of this valve would potentially add re-contamination to the Reactor Building.

These are instrument root valves for the screen wash self cleaning strainer SW-S-3. There is no emergency situation which would require these valves to cycle during the recovery period. These are the discharge check valves for the make-up pumps. Testing is not practical since the make-up pumps would be required to operate in the recirculation mode thus injecting water into the RCS. The electrical circuit breakers for the make-up pumps are presently racked out in accordance with Section 3.1.1.1 of the TMI-2 Technical Specification.

This section requires the breakers for the make-up pumps to be racked out when valve OH-Vl or OH-V17l is open. OH-Vl is maintained open in order to monitor the reactor coolant level. Additionally, Technical Specification Change Request (TSCR) 39 submitted via GPUNC letter 44l0-83-L-0013 dated January 12, 1983, proposes deletion of the make-up pumps since, due to TMI-2's present mode of operation, there is no longer a need to maintain an operable high pressure injection system. This TSCR also requires that the circuit breakers for the make-up pumps be racked out at all times. These check valves are on the lines which take suction from the borated water storage tank (BWST) and discharge to the suction side of the make-up pumps. Justification is the same as above, in that testing is not practical since it would require operation of the make-up pumps. These valves provide isolation between the OHR system and OH-V4 A&B the ReS. Testing of these valves would potentially add contamination to the OHR system. Additionally, OH-V3 and OH-V4A, B are containment isolation valves and, like category A valves, are maintained according to the Recovery Technical Specifications.

These valves provide containment isolation between the OHR system and the reactor building (RB) sump. Justification is the same as for DH-V17l. _ r; Valve No. DH-V-107 A&B NR-V-269 NR-V270 RB-V4 RB-VS2 NR-VIBI AlB SW-V23 TABLE I Explanation These check valves, located inside the RB, connect the DHR system to the reactor vessel. Testing of these valves is impractical since it would require placing the RCS on recirculation through the Decay Heat System. These normally opened check valves discharge directly to the river. Thus, testing is impractical since there is no practical way to put back pressure the valves. These normally open check valves discharge to the evaporative cooler through the RB cooler. Thus there is no practical method to initiate backflow through the valves. Also, the size of the test line does not allow practical determination of the valve seating. These valves are in the domestic water supply line to the nuclear river (NR) pumps pre-lube system. These valves are always open unless there is a loss of domestic water pressure at which time NR-VlB6A/B/C&D and NR-V234 A/B/C&D (depending on which NR pump is operating) opens and NR-VIBI A or B closes. Testing NR-V1BI A&B would require depressurizing a portion of the domestic water system and then opening a drain valve. If, however, NR-VIB6 A/B/C&D and/or NR-V234 A/B/C&D failed to open, the NR pumps would be left without lube water flow. Therefore, it is impractical to test NR-VIBl A&B. Justification is the same as above, in that if this valve failed to open, the screen wash pump would be left without lube water flow.

TABLE I (continued)

E. VALVES NOT TO BE TESTED DUE TO ALARA CONSIDERAT Valve No. CA-V137 CA-V139 Explanation Although these check valves are not located in a high radiation area, testing of these valves requires the opening of valves downstream of them which are located in a high radiation area (i.e., the make-up valve alley). The dose rates for this area exceed 120 R/hr. F. The following valves are exempt from the lSI Test requirements pursuant to NRC letter dated April 18, 1981, from B. J. Snyder to G. K. Hovey regarding Mini Decay Heat Removal System Surveillance Requirements.

Mrn-Vl Mrn-V2 MDH-Vll Mrn-VI8 MDH-V19 TABlE II VALVES 'ID BE TESTED F -Functional Test T -Stroke Test M -MOnthly Testing Q -Quarterly Testing TABlE .. Valve No. Size Valve Function rest SW-Vl A&B Oleck 8" SH-P-l A&B Discharge Check --------C F Q Valve SW-V28 A&B Gate 3/4" Lube Water to Screen Wash Solenoid B F Q Pumps DH-VI19 Angle 12" Borated Water Storage Tank Vacuun C F Q Breaker DH-V227 Angle 12" Borated Water Storage Tank Vacuun C F Q Breaker Check 10" NS-P-LA/B/C Discharge

---

C F Q A,B&C A&B Gate 1" To Instrunent Air Comp. Solenoid B F Q Motor A&B Gate 1" Fran Instnment Air Canp. Solenoid B F Q Motor NR-Vl Check 24" NR-P-l A-D Discharge Check --------C F Q A,B,C&D Valve NR-V33 A Check 8" River Water to Emergency

---

C F Q Diesel Generator Cooling NR-V34 B Check 8" River Water to Emergency

---

C F Q Diesel Generator Cooling NR-V39 A&B Butter-24" Diesel Generator Cooler Air B T Q fly Outlet NR-V40 A&B Butter-24" Inlet to Decay Heat Setvice Motor B T Q fly Coolers NR-V42 A&B Butter-24" Inlet to Decay Heat Service Motor B T Q fly Coolers NR-V82 A&B Oteck 3" Control Building River Water --------C F Q Booster Pump Discharge NR-V186 Globe 3/4" Lube Water to NR-P-l A,B,C&D Solenoid B F Q A,B,C&D TABlE II (Cont'd.)

Valve No. Size Valve FLmction _. __ ....... _ .. _ ... _--

Type of Test Test NR-V241 Butter-3" NR-P-3A/B,DischarRe Air B T Q fly RR-V25 Gate 6" Reactor Building Cooling Air B T Q A,B,C,E Coil Outlets to Evaporator Cooler DC-V6 A&B Check 12" DC-P-LA/B Outlet ---------C F Q AH-V14 A&B Cl'leck 4" Control Building Liquid ---------C F Q Cooler Pump Discharge AH-V28 A&B Butter-4" Cable Roan Fan Coil Unit Air B T Q fly \.Jater Inlet AH-V29 A&B Control 4" Cable Roan Fan Coil thlt Diaphran:n B T Q Hater Outlet AH-V32 A&B Control 3" Control R.oan Fan Coil Unit Water Inlet Air B T Q AH-V33 A&B Control 3" Control Roan Fan Coil Unit Diaphragsn B T 0 Water Outlet AH-V124 A&B Solenoid 1/4" Control Room Recirculation Solenoid B F Q InstrtJret1t Air TIlliI.t; 11 (u:mt "0.) Valve No. Size Valve Ftmction Operator Category Typ_e_

AlI-V125 MB Solenoid 1/4" Oontrol Room Recirculation Solenoid B F Q Instn.ment Air NR-V85 A&B Control 2" Control Building Mechanical Air B T Q Room Fan Ooi1 Inlet NR-V88 MB Butter-2" Downstream of Control Bldg.

B T Q fly Mechanical Room Fan Coils NR-V144 MB Butter-4" Upstream of l.iquid Chiller Air B T Q fly Condenser-Control Building NR-V145 AFtB Butter-4" Downstream of Liquid Chiller Temp. B F Q fly Condenser-Control Building Control AH-EP-S246 Soleroid 1/4" Controls Dampers D-4073, Soleroid B F Q MB ID 407SA, ID 4075B & ED 4075 AH-EP-5222 A Solenoid 1/4" Controls Damper D 4088B Solenoid B F Q AlI-EP-5227 A Solenoid 1/4" Controls NR-V85A Solenoid B F Q AH-EP-5245 Solenoid 1/4" Controls Damper D 4076 Solenoid B F Q AH-EP-S182 Solenoid 1/4" Controls NR-V144B Soleroid B&C F Q & Check AlI-EP-5205 Solenoid 1/4" Controls NR-V144A Solenoid B&C F Q & Check AH-EP-5210 Solenoid 1/4" Controls Damper D4092C Solenoid B&C F Q & Check AH-EP-S2l6 Solenoid 1/4" Controls Damper D4096A & Solenoid B&C F Q MB & Check AH-V32A " AH-EP-52l7 Solenoid 1/4" Controls Damper D4096B & Solenoid B&C F 0 A&B & Check AH-V32B AlI-EP-5222 B Solenoid 1/4" Controls NR-V85B Solenoid B&C F Q & Oleck AlI-EP-5227 B Solenoid 1/4" Controls Damper D4088A Solenoid B&C F Q & Check TABlE II <<(',-... , d. ) Valve No. Size Valve FUnction Ooerator Cater-ory Type of Test AH-EP-5235 Solenoid 1/4" Control AH-V2SA & Damper Solenoid B&C F Q A&B & Check D4074A AH-EP-5237 Solenoid 1/4" Control AH-V28B & Damper Solenoid B&C F Q A&B & Check D4074B AH-EP-5265 Solenoid 1/4" Controls Damper D409lA Solenoid B&C F Q & Check AH-EP-5266 Solenoid 1/4" Controls Damper D4091B Solenoid B&C F Q & Check NR-Vl16 Control 2" River Water Punp House Fan Solenoid B T Q A&B Coil Inlet NR-Vl17 Butter-3" River Water Punp House Fan B T Q A&B fly Coil Outlet NR-V2A,B, Butter-24" NP-PlA, B, C, & D Discharge Motor B T Q C, &D fly Valve NR-V234A, Check 1" NR PtJ11ll.ube Water Supply --------C F Q B,C, &D Check AH-EP-5356 A Solenoid 1/4" Controls NR-Vll6A Solenoid B F Q AH-EP-5358 A Solenoid 1/4" Controls NR-Vl16B Solenoid B F Q AH-EP-5356 B Solenoid 1/4" Controls Damper D5356 Solenoid B&C F Q & Check AH-EP05358 B Solenoid 1/4" Controls Damper D5358 Solenoid B&C F f1 & Check SPC-V4 A&B Check 2" SPC-P1 A/B Dicharge Check --------C F Q SPC-V6 Check 6" SPC to Res Check --------C F Q SPC-V32 Check 6" SPC to Res Check --------C F Q SPC-V40 Check 21t SPC to Res Check --------C F Q SPC-V7l Control 4" SPC-T-1 Discharge Control Motor B F M Valve 1-=.2. '" TABlE II (Cc . Valve No. Size Valve Function Operator category Type of Test SW-V20 Solenoid 3/4" Lube Water to SW Solenoid C F Q SW-V35 Check 3/4" Domestic Water Supply to --------C F Q DH-V5 A/B Gate 14" Borated Water Storage Tank to Decay Heat Remva1 Pumps M:>tor B T Q DH-V100 A&B Gate 12" Upstream of Decay Heat Rsoova1 PtIrq:> M:>tor B T Q DR-Vl02 A&B Gate 14" Upstream of Decay Heat M:>tor B T q Ramva1 PtI:tlJ DH-V103 A&B Check 10" Decay Heat Remva1 PtIrq:> Discharge Check ---------C F Q DR-Vl13 A&B Check 14" Borated Water Storage ---------C F Q Tank to Decay Heat Rsoova1 DH-V128 A&B Angle 10" Fran Decay Heat RenDva1 M:>tor B T Q Cooler DH-Vl93 A&B Gate 8" Decay Heat Rerooval Coolers M:>tor B T Q Crosstie 11: ...c./.(Af:

uzrT: 7U7.A!. ':h ""-AIStIIIC TO: eo!:! 1 IIUt D1lE: AiL.tfJ:;:.

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TMI-2 Cleanup Project Directorate Attn: Dr. W. D. Travers Director lIIIO

-I US Nuclear Regulatory Commission c/o Three Mile Island Nuclear Statj Middletown, PA -17057 l .-UHS -" QPU Nucl.ar Corporation Post Office Box 480 Route 441 South Middletown, Pennsylvania 17057-0191 717 944*7621 TELEX 84*2386 Writer's Direct Dial Number: (717) 948-8461 44l0-86-L-0075 DOCl.lTlent 10 0425A May 15, 1986

Dear Dro_ Travers:

I Three Mile Island N) J:TMI-2) Operating License No. oPR-73 Oocket No. 50-320 Relief From 151 Test Requirements for Category Band C Valves GPU Nuclear letter 44l0-84-L-0058, dated April 9, 1984, requested.relief from the Inservice Inspection (151) Test Requirements of ASME Section XI for various Category Band C Valves. Table II of the referenced letter identified those valves which GPU Nuclear had proposed to include in an 151 testing program subject to NRC approval of our request. Since the submittal date of that request, GPU Nuclear has submitted three (3) Technical Specification Change Requests (TSCR), i.e., Number 46, 49 and 51, as part of our Technical Specification Simplification Program, which are relevant.

The changes implemented by TSCR No. 46 and those proposed by TSCR Nos. 49 and 51, which are currently being reviewed by the NRC, significantly change the TMI-2 Technical Specifications.

Based on these TSCRs, GPU Nuclear has re-evaluated the need to conduct 151 testing on any of those valves identified in the reference.

Accordingly, based on the attached evaluation, GPU Nuclear requests that these valves also be exempted from the 151 testing requirements of ASt-£Section XI. GPU Nuclear Corporation Is a subsidiary of the General Public Utilities Corporation Dr. Travers May 15, 1986 4410-86-L-0075 Subject to your approval of the above request, it follows that TMI-2 should be exempt from all lSI testing requirements based on the current plant status. Therefore, exemption from the Inservice Inspection Program Requirements of 10 CFR 50.55a and the provisions of IWV-3410 and IWV-3510, for Category Band C valves is requested.

-Per the requirements of 10 CFR 170, an application fee of $150.00 is enclosed.

FRS/ROW/em!

Attachment Sincerely, /s/ F. R. Standerfer F. R. Standerfer Vice President/Director, TMI-2

Enclosure:

GPU Nuclear Corp. Check No. 00023374 ATTACHMENT 4410-86-L-0075 Nuclear Services Closed Cooling Water (NSCCW) System The NSCCW system was deleted from the TMI-2 Technical Specifications by NRC Amendment Of Order dated August 8, 1985. Thus, testing of the valves listed below is not required since the NSCCW system is no longer essential for plant safety. Valve NS-Vll A, Band C NS-V2l7 A and B NS-V2l8 A and B Function NS-P-l A, B, C Discharge Inlet to Instrument Air Compo Outlet to Instrument Air Compo II Decay Heat Closed Cooling Water (DHCCW) System The DHCCW system was deleted from the TMI-2 Technical Specifications by NRC Amendment Of Order dated August 8, 1985. Thus, testing of the below valves is not required since the DHCCW system is not essential for plant safety. Valve DC-V6 A and B NR-V40 A and B NR-V42 A and B Function DC-P-l AlB Outlet Inlet to Decay Heat Service Coolers Outlet to Decay Heat Service Coolers I Reactor Building Normal Cooling Water (RBNCW) System The RBNCW system has never a post-accident Technical Specifications required system; not essential for plant safety. Additionally, these valves are containment isolation valves per TMI-2 Surveillance Procedure 42l0-SUR-3244.0l, "Containment Integrity Verification

-Recovery Mode." Thus, the following valves should be re-classified as Category A valves which are exempt from lSI Testing per NRC letter dated April 27, 1981. Valve Function RR-V2S-A, B, C & E Reactor Building Cooling Coil Outlet to Evaporator Cooler IV Standby Pressure Control (SPC) System The SPC system was deleted from the TMI-2 Technical Specifications by NRC Amendment Of Order dated August 8, 1985. Currently, the SPC system can be utilized to recover from an RCS leak. However, the provides a secondary means of recovery; the primary means for recovering from an RCS leak would be to provide make-up by gravity feed from the BWST or by activation of the Reactor Building Recirculation System or use of both systems. Therefore, since the SPC system is not required for RCS make-up, it is not essential for plant safety. Thus, testing of the below valves is not required.

Valve SPC-V4 A and B SPC-V6 SPC-V32 SPC-V40 SPC-V71 Function SPC-P-l AlB Discharge SPC to RCS Check Valve SPC to RCS Check Valve SPC to RCS Check Valve ATTACHMENT 4410-86-L-0075 SPC-T-l Discharge Control Valve V Nuclear Services River Water (NSRW) System The NSRW (NR) valves and associated air handling (AH) valves listed in Table II of GPU Nuclear letter 44l0-84-L-0058 have been categorized in terms of the systems supported as follows: o Emergency Diesel Generator Operation o Service Building River Water Operation o Control Building Ventilation System o Control Room HVAC A. Emergency Diesel Generator Operation TMI-2 Techncial Specification Change Request (TSCR) No. 51, which was submitted to the NRC via GPU Nuclear letter 44l0-85-L-0135 dated July 31, 1985, proposed deletion of the emergency diesel generators from the TMI-2 Technical Specifications.

This proposal was based on a safety evaluation which demonstrates a high probability that recovery from a loss of off-site power can be accomplished within eight (8) hours during which time power would supplied by the station batteries.

Therefore, the safety evaluation states that the diesel generators are not required to maintain safe plant conditions.

The NSRW system supplies cooling water to the emergency diesel generators which, in turn, supply power to the NSRW pumps in the event of a loss of off-site power. Therefore, TSCR No. 51 proposed deletion of the NSRW system based on the justification that, subsequent to the deletion of the emergency diesel generators, the NSRW system will not be a safety related system; i.e., it will no longer service any Technical Specification required systems with the exception of the Control Room HVAC which is discussed separately in Section VII. ThUS, GPU Nuclear believes that the following valves do require testing pending NRC approval of TSCR No. 51: Valve NR-Vl A, B, C, D NR-V2 A, B, C, D NR-V33 A NR-V39 A and B NR-Vl16 A and B NR-Vl17 A and B Function NR-P-l A, B, C, D Discharge Valves NR-P-l A, B, C, D Discharge Valves River Water to Emergency Diesel Generator Cooling River Water to Emergency Diesel Generator Cooling ( Diesel Generator CooleOutlet River Water Pump House Fan Coil Inlet River Water Pump House Fan Coil Outlet NR-V186 A, B, C, 0 NR-V234 A, B, C, 0 AH-EP-5356 A AH-EP-5356 B AH-EP-5358 A AH-EP-5358 B ATTACHt.£NT 44l0-86-L-0075 Llbe Water to NR-P-l A, B, C, 0 Check Valves NR Pump Llbe Water SUpply Controls NR-Vl16 A Controls Damper 05356 Controls NR-Vl16 B Controls Damper 05358 B.

Building River Water (SBRW) System The only NSRW valve associated with the SBRW system is NR-V24l which is the discharge valve for the SBRW Booster Pumps NR-P-3 AlB. The SBRW system provides cooling water for the Service Building HVAC system which is not a Technical Specification required system. ThUS, testing of NR-V24l currently is not required; the SBRW system is not essential for plant safety. C. Control Building Ventilation System This system provides ventilation for the Cable, Battery Switchgear, and Mechanical Equipment Rooms in the Control Building.

The purpose of this system is to provide cooling water to the associated equipment in order to avoid degradation in severe summer conditions.

The safety evaluation for TSCR No. 51 states that no significant equipment degradation will occur during the eight (8) hours conservatively assumed necessary to restore off-site power. ThUS, justification for not requiring lSI testing of the following valves is consistent with the rationale stated in Section V, Subparagraph A. Valve NR-V82 A and B NR-V85 A and B NR-V88 A and B NR-V144 A and B NR-Vl45 A and B AH-V14 A and B AH-V28 A and B AH-V29 A and B AH-EP-5l82 AH-EP-5205 AH-EP-5222 A AH-EP-5222 B AH-EP-5227 A AH-EP-5227 B AH-EP-5235 A and B AH-EP-5237 A and B AH-EP-5245 AH-EP-5246 A and B Function Control Building River Water Booster Pump Discharge Control Building Mechanical Room Fan Coil Inlet Control Building Mechanical Room Fan Coil Outlet Inlet to Liquid Chiller Condenser

-Control Building Outlet from Liquid Chiller Condenser

-Control Building Control Building Liquid Cooler Pump Discharge Cable Room Fan Coil Unit Water Inlet Cable Room Fan Coil Unit Water Outlet Controls NR-V144B Controls NR-V144A Controls Damper 04088B Controls NR-V85 B Controls NR-V85 A Controls Damper 04088A Controls AH-V28 A and Damper D4074A Controls AH-V28 B and Damper 04074B Controls Damper 04076 Controls Damper 04073, 10 4075A, 10 4075B, and ED 4075 VI Screen Wash (SW) System ATTACHt-£NT 44l0-86-L-0075 The SW system is designed to provide flushing water for the mechanical trash racks and traveling water screens which provide water filtration for the NSRW pumps. However, since flushing of the racks can be performed manually, t,he operability of the screen wash pumps is not a requisite for operation of the NSRW pumps. Additionally, as noted in Section V above, "GPU Nuclear also proposes deletion of the NSRW system from the TMl-2 Technical Specifications.

Testing of the following valves is currently not required since the SW system is not essential for plant safety: Valve SW-Vl A and 8 SW-V20 A and 8

• SW-V28 A and 8 SW-V35 Function SW-P-l A and 8 Discharge Lube Water to Screen Wash Pumps Lube Water to Screen Wash Pumps Domestic Water Supply to Screen Wash Pumps

• These valves are secured in the "open" position; therefore, lSI testing is not required to verify their operability.

VII Control Room HVAC System TMI-2 TSCR No. 49, submitted to the NRC via GPU Nuclear letter 44l0-85-L-OllO, dated June 18, 1985, proposed deletion of certain functions of the Control Room HVAC System which require diesel generators in the event of a loss of off-site power. However, based on NRC concerns, GPU Nuclear letter 44l0-86-L-0033 dated February 26, 1986, proposed retaining operability requirements for this system in the Technical Specifications, and requested only that the requirements for back-up on-site AC power supply be deleted. This request was based on the results of analyses which indicate that probability of a" simulatenous occurrence of a Unit 1 LOCA and loss of off-site power is sufficiently low to be considered an incredible event. Therefore, emergency diesel generator power backup for the Control Room HVAC system is not required.

Additionally, TMl-2's unique condition is such that no actions are required to be taken from the Unit 2 Control Room to maintain the unit in a safe shutdown (i.e., continuous manning of the Unit 2 Control Room is not required "to maintain a safe shutdown condition).

Additionally, as previously noted, the Control Room HVAC is and will continue to be maintained operable in accordance with the TMl-2 Technical Specifications.

While the Technical Specification surveillance does not satisfy the lSI testing requirement, it is noteworthy that in granting GPU Nuclear an exemption from the lSI testing of Category A valves, which are the containment isolation valves in the case of TMl-2, the NRC based the exemption on the fact that these valves are maintained in accordance with the Technical Specifications.

ThUS, the basis for not testing Category A valves should also be applied to the following Control Room HVAC valves: Valve AH-V32 A and 8 AH-V33 A and 8 Function Control Room Fan Coil Unit Water Inlet Control Room Fan Coil Unit Water Outlet AH-V124 A and B AH-V125 A and B AH-EP-52l0 AH-EP-52l6 AlB AH-EP-52l7 AlB AH-EP-5265 AH-EP-5266 ATTACHMENT 4410-86-L-0075 Control Room Recirculation Instrument Air Control Room Recirculation Instrument Air Controls Damper 04092C Controls Damper 04096A and AH-V32A Controls Damper 04096B and AH-V32B Controls Damper 0409lA Controls Damper 0409lB VIII Decay Heat Removal System (OHRS) NRC Amendment of Order dated August 8, 1986, deleted the Technical Specifications requirements for the OHRS based on a safety analysis which concluded that forced borated water recirculation systems are no longer required in TMI-2's unique condition.

Accordingly, the TMI-2 Technical Specifications have been modified to replace the OHRS system with a Reactor Building SUmp Recirculation System and two (2) operable flowpaths downstream of the Borated Water Storage Tank (BWST) common drop line, i.e., gravity feed from the BWST. Thus, that portion of the OHRS which is not required for RCS make-up as a portion of the gravity feed from the BWST is not essential for plant safety. Therefore, testing of the following valves is not required:

Valve OH-VIOO A and B OH-Vl03 A and B OH-V193 A and B Function Decay Heat Removal Pump Crosstie Decay Heat Removal Pumps Discharge Check Decay Heat Removal Coolers Crosstie The following valves are associated with gravity feed from the BWST: Valve OH-V5 AlB OH-Vl02 AlB OH-Vll3 AlB OH-Vl19 OH-V128 AlB OH-V227 Function BWST to Decay Heat Removal Pumps Discharge from Decay Heat Removal Pumps to RV BWST to Decay Heat Removal Pumps Check Valves BWST Vacuum Breaker Valve Discharge from Decay Heat Removal Cooler to RV BWST Vacuum Breaker Valve GPU Nuclear does not consider the above valves to be essential for plant safety. In the event that gravity feed from the BWST cannot be established via the OHRS, gravity feed can be accomplished through either the SPC, MDHRS, or other alternative means, e.g., temporary hose connection via the Fuel Canal Cleanup (FCC) manifold.

Additionally, the analyses presented in the GPU Nuclear Seismic Design Criteria (

Reference:

GPU Nuclear letter 4410-85-L-0077, dated April 16, 1986), which has been reviewed by the NRC, demonstrated that a vessel draindown would not result in either a criticality or offsite exposures in excess of 10 CFR Part 100 guidelines.

Therefore, testing of the above valves is not required.

( ___ ______ ...... _ ** _:;-__ , .... _ * .v .. : * :: , -'.' .a! __ -c: /' I "". .. , ***** UIirrID ITATa NUCLEAR REGULATORY COMM'II'ON

....... TON. D.c. _ Decillber

19. 1984 Docket No. 50-320 --Mr. F. R. Standerfer.

Df rector Three Mile Island Unit 2 SPU Nuclear Corporation P.O. Box 480 Middletown, PA 17057

Subject:

Three Mile Island Nuclear Station, Unit 2 Operating License No. DPR-13 Docket No. 50-320 Technical Specification Change Requests 39, '41. 43 Recovery Operations Plan Change Requests 19, 20, 22 Exemption Request from 10 CFR 50.5Sa (Code Safety Valves) Exemption Request from 10 CFR 100, Appendix A and 10 CFR 50.36(3) (Seismic Instrumentatfon) , The Nuclear Regulator,y Commission has issued the enclosed Alendment of Order. Recovery Operations Plan Change Approval of Ex .. ption from the requirements of 10 CFR SO.SSa for e Safety Valves. and Approval of Exemption from the seismic instrwentation requi,..nts of 10 CFR 100, Appendix It. and 10 CFR 50.36(3).

The Itmendlllent of Order w.ich lIOdifies 8ny sections of the Proposed Technical

$pacifications (PTS) .. s requested by leneral Public Utilities Nuclear Corporation (SPUNC) in letters dated "anuar,y 12. 1183, Sept.ber 12, ,1183 and September 3D, " 1183. Other doc .. nts related to this request include: 'Recover,y Operations Plan (ROP) Changes w.ich tlllre reguested fn separate letters also elated January 12, Septenber 12, and s.pt.ber 3D, 1183. a ... quest for ex.ption fran the require-.. nts of 10 CFR 50.5Sa wfth ... spect to Code Safety Rel tef Valves fn a letter elated April 18, 1184. and a request for an ex.ption fran. the setllltc Iioriitoring regui ..... ts of 10 CFR SO.36(3) and 10 CFR 100, Appendix A. Pangraph YUa)"(3) in a letter elated April 18. "84. -, , As previously explafned

'n a letter issued by the staff on .JUly 17. '1184, your

• PTS and lOP change requests .re divided .nto two separate fssuaias.

The first fssuance ws 8de on .July 1" ,.84 and ws f..ecliately effective.

,... ltaff , has reviewed ,our-safety evaluations for the above doc_nts and concludes that , ftur requests addressed by this fssuance are acceptable wfth as discussed With your staff. PTS changes that are the .. bjeet of thb letter will become effecttve on January 7. '.5. ,....

to 10 cn SO.sSa, 10 en 50.36(3) and 10 cn 100. Appendix A. Para.raph

.... effeettye upon flsuance. , . .. ' .

( \ ( . . Mr. F. Since the F.bruary 11, 1980 Order '-posing the Proposed Technical cations is currently pending before the Atomic Safety and Lieensin, Board. the staff will be advising the Licensing Board of this Amendlent 0 Order through a Notice of Issuance of Amendment of Order and a Motion to Conform Proposed Technical Specifications in Accordance Therewith.

-Federal Register NoUc.s for the discussed is.suances are enclosed.

Copies of the related Safety Evaluation and revised pa,es for the Proposed Technical Specifications and the Recovery Operations P an are also enclosed.

Enclosures:

1. lInendment of Order 2. Safety Evaluation

3. Proposed Technical Specification

'age Changes Sincerely

• ..-J!,. ..... J Bernard J. $n/le'r. Program Director Three Mile Island Program Office Office of Nuclear Reactor Regulation

• • Recovery Operations Plan Change Pages 5. Exemption from 10 CFR 50.55a ,. Exemption from 10 CFR 100, Appendix A, Paragraph Yl(a)(3) and 10 CFR 50.36(a) 7. Notice of Environmental Assessment and Finding of No Significant I_pact 8. Federal Register Notic.s ec: J. Barton R. Rogan S. Levin R. F ..... ennan d. Byrne Service Distribution List (see attached)

-\. Enclosure 5 UNITED STATES NUCLEAR REGULATORY COMMISSION In the Natter of I IENE_RAL PUBLIC UTILITIES NUCLEAR t CORPOIATION ) (Three Mile Island Nuclear Station, .. ) Dlit 2) ) EXEMPTION

1. Docket No. 50-320 GPU Nuclear Corporation, Metropolitan Edison Caapeny, .Jersey Central Powr and Light CoInpeny and Pennsylvania Electric Conpany (collectively, the licensee) are the holders of Oper.ting License MD. DPR-73, had . . authorized operation of' the Three Mile Island "'clear Station. Unit 2 (1MI-2) at po-.er llYel s up to 2772 lIIga-atts thennel. The facility, _ich is located in Londonderry Township, Onphin County, Pennsylvania, is a pressurized

.. ter reactor previously used fOr the commercial generation of electriCity.

-. 'y Order for Modification of Ltcense, dated .July 20, 1979, the licensee's to operate the

.. s suspended and the lic.nsee's authority .s limited to .inttnance of the in the present s .... tdown cooling lIOde (44 Fed. Reg. 45271). By further Order of the Director, Office of .clear .. actor Regulation, dated F.bruary 11, 1980, a new s.t. of fO .... ,. Hcense NqUir.ents .s illposad to reflect the post-accicJent condition of. the facl1 ity and to assure the continued .intenance of the safe, stable, long-tenn cooling condition of the facility (45 FId *. leg .:,1212) * . this Hc.nse provides, ""I other things, that it is subject to an "".5. regulatiOns and Orders of the C..tssion now or ...... fter in eff.ct. -.. _ ....*

. r*" . ,". F="" "."" II. On Apr11 18, 1984, General Publ ic Utilities

"'clear Corporation

('PUNt) . . .-requested an Exemption from the requirMents of 10 CFR 50.55a wtth -respect to Code Safety Valves for 1MI-2. Thfs provision of *the CClllllfssion's regulations currently requires that cOlftpoftents

... 1ch are part of the reactor coolant pressure boundary .-et the requ1r_nts for Class '. cCII'Iponents in Section III of the ASHE B011er and Pressure Vessel Code. As stated in Table 5.2*1 of the 1M1-2 Final Safety Analysis Report (FSAR), _ the lMI-2 Code Safety Valves (pressurizer) .et the requir_nts of ASME Section III, Article 9, Scnmer 1969 Addendllll.11-910.1 of Article 9 states that, -Each ,essel within the scope of the Code shall be

"'11e in service fr(lll consequences arising fran the application of steady state or transient conditions of pressure and (coincident) temperature

"'ich are in excess of the design conditions

• 11-910.4,11-910.5,11-910.6,11-910.7.

/ I and 1-910-8 state ,arious design and location requirtmlnts fOr the relief ,al,es. As stated in the FSAR, the 1MI-2 reactor coolant syst .. has a design pressure of 2500 ps1g with the pressurizer Code Safety Val,es re11ew1ng at approl1Ntely 2450 ps1g with a 690,000 pounds per hour capacity.

III. In the current syst .. configuration, with the reactor vessel Mad rlllOVed to .

defue11ng, *the code safety yalves are not useable and are not NqU1rlll in order to re11eve ,yst. pressure.

It is also the ltaff's opinion that the no,..l _1nteunce perfot'llld and presently ftIIded on tM .. val,es -

r... . ... , # /. '. / \; 5 Of: / c:. -. . (bench testing, blowdown, sl.l fnspection, Itc.) .,uld be .n .nKlSs.rY.inte-n.nce burden .nd result 1n .n unjustified r.diologic.l dose to the pl.nt .,rlters.

-As di.scussed it:' the concurrently saflty ev.luation, the .... ctor cool.nt .' S1Stem will .... in open to the re.ctor bunding, ** sphere throughout the recovery perfod. The s1St .. 'S configur.tfon fnherently provides ovlrpressure protection bec.use of the '.ck of

• closed 51st .. th.t fs nKess.ry for. significant pressure . bufldup. The only themaodynamic lVent that can occur 1n the RCS that .,uld have potential negative consequence fs a s.ysten heatup. lecause of the open vessel,.,e lVen

• heatup .,uld h.ve no pressure consequences unless the containment atmos-. phere increased.

Because of the vol ... of the contailDlnt

(.pproximately 2 x 10 6 cu. ft.) and the IIIIOUnt of decay .... t prlsent (approxi.tely 15 lew), any.stgn1ffcant contailDlnt pressure buildup .,uld occur over a period of d.1S, 1f not weks. Should a currently unforeseen lVent occur that could potlnt1.l1y cause pressure to fncrease or .ould requirl that

• pressur-fzld syst. be reestablhhlCl, the staff and the ltclnSle .,uld have suffkient response ti. to decide a course of action 1t be pl.cing the .. ad back on the vessel, 1nstalHng

• pressure rel1ef c_ponlnt, or llaving the systen as ts. Therefore, 1t fs the staff's opinton that a prlssure relief device for the RCS need not be tn pl.ce at this tf.. If a decfsion is Mde 1n the future to repressurize the ReS, a Mxi ... pressure rating and approprf.te overpresSure . '. protection

.. st be specift. fn a safety lValuation approved by the itaff, and fn procedures approved pursuant to Sectton 6.8.2 of the PTS. . .. -Itcause of the IUspension of the lfcl"sle's authority to oper.te facilfty tn other than the prlslnt recoveryllOde .s deftned fn the pr.opostel teetllical

-"

r , * \ '-..... : .' .

OF // , ..* 1._ b to specifications.

c.rtain of the regulations

..... fch are intended to appl, to no ... 1 operatfng plants. are sfmp1y inappropriate and. ure significantly

• .. are--unnecessary to protect the publfc health and safety. Cfven the unfque status of the plant in tenas of prf.r1 slstem t.nperature and pressure.

avaflable ffssion product inventorl.

the abflfty to cool the reactor without forced circulation (1oss-to-ambfent).

and the low decay heat rate ** 'ntenance of the fac11 fty with the eXllnptions grinted hereby will provfde an adequate level of safety. lYe Accordingly.

the Ccnnfssion has detenained that. pursuant to 10 CFR 50.12. an exemption is authoriZICI by law and will not endanger l1fe or .... operty or the CClllllOn defense and security and is otherwise in the publfc fnterest.

The Commission hereby grants an exemption to the requf,..ents of 10 CFR Part SO, A. Crfterion

2. 50. and 51. It is further determined that the .... ptfon does "ot authorfze a change in effluent types or total ..,.,nts "or an increase in 'powr level and will not result in anI Significant environmental tapact. In light of thfs deter-.ination and as reflected fn the Envfro,..ntal AssesSlllnt -aad Iotiee of finding of No Significant Environmental llpact prepared pursuant to 10 CFR 51.21 and 51.:30 through 51.32. issued concurrently it .. 5 * , . -_ .. -

r* r " ,.-. ( ,"','j. 7 -concluded that the instant action is insignificant fran the standpoint of enviromaental impact and an environmental fllpact statement MId .t be

• Effective Date: December 19. 1984 Dated at Bethesda.

Maryland Issuance Date: Dec_ber 19. 1984 FOR TM£ NUCLEAR REGULATORY COMMISSION Harold R. Denton. Director Office of NUclear .. actor Regulation

..

OF Ie.

• GPU Nude. Corporation Post Office Box 480 Route 441 South Middletown.

Pennsytvania 17057-0191 717 .... -7621 1141 Program Of'fice Attn: Dr. B. J. Snyder Program Director US Nuclear Regulatory Commission .ash1rgton, DC 20555

Dear Dr. Snyder:

TELEX 84-2386 Writer's Direct Dial Number: (717) '48-8461 44lo-84-L.-OO51 Docunent 10 048()J April 18, 1984 Three Mile Island tU:1ear Station, lhit 2 (1l4I-2) Q:lerating Ucense No. DPR-73 Docket No. 50-320 Exemption Reql.l!st from 10 a:R SO.558 with respect to Code Safety Valves Your letter of January 13, 1984, which provided conrnents on various Technical Speci fication and Recovery l>>erations Plan Change Requests required that GPUNC sutJnit a specific rellef from the ASME Cede requirenents for safety valves of 10 a:R Part SO. Based on the attached justification GPlt£ requests an exetqJtion from 10 CFR .5O.55a. As this request is stbnitted in conjunction with Ted'vUcal Specification Change Request No. 39, no additional fee is required.

Please call Mr. J. J. Byrne of my staff if you have any questions on this infcmnation.

EEKI JJ3! jep Attachment S1n:erely, lsI E. E. Kintner E. E. Kintner Executive Vice President cc: Deputy Program Director -1MI Program Office, Mr. L. H. Barrett GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation

• \ .lJSTIFlCATION FOR DELETING FOR ASME CODE SAFETY VALVES As stated in -Table 5.2-1 of the lNI-2 Final Safety Analysis Report (FSAR), the lNI-2 Code safety valves meet the requirements of ASME Section III, Article 9, Sunner 1969 Addendum.

This subsection of the code requires in Paragraph N-9l0.l that each vessel within its scope be protected while in service from consequences arising from the &plication of steady state or transient conditions of pressure and . (coincident) temperature which are in excess of design conditions.

Paragraph N-9l0.4 of this subsection of the code amplifies this section in that it requires the total capacity of those pressure relief devices be sufficient to prevent a rise in pressure of more than above the vessel's design pressure at the antiCipated design temperature.

The code safety valves presently in the lNI-2 Technical SpeCifications were evaluated in the FSAR to meet these requirements while TMI-2 was operating, however, in its present condition as discussed in Technical Specification Change Request No. 39, the design conditions have changed and there is no longer a need to protect the lNI-2 Reactor Coolant System Pressure Boundary from an overpressure event. Therefore, based on this Article of the ASME CodeJ no overpressure protection is required at TMI-2, thus, an exemption from the AStE code requirements would not jeopardize the health and safety of the public. Document 10 0480U

.' * . * : "Gil, . . UNITED STATES U'U-.l .... ('\ NUCLEAR REGULATORY COMMISSION

-.1It Distribution

,. 0 :1 , I , i;,./ .. ' \ .... l* WASHINGTON.

O. C. 20115

/0 OF / 6. AI's 8.tl.,](I)

Subject:

• • • • April 27. 1981 Assip: d To: No. 50-320 Mr. Gale K. Hovey Vice President and Director of TMI-2 Metropolitan Edison Company P.O. Box 480 Middletown, Pennsylvania 17057

Dear Mr. Hovey:

The NRC staff has reviewed your request of April 18, 1980 (Met. Ed. letter TLL 176 from R. C. Arnold to Harold Denton) for relief from the Inservice Inspection Program requirements of 10 CFR Part 50.55a. Your application proposed that in lieu of complying with the requirements of 10 CFR Part 50.55a, testing/surveillance be performed in accordance with the Recovery hnical Specifications/Recovery Operations Plan issued with the February '. IE , 1980 Order. You determined that the provisions of IWA-2400 of Section XI l the ASME Boiler and Pressure Vessel Code, 1974 Edition, Summer 1975 Addenda wnich provide for extending the inspection interval for a period of time equiv alent to the length of the shutdown are applicable to TMI-2. We agree that th inspection interval should be extended for a period equivalent to the shutdown period. Due Date:

Dist. Arnold-AD.BG.

Barton-AD.Be.

Clark-PAR.

DeVine-AD.

BG. Elam-AD.BG.

Fenti-TR.259 Fuller-AD.Br..

Harding-TR.68 Herbein-TR.ll8 Heward-PAR.

Hockley-HEAR.

Hol%worth-EG&G Hovey-AD.BG.

Hukill-TR.l84 Kanzanas-PAR.

King-AD.Br..

Kunder-AD.BG.

Lacey-JCP&L Kanganaro-PAR.

SchUlauss-PAR.

Thorpe-PAR.

Tipton-PAR.

Wallace-PAR.

Walsh-PAR.

J Wilson-RlOl R Wilson-PAR.

DDCC-1MI DDCC-PAR *. Mf¥ G. AU,1I$(JItI 4', You also determined that testing of pumps during shutdown periods is not required as per IWP*3400 of the Code and stated that you intend to discontinue testing pumps which are no longer considered safety related. In general, we find the discontinuance of pump testing for those systems no longer considered safety related acceptable.

All pumps in those systems required to mitigate the quences of an accident and maintain the reactor in its present safe shutdown condition are included in and required operable by the Recovery Technical fications.

These pumps should be tested at least once per 31 days. These tests should include as a minimUm running for fifteen minutes and measurement of at least one of the following:

discharge pressure flow-rate or differential pressure.

Specific relief requests for individual pumpsare required to be submitted if the above criteria cannot be met. Your 'request for relief defined Category A valves (per IWV-2110 of the Code) as being exclUSively containment isolation valves and stated that these valves will not be full stroked exercised in order to maintain containment isolation but will be maintained according to the Recovery Technical Specifications.

We.agree that containment isolation valves which are closed should remain closed and not 'exercised.

However. if a containment isolation valve is opened and then closed \.

a period of time, 10 CFR Part SO. Appendix J. Type "c" testing will be to verify its containment isolation function.

e " I' , ri I, f;' " " " , r/ " I

• t ., I , \ w

( , Mr. Gale K. Hovey You also stated that Category Band C valves in systems out of service will not be tested as per the Code but will be tested according to the Recovery Technical Specifications.

We recognize that Category Band C valves in systems out of service need not be tested during the shutdown period. However, we believe that all Category Band C valves in safety related systems in service should be exercised at least once per 92 days where practical to determine their operational readiness.

Relief from the test requirements for Category Band C valves in safety related systems specified above will have to be submitted on an individual valve basis. Based on meeting the above requirements, we find that granting the specific relief stated herein is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest, and, therefore, grant the requested relief. However, you should note that this relief does not apply to the Mini Decay Heat Removal System since relief for it is being considered in a separate action. A copy of the Notice of Issuance is enclosed.

Enclosure:

Notice of Issuance cc: See attached list Sincerely.

---?Oq ...... ,1;£ -' _ Bernard J. Snyd t 'Program Director TM! Program Off ce Office of Nuclear Reactor Regulation

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-. UNITED STATES NUCLEAR REGULATORY COMMISSION DOCKET NO. 50-320 METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER AND LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY GRANTING OF RELIEF FROM ASME SECTION XI INSERVICE INSPECTION (TESTING)

REQUIREMENTS 13 Of I b 7590-01 The U.S. Nuclear Regulatory Commission (the Commission) has granted relief from certain requirements of the ASME Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components" to Metropolitan Edison Company, Jersey Central Power and Light Company, and Pennsylvania Electric Company in accordance with the provisions of 10 CFR ISO.55a. The relief relates to the revised inservice testing program for pumps and valves for Three Mile Island Nuclear Station, Unit 2, located in Oauphin County. Pennsylvania.

The ASHE Code requirements are incorporated by reference into the Commission's rules and lations in 10 CFR Part 50. The relief is effective as of its date of issuance.

The relief consists of exemption from the requirements for measuring certain parameters in the Pump Testing Program and revised schedules for conducting valve stroking tests in the Valve Testing Program. Relief was also granted from the requirement for inservice inspection of Class I, II, and III components for a period equivalent to the length of the TMI 2 shutdown.

The request for relief complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations.

The Commission has made appropriate findings as required by the Act l* and the Commission's rules and regulation in 10 CFR Chapter I, which are set forth

. . . . . 7590-01 fn the letter granting relief. Prior public notice of this action was not required sfnce the granting of this relief from ASHE Code requirements does not involve a significant hazards consideration.

The Commission has determined that the granting of this relief will not result in any significant environmental impact and that pursuant to 10 CFR §51.5(d){4) an environmental fmpact statement or negative declaration and environmental impact appraisal need not be prepared in connection with this action. For further details with respect to this action, see (1) the request for relief (2) dated April 18, 1980, and (2) the Commission's letter to the licensee dated April 27, 1981.

• These items are available for public inspection at the Commission's Public Document Room, 1717 H Street, N.W., Washington, D.C. 20555 and at the Government Publications Section, State Library of Pennsylvania, Education Building, Commonwealth and Walnut Streets, Harrisburg, Pennsylvania 17126. A copy of item (2) may be obtained upon request addressed to the U.S. Nuclear Regulatory Commission, ,Washington, D.C. 20555, Attention:

Director, TMI Program Office. Dated at Bethesda, Maryland this 27th day of April. 1981. FOR THE NUCLEAR REGULATORY COMMISSION Three Mile Island Program Office Office of Nuclear Reactor Regulation

• ... ( Metropolitan Edison Company Post Office Box 480 Middletown.

Pennsylvania 17057 717 944-4041 Office of Nuclear Reactor Regulation Attn: Harold Denton, Director U. S. Nuclear Regulatory COmmission Washington, D.C. 20555

Dear Sir:

Writer's Direct Dial Number April 18, 1980 TLL 176 Three Mile Island Nuclear Station, Unit II (TMI-2) Operating License No. DPR-73 Docket No. 50-320 Exemption from the Requirements of 10 CFR 50.55a This letter is written to formally request relief from the requirements of *10 CFR 50.55a concerning the Inservice Inspection Program. As a result of the accident which occurred on March 28, 1979, the unit will be shutdown for an extended period of time (at least until 1984). In lieu of the requirements of 10 CFR 50.55a, testing/surveillances will be perforoed in accordance with the Recovery Technical Specifications/Recovery Operations Plan as addressed in the February 11, 1980 Order. The enclosed justification provides the basis for this request. RCA:LWH:SDC:hah Enclosure cc: J. T. Collins B. Grier V. Stello Sincerely, lsI R. C. Arnold R. C. Arnold Sr. Vice President Metropolitan Edison Company is a Member of. the General Public Utilibes System ..

\ Enclosure 1 TLL 176 ff.fJE !, ,(\,f: This facility is required by 10 CFR 50.55a Section G, Paragraphs 1 and 4 to meet the inservice inspection requirements of Section XI of the ASME Boiler and Pressure Vessel Code and its addenda. The edition to which this facility ascribes is the 1974 edition with the summer of 75 addenda. As a result of the March 28, 1979 accident, the Unit II facility has been, and continues to be out of service. The out of service term has already exceeded one year and will continue until at least 1984. According to article IWA-2400, Inspection Intervals, paragraph (a) the inspection interval for Class I, II and III components may be extended for a period equivalent to the length of the shutdown provided that it is equal to or greater than one year. As of the end of March, 1980, we have met this criteria and are therefore requesting the extension of the inspection interval for articles IlVB, UIC and n,"D of Section XI of the ASME code. Article IWP-3400, Frequency of Inservice Tests, paragraph (a) allo\-1s for discontinuing the inservice tests of pumps during shutdown periods. It is our intention to discontinue testing pumps which are' no longer safety related. Those pumps that are safety related, in the facilities present condition, will be tested in accordance with the Recovery Technical Specifications and the Recovery Operations Plan. Article IWV-3000 deals with inservice inspection of safety related valves. Section IWV-3410, Valve Exercising Tests, paragraph(f), allows an exemption to the exercising requirements of category A and B valves in systems out of service while paragraph (b)(l), allows for exercising category A valves only as far as practical, if at all, during plant operation.

In this facility, category A valves are exclusively containment isolation valves. While many of the systems which these valves are a part, are out of service, the valves remain in service to maintain containment isolation.

These valves will not be tested as allowed for in section IWV-3410(f) and (b)(l). These valves will be maintained as required by the Recovery Technical Specifications and the Recovery Operations Plan.

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. c._ Mr. &ale K. Hovey Vice President and Director of TMI-2 Metropolitan Edison Company P.O. Box 480 Middletown, Pennsylvania 17057 APR 1 t r..." .. "' . ., II-tt/,II.3 PAGE /

SUBJECT:

MINI DECAY HEAT REMOVAL SYSTEM SURVEILLANCE REQUIREMENTS

REFERENCES:

1. Letter, from G. K. Hovey to J. T. Collins, TLL 645, December 9, 1980. 2. Letter, from G. K. Hovey to B. J. Snyder, LL2-81-0031, February 13, 1981.

Dear Mr. Hovey:

OF /2-t We have reviewed your letter of February 13, 1981 (reference 2), and have approved your request for relief from the requirements of Section XI of the ASME Boiler and Pressure Vessel Code in accordance with the provisions of 10 CFR, Part SO.55 (a)(g)(6)(i) with the following exceptions.

In addition to its deSign function as a decay heat removal system, other potential uses for the Mini-Decay Heat Removal System identified to date include a backup 18ans of Reactor Coolant System pressure control in the event of failure of the Standby Pressure Control System, or the inability of the SPC system to maintain RCS pressure and inventory during a gross reactor coolant system leak or small break loss of coolant accident.

Even though the MDHRS is one of several back-up .ades available, it is identified as the preferred mode in your existing approved procedures.

An accident analYSis for the MDHRS that was reviewed and approved by the NRC in our Amendment of Order, dated November 14. 1980, assumed MDHRS isolation by the system's main isolation valves (MDH-Vl, MDH-V2. MDH-V18, MDH-V19), and a cOlI'C)lete draining of the Mini Decay Heat Removal System's vol&.llle onto the floor and into the drains of the auxiliary building.

This analySis demonstrated that, provided the isolation valves perform their function as deSigned, the sequences of the postulated accident would be acceptable.

This accident analysis also discusses the possibility of electrical energization of all pressurizer heaters (1638 KW) resulting in a volumetric expansion of the reactor coolant and requiring a COII'C)ensating relief of 8.6 gpm. The MDHRS has an installed relief of 53.5 gpm. The NRC staff reviewed your results of this potential overpres-..

. * ). . .. ( <. Mr. Gale K. Hovey r surization-event and perfonned an independent check. the results of which agreed with your conclusion.

The balance of the valves in the MDHRS are either for .. intenance convenience.

flow control. or instrumentation isolation.

While failure of any of these valves would necessitate the system being shutdown and isolated, they are not re11ed upon in the safety analysis.

The four isolation valves shall be inserv1ce tested in accordance with the quirements of Article IWV-3000 of Section XI of the ASME Code at least once within 31 days prior to the initial system startup and in accordance with the ASHE Code thereafter.

In addition, each valve in the main flowpath of the MDHRS shall be locked in its emergency use position and verified to be in that position at least once per 31 days. The only exception to this valve positioning requirement would be for testing the PlJll1)s.

after which the valves shan be returned to their emergency use position.

The four pressure relief valves, MDH-V4A, MDH-V4B, MDH-VaA, MOM-V8B, were tested prior to installation, the results of which have been reviewed by the staff and accepted.

Relief from additional testing is granted for these valves because of their passi,e role during normal system operation and ALARA considerations.

It should be noted that we do not concur with your reasoning that additional valve testing promotes valve degradation.

Our discussions with the diaphram valve vendor (ITT Grinnell) indicated that the valves in the MDHRS have sufficient conservatism built into their design to permit periodic cyc11ng in accordance with the requirements of the ASHE Code. Therefore, with proper operation.

periodic cycling of these valves would not have been expected to degrade their reliability or increase their failure probability.

Therefore, based on the above discussion, all Mini-Decay Heat Removal System valves are granted exemption from the inservice testing requirements of the ASHE Code with the exception of Main Isolation Valves MDH-Vl. MDH-V2. MDH-V18 and MDH-V19. Also, according to the above discussions on the present .odes in which the Decay Heat Removal System would be used and as stated in existing approved cedures, we grant the requested re11ef from Article IWP-3000 of Section XI of the ASHE Code and concur with your request to test each of the two MDHRS pumps on a 3 .onth staggered basis with each puq> being tested every 6 .onths. Per your justifications in reference

1. we agree with your proposal to leasure inlet sure, differential pressure.

and lubrication level according to the criteria of Article IWP-3000 and to .. asure vibration amplitude USing velocity via the Vibralarm as a lethod of also .onitor1ng bearing performance.

In s .... ry. the TMlPO concurs with the proposed testing schedule for the Mini-Decay Heat Removal System as discussed in reference

1. Full re11ef from Article IWV-3000 of the ASHE Boner and Pressure Vessel Code is granted for all valves in the MDHRS with the exception of the four .. in isolation valves MDH-Vl. MDH-V2. fl)H-VI8 and MDH-VI9. The four isolation valves shall be inservice tested in dance with the ASHE Code at least once within 31 days. prior to initial system startup. other than for PIlIP testing. and shall be inservice tested per Article IWV-3000 after initial startup.

):. , '-PAGE 3 Of /L Mr. 6ale K. Hovey It is our position that the above described and approved 1nservice testing t criteria for MDHRS pumps and valves and the positioning criteria shall be ianplemented (1.e., completion of the first periodic PllllP test and first verification of valve positions) within 7 after receipt of this letter. Sincerely, Bernard J. Snyd TMI Program Off ce Office of Nuclear Reactor Regulation

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("II , 0 ** PAGE. S OF (2. Mettopalitan Edison Company Post Office Box 480 Middletown, Pennsytvenil 17057 T.HI Prosram Office Attn: Mr. Bernard J. Snyder Prosram Director TMI Prosram Office U. S. Nuclear Resulatory Commission Washinston, D.C. 20555

Dear Sir:

February .13. 1981 LL2-81-0031 Three Mile Island Nuclear Station, Unit 2 (TMI-2) Operatins License No. DPR-73 o Docket No. 50-320 Mini Decay Beat Removal System Surveillance Requirements ( letter of December 9, 1980 (TLL 645) detailed our proposed veillance for the Mini Decay Beat Removal (MDHR) System. In your response of January 7, 1981, you requested that we perform additional surveillance on the HDHR System, which you believe to be more sistent with the intent of 10CFR 50.55 a (g) (6) (i) and with existins conditions.

After reviewing your letter and evaluatins the exiating plant conditions we are of the opinion that your reasons for requestins these additional surveillances are not consiatent with current intentions for the use for the HDHR system. Our orisiul intent in buildinS the MDHR syatem was to have a amall d1spoaable aystem for removal of the relatively hiSh levels of decay heat (approximately 900 KW) existing in the reactor core in the months immediately follow1nS the accident, bence avoiding the need to operate the installed Decay Beat Removal System, and eliminating tbe potent1ar for radiation expoaure to peraonnel and l .. kase of highly contaminated reactor coolant into the Auxiliary Buildins.

In the t11De it baa taken to 'build and license the MDHR. .,atem. the decay beat lenerated by the core has decayed to approximately 45 EW. 1oaa-to-.. bient cool ins has 'been demonstrated to 'be fully capable to .. inta1n core cooling. Thus, MDHR 1a not presently needed for core cool1ns and is simply one of several .odea available to provide core cooling if desired. Another potential use of the HDHR system is to provide a 'back-up .. ana of ,ctor Coolant System presaure control in the event of failure of tbe ; syata, 'but iD this MDBJl is a,a1n simply one of aeveral lback-up .odes ava11a'ble, (iDcludiDR the Decay Beat System) aDd bas been 1Dcluded iD plant procedurea for our coaveD1ence.

'l'HI-2 DiatribuUoa D1It. C X Arnold-AD.IC.

v Arnold-PAR.

J' lartoa-AD.IC. , Clark-PAR

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aD-AD.IC.

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BudiDa-n.

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.; BevaH-PAR.

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.' Buk1ll-n.184 bD&uas-PAR.

kiDS-AD.IC.

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Thorpe-PAR.

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... . -.. -LJ..-'-O,L-uu,),L Of /2-For these reasons, we have concluded that the MDHR ayatem i. not required in order to protect the public and health and safety. We presume that NRC concurs in the judgement, by vertue of the fact that NRC considered public health and safety to be adequately protected prior to the time MOHR became operational.

Furthermore, we have performed an accident analysis on the MOHR system Which was presented in Technical Specification Change Request No. 24b and Which was by the NRC. This analysis assumed isolation of this system by the ay,tem iaolation

valves, V2, V18, and V19, and a complete draining of the fluid in the MOHR ayatem onto the floor of the Auxiliary Building.

This analysis determined the off-aite effects of the accident, and confirmed that the health and aafety of the public was not jeapordized.

In light of the above discussion, we have reevaluated our original aub-'mittal with respect to your letter of January 7, 1981. We agree that some form of periodic testing of the MOHR pumps prior to system tion is appropriate, as a matter of good engineering practice.

Bowever, we believe such testing, although referenced in Section 4.7.3.3 of the Recovery Operations Plan, is not required in order to conform with any specific article pertaining to safety code component test requirements, as the MOHR system was not designed to be a safety-related system. On this basis. the inservice testinR reauirements of Article rwP-1000 nf Section XI of the ASHE Code is not applicable.

We can and will test each MOHR Pump in a recirculation mode with the recirculation valve in a specified position.

This test will allow us to monitor Inlet Pressure, Differential Pressure, Vibration Amplitude, and Lubricant Level in the manner specified in TLL 645. MOHR flow rate cannot be measured in this test due to the location of the flow rate instrument.

This test will be performed.

using uncontaminated.

unborated water (to prevent premature seal degradation), on a staggered basis so that each pump will be tested every six months, and a pump will be tested every three months. With respect to valve operability, we feel that additional valve testing is neither necessary nor appropriate.

since such testing can promote valve degradation.

Again, this is not a matter of public health and safety, but rather one of good engineering practice.

In our judgement, repeated unnecessary operation of valves will have a net detrimental effect on the readiness of MOHR for operation, ahould we choose to use the system, and therefore we do not intend to perform the additional valve operability testing you proposed.

In summary. we will add the MDHR pump testing (as described above) to the surveillances discussed in TLL 645. In our opinion, additional testing beyond that point would be an unreasonable burden on our limited resources, and could provide no benefit to the health and safety of the .* public. We wish to reiterate that our incentive for performing such surveillances is not to comply with any procedural requirements applicable to safety related system (which MOHR is not), but simply to provide us with reasonable assurance of system availability.

'-.. GKH:JJB:djb Sincerely, ,I, *. K.BOVI( G. It. Bovey Vice-President and Director, 'l'Ml-2 cc: L. Barrett, Deputy Director-'l'Ml Program Office

<# ** ----.... --.....

12-. * [gO =t: rIa iJlJ Mttropoli1an Edison Company Post Office Box 480 Middletown.

Pennsylvania 17057 \ ( <. -.. All' Writer's DiNe_ Dial Nu SUB ::r:. "'\ -------... * -v' TMI Program Office December 9. 1980-. TLL 645 Attn: Mr. John T. Collins, Deputy Director U. S. Nuclear Regulatory Commission c/o Three Mile Island Nuclear Station Middletown, Pennsylvania 17057

Dear Sir:

Three Mile Island Nuclear Station, Unit 2 (TMI-2) Operating License No. DPR-73 Docket No. 50-320 Recovery Operations Plan, Surveillance lequirements Relief Request This letter is written to formally request relief from Section XI of the Boiler and Pressure Vessel (B&PV) code and applicable Addenda in accordance with 10 CPR 50, Section 50.55a and Section 4.7.3.3 of the Recovery Operations Plan for specific Mini Decay Beat Removal System (MOHRS) inservice inspection criteria.

The enclosed evaluation details the MDHRS Inservice Inspection ments that we propose, the relief requested and provides a fication for each request. Your approval of these proposed surveillance requirements is requested.

GICH:JJB:dad cc: Bernard J. Snyder Inclosure Sincerely, ,./ ClI.BOVIY G. K. Hovey Vice-President and Director, TMI-2 (assnd. to) It It It It It It It Arnold-AD.Be Arnold-PAR.

Barton-AD.Be Clark-PAR.

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'-12/80 PAGE Of TMl-2 bcc Ust (external distribution)

Mr. E. L. Blake, Jr. Shaw, Pittman, Potts & Trowbridge 1800 M Street, M.W. Washington, D. C. 20036 Mr. T. F. Hartley, Jr. Marsh and McLennan, tnc. 1221 Avenue of the Americas New York, New York 10020 Mrs. Pat Higgins Edison Electric Institute 1111 19th Street, N.W. Washington, D.C. 20036 Mr. George Kulynych Babcock and Wilcox, Inc. P.O. Box 1260 Lynchburg.

Virginia 24505 U. S. Nuclear Regulatory Commission clo Document Management Branch Washington, D.C. 20555 Dr. Steven Long, Director Power Plant Siting Program Department of Natural Resources Taves State Office, Building B-3 580 Taylor*Avenue Annapolis, Maryland 21401 Mr. Thomas Gerusky, Director Bureau of Radiation Protection PA Department of EDvironmental Resources Fulton National Bank Building Harrisburg.

Pennsylvania 17120 Mr. Kent Hamlin American Nuclear Insurers The Exchange -Suite 245 270 Farmington Avenue Farmington.

Connecticut 06032 LERs only Hr. Bill Lavallee Nuclear Safety AnalYSis Center P.O. Box 10412 Palo Alto. California 94303

• 1

.. f ( * ,

1 u( /"2-J. T. Collins Page 2 Recovery Operations Plan, Surveillance Requirement Relief Request 4.7.3.3 of the Recovery Operations Plan requires each Hini ;Decay Beat" Removal System (KOHRS) pump and valve in the flow path to be in with the inservice inspection requirements of Section Xr-of the Boiler and Pressure Vessel Code except where specific written relief has'been granted by the Commission.

This request is submJtted in accordance with this requirement.

The relief requested is broken into two parts. Part A requests relief from the aonthly testing requirement for the KOHR pumps. This particular request is to perform this test semi-annually after the KOHRS has been used for reactor decay heat removal instead of .onthly as specified in the ASHE code. Part B of this request details the apecific testa required by the ASHE code and the relief requested from three of these teats, i.e Vibration Amplitude .. nt, Bearing Temperature Heasurement, and Valve Exerciae Testing. The goal of the first two tests is .et by using an alternative

.. alurement using installed vibration measuring equipment.

The third te.t involving valve exerci.ing, need only be perforaed on the aystem ilolaton valves to ensure proper KOHRS operation and therefore testing of additional valves is unnece.sary.

REASON FOR CHANGE A) Periodicity of KOHR Pump Testing The testing requirements of Section Xl of the ASHE B&PV Code requires inservice te.ting of each pump .onthly. we are requesting that this periodicity .. nt be relaxed to semi-annually after the KOHRS has been used for reactor decay heat removal. Vse of the MOHRS il only one of aeveral .odes available to .. intain adequate core cooling. Therefore monthly te.ting of the MOBRS pumps il not Dece ** ary becaule the MOHRS il not ellential to reactor lafety. Additionally telting the MOHR pumps .onthly in a borated water envira.ent increales the chance of .. ture leal degradation, and as the KOHR pumps will becoae highly contaminated by the primary Iyltem water, repairs to the pump. would relult in lignificant perlonnel

• * ) Specific KOHRS Pump and Valve Inlervice Inlpection Requir ... ntl. In addition to thil periodicity requir ... nt, Articlel lWP-3000 and lWV-3000 of Section Xl of the ASHE B&PV Code lpecify particular Inlervice Inlpections required for pumpi and valve.. The Inlervice teltl applicable to the MORRS pumPI and valvel and our .. thod. of telting are lilted below. Specified Telt Requir ... nt. Coapliance Hethod Inlet Prellure.

Pi Per lWP-3000

• , .

• PAGE /0 OF 12-J. T. Collins P.ge 3 Recovery Oper.tions Pl.n. Surveill.nce Requirement Re 1 ief leque.t Differenti.l Pre ** ure, P F10li R.te. Q Vibr.tion Amplitude, V Bearing Teaper.ture.

Tb Proper lubricant level 2) C.tegory A & B Valves V.lve Exercise Te.t Valve Leak Rate Test (Catelory A only) 3) Category C valves :: Per IWP-3000--Per IWP-3000 Me.sure Vibr.tion in velocity Rot .... ured Per IWP-3000 MDH-VI.V2.

VIS. VI9 only Per IWV-3000 Per IWV-3000 For those compli.nce .ethods which .re not identic.l to the ones .pecified in Articles IWP-3000 and IWV-3000 of Section Xl the ASHE B&PV Code. IOCFRSO.SSa exemption is requested .nd a justification

i. provided.

1) MDHR Pump Vibration

2) Article lWP-3000 of Section Xl of the ASHE B&PV Code .pecifies .n t.ble range for pump vibr.tion in mil ** di.placement.

The .y.tem that is installed on the MDHR pumps to mea.ure vibration.

the Vibralarm, .. asures vibration velocity.

Velocity me ** urement i. convertible to di.placement.

We intend to u.e, hOliever the velocity .. a.urement in place of the di.pl.cement aea.urement .pecified in the ASHE Code. Additionally the Vibralarm cont inuou.ly aonitor. pWlP vibr.t ion .nd provide ** n "ALERT" .nd a "SHUTDOWN" indication for the oper.tor.

Aa these indic.tions pond to the Alert Ranle .nd the lequired Action Ranae .pecified in the ASHE Code for pump vibr.tion ve viii u.e the.e indication

• • Which .re aonitored r.ther th.n t.kina aonthly .... urement **** pecified in the ASHE Code. MDHR Pump Be.rina Teaper.ture Article IWP-3000 of Section XI of the ASHE B&PV code .1.0 require ** n . Innnice te.t .... urina pump be.rina teaper.ture..

The MDHRS l.cks in.t.lled in.trument.tion to .... ure be.rina teaper.ture, but the Vibr.larm .y.tem c.n be u.ed to aonitor the ,...p.' be.rina hou.ina ** Me ** urina MDHRS pWlP be.rina teaper.ture would require entry into the MDBRS puap cubicle. Which i. c.lcul.ted to be

• 2-8 IIBr r.di.tion aone. to directly ... sure this par .. eter. Aa be.rina teaper.ture

i. not con.idered to be .t.ble until three .acce ** ive readina. taken at ten ainute interv.l.

v.ry by no .ore than 3%, this .... ur ... nt .ill re.ult in .ianificant expo.ure to personnel which i. unde.irable fro. an &LARA standpoint.

l . . J. T. Collins Recovery Operation.

Plan, Surveillance Requireaent Relief Reque.t ... II Pale 4 :: --.ince an installed Iy.tem exi.tl which adequately .onitors performance we intend to u.e it in.tead of takinl direct .ents of bearinl temperature.

3) MDHR System Valve Exercile Te.ts Article IWV-3000 of Section Xl of the ASHE B&PV Code requires the .. nce of a valve exerci.e test on Catelory A and B valves. Our intention however is to perform this test quarterly on the follovinl MDRRS valves only: MOH-Vl/2 MDHR System Inlet Isolation Valves MOH-V18/19

-MDHR System Outlet Iiolation Valves These are the valves that will be used to ilolate the MORRS in an emerlency and because of the high radiation levels in the MOHR cubicle. durinl Iystem operation these valves will have to remain Ihut while repairl are beinl .. de. Thus testinl other MDHRS valves is not needed to enlure Iy.tem reliability.

JUSTIFICATION OF ALTERNATIVES A) Periodicity of MDHR Pump Tests U.e of the MDHRS i. one of leveral .ode. available to .. intain adequate core coolinl. Therefore .anthly verification of the Itandby MORR Pump's operability

i. not vital to reactor lafety. Additionally te.tinl of the Itandby pump .. y caule .echanical leal delradation requirina pump repairl to be perfor.ed, and lianificant expolure to perlonnel.

In order to reduce the number of timel thele repair ... y be necellary but Iti11 .. intlin rea.onable al.urance that the Itandby MDHR pump will be operable, if needed, we intend to perform the in.ervice teltl on the Itandby MDHR pump lemi-annaally after the MDRRS has been uled for reactor decay heat re.aval * * ) Specific MDHRS Pump and Valve Inlervice Inlpection Requirement.

1) MDIlll Pump Vibrat ion Vibration il a ba.ic parameter for delcribina pump .echanical ilticl. Thil parameter il aealured on the MDHR pumPI by the Vibralarm vibration .onitorina IYltem which indicate.

vibration velocity.

Thil varies from Article IWP-3000 of Section Xl of the ASHE I&PV Code which lpecifiel a vibration dilplac ... nt .ealurement, but ttiil variation il acceptable becaule pump vibration i * .ealured with relpect to an eltab1ilhed and acceptable baleline.

Additionally, The Vibra1arm

...

• J. T. Collins Page 5 ... -* Recovery Operation.

Plan, Surveillance Requirement Re lief Reque.t -continuously aonitor. p .. p vibration and provides an "ALERT" and a "SHUTDOWN" indication for the operator, which corre.pond to the Alert Range and Required Action range .pecified in the ASHE Code. Therefore continuous aonitoring of vibration velocity of an operating MORR pump by the Vibralara .atisifie.

the intent of the vibration .ea.uraent cation in the ASHE Code. 2) MOHR Pump Bearing Teaperature The reason for .easuring puap bearing temperature is to give an tion of the .echanical characteristic.

of the MORRS puaps. AI there is no direct aea,urement of MORR Puap bearing teaperature, a substitute aeasurement

i. provided.

Thi ** ub.titute aeasurement

i. supplied by the MOHR pumps "Vibralara" vibration aonitoring .y.tem which continuously aonitors the pump'. bearing hou.ings for iapending failure .0 that corrective action can be taken prior to puap failure. U.ing this .ystem to aonitor the MOHR p .. p bearing. al.o reduce. per.onnel expo.ure to radiation becau.e entry into the MORR p .. p cubicle to take teaperature aeasureaents will not be required.

3) MORR System Valve Exercise Te.ts The accident analysi. perforaed on the MORRS and pre.ented in Technical Specification Change Reque.t Mo. 24b a .... ed MORRS i.olation by these .ystem i.olation valve. and a coaplete draining of the rest of the MORRS. Thi. analy.is demonstrated the off-site affect. of the accident did not jeapordice the health and .afety of the public, and therefore the.e are the only valve. that need to be te.ted. Additionally, not te.ting other .y.tem valve. will not adversly iapact .y.tem operation becau.e the hilh radiation level. in the MORR cubicles during .y.tem operation calculated to be 2-8 R/Hr, require the entire Iy.t .. to be Ihutdown and flu.hed prior to entering an MOHR cubicle to perfora _intenance.

• UNITED STATES NUCLEAR REGULATORY COMW,ISSION WASHINGTON, D. C. 20555 Docket"" No; 50-320 Mr. F. R. Standerfer Vice President/Director Three Mile Island, Unit 2 GPU Nuclear Corporation P.O. Box 480 Middletown, PA 17057

Dear Mr. Standerfer:

28, 1984 Three Mile Island Nuclear Station, Unit 2 Operating License No. DPR-73 Docket No. 50-320 GPUNC SeiSMic Design Criteria PAGE / OF sS-On October 4, 1984, you provided the staff with your interpretation of the GPU Nuclear General Project Desi9n Criteria (GPDC) as it relates to the Recovery Qual ity Assurance Plan (RQAP) .. aJld Regulatory Guide 1. 29. You state in that letter that GPU Nuclear's RQAP, Appendix C identifies Regulatory Guide 1.29 as being applicable for plant modifications that will renain after plant startup. You also stated that temporary recovery modifications will not have severe accident phenomena used as a design basis. You reference the staff's October 17, 1983 approval of Revision 2 of the QA plan as an indication of NRC concurrence with this philosophy.

In addition, you state that based on the above, an exemption from 10 CFR 50, Appendix A, Criterion 2 is not requi red. In response, you should note that on July 17, 1934, the staff issued GPU Nuclear an eXeMption from 10 CFR 50, Appendix A, Criterion 2 in relation to reactor building penetrations only (fee Enclosure, 49 F.R. 30384 July 30, 1934). On NoveMber 5, 1984, the staff also ssued a letter which re-eMphasized that the staff only deleted containment penetrations from Criteria 2 so that requirements could be applied on a case-by-case basis with respect to that type of structural component.

The-staff concurs with your statement that temporary recovery modifications do not have to meet design basis severe natural phenomena so long as; 1) the structure is temporary (all recovery modifications are not necessarily temporary in the staff's view); 2) a breach of that component by natural phenomena will not cause a radiological release in excess of 10 CFR 100 limits or the failure of that component will not compromise the ability to the reactor in a safe shutdown condition.

• tir. F. The staffis original statement in our August 10, 1984 letter is still applicable whereas " *** the GPDC must comply w1th Title 10 of the Code of Federal Regulations.

Staff concurrence with your design does not relieve you from 'the necessity of requesting exemptions from the code when appropriate." Unless the GPDC and the RQAP comply with 10 CFR, you must either request an exemption from the Code or revise the appropriate GPUNC document.

As previously stated, all other plant structures or components that are not classified as penetrations must still meet the requirements of Criteria 2 unless you justify in an exemption request, and you receive the staff's approval, that the consequences of a failure are acceptable.

cc: T. F. Demmitt R. E. Rogan S. Levin R. L. Freemerman J. Byrne A. W. Hiller Service Distribution List (see attached)

Sincerely, ... .( d . ."J-,.fl.

"-Bernard J.

Program Director Three Mile Island Program Office Office of Reactor Regulation

.' .:.. * \ .. TMI-2 LIST PAGE 3 01'. T1IaIIIs "'r ley Regional Region I U.S. Nuclear Regulatory 631 Park A"enue King of Prussia. PA 19406 John F. IIolfe. £iq ** Chaif'llln.

Judge 3409 ShQllerJi St. Chevy Chase. III. 20015 01'. Oscar H. Paris Judge Slfety and LiceRSin, IoIrd Panel U.S. Nuclear Regulatory washiftftOn.

D.C. 20555_. __ 01'. FNclericlt H. SIMIn AdII1 n1l trlti". Judge Slfety and Licensin9 Ioard Panel U.S. Nuclear Regulatory washtngton.

D.C. 20555 Karin W. Carter Assistant Attorney lieneral 505 E.ecuti"e House P.O. lu 2357 Harrisburg.

PA 17120 Dr. Judith H. Johnsrud En"iroftlental Coalition on Nuclear Power

• 33 Or lando Ave. . State College. PA 16801 IieOrge F. Trowbridge.

Esq. Shaw, pittllln, PotU and Trowbridge 1800 M. St ** IN. washington.

D.C. 20036 Slfety and Ltcensing Board Panel U.S. Nuclear Regulatory washington.

D.C. 20555 Slfety and Licensinl Appeal Panel U.S. Nucl .. r Regulatory washington, D.C. 20555 Secreury U.S. Nuclear Regulatory ATTN: Chief. Docketing' Service Branch washington.

D.C. 20555 Mr. Llrry IIocheftdoner Dauphin County P.O. 80. 1295 Harrisburg.

PA 17108-1295 John E. IHnn1ch, Chl1'1M1rson.

Dauphin County Iolrd of Dauphin County Courthouse Front .nd Market Streets Harrisburg.

PA lnOl Daup"'n County Office of _rgency Preparedness Court House. RooII 7 Front , Market Streets Harrisburg, PA 17101 U.S. Enviroftleftul ProtlCtion Agency ..,ion III Office ATTN: EIS Coordfnator Curtfs Buildfng (Si.th Floor) 6t11

• walnut St .... U Phnldalp1111.

PA lY106 1'hoIIIs M. lieNs ky. Df1"eCtor Bur .. u of Radf.t1on ProtlCt1on

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PA 17120 Mr. Edwin Kintner Eaecuti"e Vice President liener.l Public Utilities Nuclelr Corp 100 Iftterpace Plrkwy

• Parsippany.

N.J 0705

• t Lf -;h 5 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 Nov ... ber 5, 1984 1891 0;. Docket °No: 320 ,..,bc.. 7.p '" _A.? :: \':to'" ) ocr .. ( ---/ Mr. F. R. Standerfer, ° Three Mile Island Unit 2 GPU Nuclear Corporation P.O. Box 480 Middletown, PA 17057

Dear Mr. Standerfer:

The purpose of this letter is to clarify the staff's intent in issuing the July 17, 1984 Exemption from the requirements of 10 CFR 50, Appendix A, Criteria 2, 50 and 51. This intent was previously discussed with GPU licensing personnel prior to the issuance of the above document.

Because of the unique status of TMI-2, the staff did not believe that the requirements of Criteria 2, 50 and 51 were applicable in all instances and therefore issued the Exemption.

However. the staff does believe that requirements similar to Criteria 2. 50 and 51 may be applicable for some structural and component designs at TMI-2. The staff's action merely allowed for a case-by-case application of natural phenomenon design criteria and should not be interpreted as a pennanent deletion of associated design requirements. . The staff's analysis to support the Exemption was bounded by estimated source terms, release pathway area and air flow parameters as discussed in the Enclosure.

You should review the staff's analysis and provide the staff with a safety evaluation if a postulated failure of present or future penetrations would exceed the NRC's offsite dose consequence estimates.

All documents forwarded to the NRC for review or approval that discuss penetration modifications or containment integrity should also address natural phenomenon effects if applicable.

You should note that the staff's analysis primarily considered radiological releases that were filtered (via the auxiliary and fuel handling buildings) prior to release to the environment.

The only unfiltered release pathway considered by the staff was penetration 401, which for the scenarios sidered, would release a maximum of 20% of its activity directly to the environment.

\ PAGE 5 OF 35 Mr. F. Therefore, notwithstanding the July 17, 1984 Exemption to Criteria 2, 50 and tbe staff will also apply appropriate natural phenomenon design criteria on a case-by-case basis to procedures and design changes reviewed by the NRC in accordance with Section 6.8.2 of the Proposed Technical cations for penetrations and structures.

Enclosure:

As stated cc: T. F. Demmitt J. J. Barton R. E. Rogan S. Levin IJ. L. Freemennan J. Byrne A. W. Miller Service Distribution List (see attached)

Sincerely, .. J O..,J,.A--

ernard J.

Director Three Mile Island Program Office Office of Nuclear Reactor Regulation

.. Dr. n-I .... ",., ",fonal hl1011 I U.S. Nucl .. r Regulltory eo..f111011 131 Pa"k Avenue Itfnt of ,",ssfl. PA 19406 .10M F. 1IIo1f ** Eici ** C ... f,.lI.

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• RoIIert L. KnuPP. Esquirl Asliltant Solic1tor Knupp and Andrews P.O. lox

• 407 I. Front St. Harrisbu",.

PA 17108 John Levin. Elquire ...... ylv.nil Public Utilities to.I. '.0. loa 3265 "'rri Ibu",,'A 17120 Hono".bl.

"'rk Collan 112 E-E Ma1" capital .. Udhag HarrilburV_.A 17120 ..... Edw1n K1ntnar Eatcutiv.

V1cI President Itftarll Publfc Utilities Mllel ** r Corp. 100 Intarpacl Par., ""i"."y. KJ 07054 CF rAUL / TMI PROGRAM OFFICE FAILURE ANALYSIS FOR PENETRATIONS MODIFIED DURING THE RECOVERY PERIOD IN SUPPORT OF 10 CFR 50. APPENDIX A. CRITERIA 2, 50, AND 51 EXEMPTIONS INTRODUCTION AND ASSUMPTIONS Calcu1ations*were performed to estimate the offsite dose consequences of various accident scenarios involving breach of non-seismic containment trations.

The scenarios were selected to be representative of the types and conditions which could occur at TMI-2 during defueling activities.

The scenarios were chosen to be at the severe end of the spectrum, i.e., minor fires and cracks in the penetrations were not considered.

A limited number of representative isotopes and critical organs were used to simplify calculations.

This Simplification will set limiting conditions and account for greater than 90% of the dose. A more'complete source term which will account for greater than 98% of the offsite dose is being sent to RAB for dose assessment.

Dose conversion factors are from Regulatory Guide {RG} 1.109 except for transuranics.

Since RG 1.109 does not list values for Pu and Am, NUREG/CR-1972 was used for these isotopes.

Previously published (by NRC) values for short term (accident) atmospheric dispersion were used. In general the results are ratioab1e, i.e., one may double the fraction assumed to go airborne and it will double the dose. There are exceptions to this. Lengthening the duration of the events beyond two hours will not increase exposures in direct proportion due to meteorological sector averaging beyond two hours. Increasing the reactor coolant system leak will not increase doses in a linear fashion in that a stream of does not produce drop-lets as efficiently as a spraying leak. The assumptions used regarding release fractions are conservative, probably by more than one order of magnitude.

A list of references is provided following the last scenario.

Additional release fraction models and. experimental data from Battelle-Pacific Northwest Laboratories was utilized.

I ' , I f O<J1INANT FACTORS Several factors dominate the offsite dose consequences at the following accident scenarios.

Compared to scenarios for an operating reactor the isotopic mix is smaller and different.

In an operating reactor, volatiles (iodine) and noble gases are the primary dose contributors.

They are essentially absent at TMI-2. The particulates Which dominate dose calculations atTHI-2 have a tendency to settle which iodines and noble gases don't. They are essentially absent at TMI-2. In an operating plant one can usually generate peak containment pressures of 50-60 psig to provide a driving fOrce to propel the isotopes out of containment.

At TMI-2 this force is largely absent. HEPA filters are effective at removing particulates, even when assumed to operate at one thirtieth their design efficiency.

One must much lower decontamination factors for iodine and one for noble gases.

'---2-Plutonium 239, and other transuranics become the limiting isotopes due to the fact that their dose conversion factors are of magnitude higher than the fission products considered.

The actual curie releases are what smal1eT..

In operating reactors plutonium contributes relatively little offsite dose due to physical characteristics (i.e., solid) and the relative abundance of other isotopes.

The Whole body doses listed for Pu-239 and transuranics are actually organ doses which are "equivalent whole body doses." The dose consequences only consider the initial releases.

The scenarios considered would cause continual releases; however, after the first hour or two (which were factored in) the rates drop by orders of magnitude.

The licensee has several methods available to tenminate the release, during the initial period was assumed that these actions are not taken. These actions include plugging the penetration and/or starting a train of RB purge to put the building under negative pressure and provide a filtered releases path. Most of the scenarios including the limiting scenario are self-extinguishing.

Any long-tenm releases were considered improbable and small, and therefore were neglected.

SCENARIO I -FIRE ANO PENETRATION FAILURE In this scenario a seismic event fails various penetrations and also knocKs over a temporary lighting device which starts a fire in the reactor building.

Several fires were considered including contaminated surfaces (i.e., cable trays), a fire in the "0" ring supported by reactor coolant pump lubricating oil and a fire in the radioactive materials storage area. The fire in the radioactive materials storage area produced the highest dose consequences.

The stroage area is assumed to contain a variety of materials including tools, equipment (i.e., TV cameras, hoses, etc.). The materials would be enclosed in polyethylene (PE) or polyvinyl chloride (PVC) wraps or bags. A total of 8 Ci of Cs-137 and 4 C1 of Sr-90 is assumed in the storage area. This number is somewhat conservative in that personnel exposure ations would preclude accumulation of that much activity in one location.

The isotopic distribution is representative of the average expected over the defueling.

Conservatism in the total curie content covers the expected shift from low initial Sr fractions to perhaps more than 50% as defueling activities progress.

The PE, PVC, rags, paper and other wiping and wrapping rnlterials are assumed to contain 10% of the total activity.

The release fraction due to the fire is 5 E-2 for these materials.

The remainder of the activity is on tools and components.

The release fraction for these materials is 1 E-2. The total airborne activity generated in the reactor building by the fire is: (0.8 Ci x .05) + (7.2 Ci x .01) * .112 Ci Cs-137 (0.4 Ci x .OS) + (3.6 Ci x .01) * .056 Ci Sr-90

\.." PAGE '1 OF3 5 During the event the fire creates a 2 PSI RB overpressure (16.7 psia), contaminated air escapes through open penetrations until equilibrium is reached witn outside air. This release represents 12% (2+16.7) of the containment air or 240,000 cubic feet. If a penetration in the vicinity of the fire (i.e., 561 or 565) failed, "essentially all of the airborne activity could escape in the 240,000 cubic feet. If the penetration, which fails is remote from the fire location, the maximum fraction of airborne activity leaving the RB would approximate 12% due to mixing by the RB recirculation system. Assuming that penetration 561, 565, or both fail, if the auxiliary building ventilation system is operating it will remove 99% of the activity (accident assumptions for 99.97% efficient HEPA filters).

If the ventilation system is not operating, 90% of the activity will fallout in the auxil iary building (a large dead air volume) and 10% will exfiltrate from the building.

The worst case is with the ventilation inoperable and results in the release of 11 mCi of Cs-137 and 6 mCi of Sr-90. If penetration 401 fails (remote from radioactive material storage area) 12% of the airborne activity (13 mCi of Cs-137 and 7 mCi of Sr-90) could be transferred to the basement of the service building.

No fallout was assumed since this would not be a large dead air space. The 13 mCi of Cs and 7 mCi of Sr are assumed to be released directly to the environment.

This represents the limiting case since a combination of failed penetrations would lower the activity escaping through penetration 401. Dose calculations:

The astivity is all released within a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period (X/Q

• 6.8 3 E-4 sec/m). An adult at the exclusion boundary breathes at 1.2 m /hr. For Cs-137 (1 hr/3600 sec) (1.2 m 3/hr) 6.8 E-4 sec/m3) (0.0065 Ci/hr) (2 hr) (1 E12 pCi/Ci) (5.35 E-5 mrem/pCi)

• .16 mrem whole body dose (1 hr/3600 sec) (1.2 m 3/hr) (6.8 E-4 sec/ml) (0.0065 Ci/hr) (2 hr) (1 El2 pCi/Ci) (5.98 E-5 mrem/pCi)

• .18 mrem bone dose For Sr-90 (', hr/3600 sec) (1.2 m 3/hr) (6.8 E-4 sec/m3) (0.0035 Ci/hr) (2 hr) ( E12 pCi/Ci) (7.62 E-4 mrem/pCi)

• 1.21 mrem whole body dose (11 hr/3600 sec) (1.2 m 3/hr) (6.8 E-4 sec/m3) (0.0035 Ci/hr) (2 hr) ( E12 pCi/Ci) (.124 E-2 mrem/pCi)

• 19.7 mrem bone dose

16 OF 35 In Scenario I all the airborne activity from the reactor building escaped to the auxiliary building.

Therefore.

change in the number and size of the penetrations_

can't increase the potential release. SCENARIO I I -LEAKS AND SPI LLS In this scenario a seismic event causes the failure of penetration(s) and a leak in the reactor coolant system (RCS) or an RCS cleanup system. At the time of the leak the RCS activity concentrations are assumed to be elevated due to defueling activities.

The following concentrations are assumed 15 uCi/ml Cs-l37, 7.5 uCi/ml Sr-90. 1 uCi/ml Ce-l44. and 5 E-S uCi/ml Pu-239. In the leak of the processing system the leak is assumed to be at the pump outlet prior to demineralizers.

The leak rate is 25 gpm and the system is turned off (isolating the leak) after 1 hr. The 3 fraction becoming airborne due to spraying.

splashing.

and free fall is 10-. The resulting airborne activity is (25 gal/min) (60 min) (37S5 ml/gal) (.001) (activity conc Ci/ml). The results are .OS5 Ci Cs-l37 ** 042 C1 Sr-90 ** 005 Ci Ce-l44. and 2.S E-10 Ci Pu-239. If the leak occurs in the RCS it would be unpressurized (other than static head). The leak is assumed to continue until the water drains to the level of the reactor vessel nozzles the total volume is assumed to be 20,000 gallons. Due to the laral volume and lack of pressurization the fraction becoming airborne is 10

• For this case .11 Ci Cs-l37 ** 056 Ci Sr-90 ** OOS Ci Ce-144 and 3.7 E-10 Ci of Pu-239 become airborne.

Assuming simultaneous failure of several penetrations air could be drawn into the reactor building through penetration 401 and out penetrations 561 and 565 by the auxiliary building (AB) and fUel handling building (FHB) exhaust fans. If the highly contaminated air remained below the RB 347 ft. elevation, total airflow through the penetrations would not become a limiting factor. The air would pass through HEPA filters (accident OF of 100) and 1% of the activity would be discharged through the vent stack. A single failure of penetration 401 coupled simultaneously with the passage of a 1 psig low pressure front could result in a direct release pathway. This would result in the release of 6.S% of the containment air prior to reaching pressure equilibrium.

Due to the location of penetration 401 in the lower portion the RB it is assumed that 20% of the activity is entrained in the air (6.S% RB volume) which is released.

The resultant offsite doses (assuming short tenn release) are given by (1 hr/3600 sec) 1.2 (6.S E-4 sec/m-3) (1 El2 pCi/Ci) (Ci released) (dose conversion factor mrem/pCi)

• mrem.

(2.27 E5) (Ci) (DCF)

• mrem For--Cg...J37 PAGE ;/ Of (2.27 E5) (0.22 Ci) (5.35 E-5 mrem/pCi)

• 0.27 mrem whole body dose (2.27 E5) (0.22 Ci) (5.98 E-5) * .30 mrem bone dose For Sr-90 (2.27 E5) (.011 Ci) (7.62 E-4 mrem/pCi)

• 1.9 mrem whole body (2.27 E5) (.011 Ci) (1.24 E-2 mrem/pCi)

• 31 mrem bone dose For Ce-144 (2.27 E5) (0.0016) (2.3 £-5) * .01 mrem whole body (2.27 E5) (0.0016) (4.29 E-4) * .16 bone dose For Pu-239 (2.27 ES) (7.4 E-l1) (.514 mrem/pCi)

• 9 E-6 mrem4whole body equivalent (2.27 E5) (7.4 E-ll) (9.139 mrem/pCi)

• 1.5 x 10-mrem bone surface dose The sum of other transuranics which could potentially cause significant dose contributions is 84% of the Pu-239 dose. This is based upon their ratios in the fuel and the ratios of dose conversion factors. These isotopes are Pu-238, Pu-240, Pu-241 and Am-241. In Scenario II the assumptions for penetration 401 assumed 20% of the reactor building airborne escaped to the environment.

The worst case for multiple failures on any size in the auxiliary and fuel handling buildings would be 10%. Therefore, they cannot become limiting in this scenario.

SCENARIO III -DROPS OF CANISTERS The licensee's RCS cleanup and defueling methodologies have not been finalized.

The RCS cleanup system will probably be located outside the RB; however, it is included inside since this is a potential alternative.

A limit of 800 kg of fUel and rubble is assumed for fuel transfer canisters.

The limiting activity in RCS cleanup canisters is placed at 42,5'00 Ci Cs-137, 21,000 Ci Sr-90, 100 Ci Ce-144 and Pu-239 5 Ci. Cs and Sr would be limited by water throughout (media depletion);

Ce and Pu be limited by solids accumulation (plugging).

In the canister drop 10-is assumed to become airborne resulting in an airborne source in the RB of 4.25 Ci Cs-137, 2.1 Ci Sr-90, 0.01 Ci Ce-144 and .0005 Ci Pu-239. The canister drop would occur on the upper elevations, 10% of the activity is assumed to pass out penetration 401 (in the basement) due to stonn front passage. This results in a release to the environment of .425 Ci Cs-137, .21 Ci Sr-90, .001 Ci Ce-144 and 5 £-5 Ci Pu-239.

PAGE /2-Of 35 l. Resulting offsite doses are: Cs-137 Sr-90 Ce-144 Pu-239 Other TRU Whole Body (mrem) 5.10 36.02 .01 58.50 equiv. 49.00 equiv. 149.00 equivalent Bone (Bone Surface) (mrem) 5.70 591.00 0.01 103.00 87.00 787.00 equivalent With a limit of 800 kg of fuel in a canister, the activity per canister would 4 be 4350 Ci Cs-137, 5650 Ci Sr-90, 1570 Ci Ce-144, and 83 Ci Pu-239. With 10-airborne in the fuel canister drop accident the airborne source becomes .435 Ci Cs-137, .565 Ci Sr-90, .157 Ci Ce-144, and .0083 Ci Pu-239. This canister drop would also occur in the upper elevations of the RB with 10% of the airborne activity escaping through penetration 401. The release to the environment becomes .0435 Ci Cs-137, .0565 Ci Sr-90, .0157 Ci Ce-144, and .00083 Ci Pu-239. . Resulting offsite doses are: Cs-137 Sr-90 Ce-144 Pu-239 Other TRU Whole Body (mrem) .50 10.00 0.10 97.00 equiv. 81.00 equiv. 189 mrertl Bone (Bone Surface) (mrem) 0.60 160.00 0.20 1718.00 1443.00 3161 mrem (312 Rem) In Scenario III 10% of the reactor building airborne passed out penetration 401 this is greater than the upper limit for a series of large penetration failures on the auxiliary and fuel handling building.

SCENARIO IV -PYROPHORIC EVENT AND PENETRATION FAILURE In this scenario the seismic event simultaneously fails penetration 401 and many of the incore instrument guide tubes. Core rubble spills out into the reactor vessel cavity, the rubble is assumed to contain finely divided zirconium.

The moist zirconium undergoes a pyrophoric reaction When exposed to air. The total quantity of rubble is assumed to be 1000 Kg including 100 Kg of finely divided zircalloy (predominantly zirconium).

The fuel which is mixed with the zircalloy is already in an oxide fonn and will not react. However, due to the zirconium reaction, it is assumed that 10-4 of the fuel will become airborne particulates.

• \.. '-' PAGE/.5 Of 3 5 The release to the RB atmosphere is 0.5 Ci Cs-137. 0.65 Ci Sr-90. 0.18 Ci Ce-144 and 0.0091 Ci Pu-239. The following release mechanism is assumed; half is by the water from the leak and deposition due to the high density of the particles and that 201 of that remaining airborne escapes through the open penetration.

Resulting offsite doses are: Cs-137 Sr-90 Ce-144 Pu-239 Other TRU

• Whole Body (mrem) 0.6 11 0.1 113 equiv. 95 equiv. no ",rem Bone (Bone Surface) (mrem) 0.7. 186 0.2 2002 equiv. 1682 equiv. 3871 mrem (3.9 Rem) In Scenario IV 20% of the activity building airborne passed out penetration 401 which exceeds the limiting case for multiple large breaks in the auxiliary and fuel handling building penetrations.

CONCLUSIONS AND RECOMMENDATIONS The results of the scenarios show the worst case offsite dose commitments exceeding, but within a factor of 10 of 10 CFR 20 limits. The dose to the most exposed organ is well within (i.e., less than 2OS) the exposure lines for Whole bOdy dose in 10 CFR 100 (using ICRP 30 methodology).

The equivalent whole body dose is on the order of 10 CFR 20 annual limits. The dose commitments are all less than those for which evacuation would be recommended per NUREG-0654.

In general, if the auxiliary and fUel handling building ventilation systems are operating, the activity released would be less than 1%. If the ventilation is down the lack of driving head and building plateout will each limit releases to 10%. In aggregate they would limit releases to <5%. This analysis 1.s valid for up to 20 ft2 of penetrations in the auxil iary or fuel handling

• .J .,

• l \.. PAGE If 35 References

1. Chan, M.K.W., Mishima, J.

Characteristics of Combustion Products:

A Review of the Literature NUREG/CR-2658 PNL-4174 2. McGu1-re*,-

S. (1984). Personal COl'llllUnication

3. Mishima, J. (1966) Plutonjum Release Studies II, Release from Ignited, Bulk Metallic Pieces BNWL-357 4. Owczarski, et al., unpublished data (expected to be published as NUREG document late 1984) 5. Schwendiman, L.C.; Mishima, J.; and Radasch, C.A. (1968) Airborne Release of Particles in Overheating Incidents Involving Plutonium Metal and COmpounds

6. Sutter, S.L., et al. (1981) Aerosols Generated by Free Fall Spills of Powders and Solutions in Static Air NUREG/CR-2139, PNL-3786 7. Sutter, S.L. (1983) Aerosols Generated by Release of Pressurized Powders and Solut,ions in Static Air NUREG/CR-3093, PNL-4566 8. Sutter, S.L. (1982) Accident Generated Particulate Materials and Their Characteristics

--A Review of Background Information NUREG/CR-265l, PNL-4154 F'I-li3 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 Docket No. 50 ... 320 '." -".-;' Mr. B. K. Kanga, Director Three Mile Island Unit 2 GPU Nuclear Corporation P. O. Box 480 Middletown, PA 17057

Dear Mr. Kanga:

July 17, 1984

Subject:

Three Mile Island Nuclear Station, Unit 2 Operating License No. DPR-73 Docket No. 50-320 . PAGE /5 OF .3..5 Technical Specification Change Requests 39, 41, 43 Recovery Operation Plan Change Requests 19, 20, 22 Request for Exception to 10 CFR SO, Appendix A, Criterion 56 (Containment Penetration Design) Exemption Request from 10 CFR SO, Appendix A, Criteria 2, SO and 51 {SSE Requirements for Containment Penetrations}

Approval of Exemption to 10 CFR 50, Appendix A, Criterion 57 Approval of Alternate Design to 10 CFR SO, Appendix A, Criterion 55 The Nuclear Regulatory Commission has issued the enclosed Amendment of Order; Approval of Exemption from the SSE Design Requirements of 10 CFR 2, 50 and 51; Approval of Alternate DeSign to 10 CFR SO, Appendix At Criteria 55 and 56; Approval of Exemption from 10 CFR SO, Appendix A, Criterion 57; and Recovery Operations Plan Change Approval.

The Amendment of Order which modifies many sections of the Proposed Technical Specifications was requested by General Public Utilities Nuclear Corporation (GPUNC) in letters dated January 12, 1983, September 12, 1983 and September 30, 1983. Other documents related to this request include: Recovery Operations Plan Changes requested in separate letters also dated January 12, September 12, and September 30, 1983; a request for an exemption from 10 CFR 50, Appendix A, Criterion 56 and a request for exemption from Criteria 2, 50 and 51 in a letter dated April 24, 1984. The staff has divided your various requests into two separate issuances.

This issuance addresses those items that are immediately effective pursuant to 10 CFR 2.204. The justification for this type of issuance is discussed herein. The second issuance, which is addressed under a separate letter, discusses requested changes that are also being issued pursuant to 10 CFR 2.204 but are not immediately effective and allQws the licensee to demand a hearing within 20 days from the date of the Notice of Issuance.

\ Mr. B. K. Kanga The staff has reviewed your safety evaluations in the above documents and concluded that your requests are acceptable w1th minor changes as discussed with your staff. me staff has discussed with you your request for exemption from 10 CFR 50, ApPendix A, Criterion 56 and concluded that what you are actually seeking 1s an approval of alternate design. These discussions also revealed that this same alternate design should be applied to the requirements of Criterion 55, Reactor Coolant Pressure Boundary Penetrating Containment and based on this alternate design an exemption should be granted relative to Criterion 57, Closed System Isolation Valves. As previously stated, the Amendment of Order, Recovery Operations Plan Change Approval, the Approval of Alternate Design for 10 CFR 50, Appendix A, Criteria 55 and 56; the Exemption from 10 CFR 50, Appendix A, Criteria 2, SO, 51, and 57 are effective upon issuance.

Since the February 11, 1980 Order imposing the Proposed Technical cations is currently pending before the Atomic Safety and Licensing Board, the staff w111 be advising the Licensing Board.of this Amendment of Order through a Notice of Issuance of Amendment of Order and a Motion to Conform Proposed Technical Specifications in Accordance Therew1th.

Federal Register Notices for the discussed issuances are enclosed.

Copies of the related Safety Evaluation and revised pages for the Proposed Technical Specifications and the Recovery Operations Plan are also enclosed.

Enclosures:

1. Amendment of Order Sincerely, -I"i*-,)------Bernard J. ,Program Director Three Mile Is1 d Program Office Office of Nuclear Reactor Regulation

2. Exemption from 10 CFR 50, Appendix A, Criteria 2, 50 and 51 3. Approval of Alternate Design to 10 CFR 50, Appendix A, Criteria 55 and 56 4. Exemption from 10 CFR 50, Appendix A, Criterion 57 5. Safety Evaluation

6. Proposed Technical Specification Page Changes 7. Approved Recovery Operations Plan Change No. 20 8. Notice of Environmental Assessment and Finding of No Significant Impact 9. Federal Register Notices cc: J. Barton J. Byrne J. Larson Service Distribution List (see attached)

I \. l PAGE /7 OF 35 ENCLOSURE 2 UNITED STATES NUCLEAR REGULATORY In the Matter of GENERAL PUBLIC CORPORATION (Three Mile Island Nuclear Station, Unit 2) 1 ) EXEMPTION I. Docket No. 50-320 GPU Nuclear Corporation, Metropolitan Edison Company, Jersey Central Power and Light Company and Pennsylvania Electric Company (collectively, the licensee) are the holders of Facility Operating License No. DPR-73, which had authorized operation of the Three Mile Island Nuclear Station, Unit 2 (1l1I-2) at power levels up to 2772 megawatts thermal. The facility, which is located in Londonderry Township, Dauphin County, Pennsylvania, is a pressurized water reactor previously used for the commercial generation of electriCity.

By Order for Modification of License, dated July 20, 1979, the licensee's authority to operate the facility was suspended and the licensee's authority was limited to maintenance of the facility in the present shutdown cooling mode (44 Fed. Reg. 45271). By fUrther Order of the Director, Office of Nuclear Reactor Regulation, dated February 11,1980, a new set of formal license requirements was imposed to reflect the post-accident condition of the facility and to assure the continued maintenance of the current safe, stable, long-term cooling condition of the facility (45 Fed. Reg. 11292). The operating license provides, among other things, that it is subject to all rules, regulations and Orders of the Commission now or.hereafter in effect.

( II. By letter dated 1984, the licensee requested exemptions fran , . . 10 CFR 50, Appendix A, Criteria 2, 50, 51, and 56 regarding the design of containment penetrations after the removal of the reactor vessel head. Criterion 2 deals with design bases for protection against natural phenomena (i.e., earthquakes, tornados).

Criterion 50 relates to designing to' withstand pressure and temperature transients associated with loss of coolant accidents.

Criterion 51 pertains to fractures of the containment boundary.

Criterion 56 is concerned with containment isolation valves and is discussed in the NRC's Approval of Alternate Design issued concurrently herewith.

111. With respect to Criterion 2 the staff has evaluated the potential offsite dose of a containment isolation valve failure when challenged by natural phenomena.

The failure of the penetration by itself does not present a potential hazard unless accanpanied by a simultaneous event in the ment building which would cause the of radioactive material.

The staff has evaluated the potential offsite dose consequences of the failure of one or more coupled with a broad range of accidents in the containment building.

Calculations were performed to estimate the offsite dose consequences of various accident scenarios involving breach of non-seismic containment penetrations.

The scenarios were selected to be representative of the types and conditions which could occur at THI-2 during defUeling activities.

The scenarios were chosen to be at the severe end of the spectrum, i.e., minor reactor building fires or small cracks in the penetrations were not considered.

/

( A representative source term-for the offsite dose calculations was developed by the THIPO and" consequences were evaluated by the staff's Radiological Assessment Branch. With regard to Criterion 50, mechanisms and conditions which could produce temperature and pressure transients during a loss of coolant accident are essentially absent and will remain so during defueling.

This is due to the fact that the reactor coolant system will be at atmospheric pressure and tenperatures less than 110°F during defuel ing vs. -design temperatures in excess of 600°F and design pressures in excess of 2300 psig for an operating reactor. The staff also has evaluated other potential temperature and pressure producing mechanisms in coincidence with containment penetration failure. These include fires, failure of systens containing pressurized gases (i.e., nitrogen, air), and natural phenomena which cause pressure transients (i.e., tornadoes, .hurricanes, storm fronts). Potential penetration failures associated with the brittle fracture requirements of Criterion 51 are enveloped#by the evaluations performed for Criterion 2 and Criterion

50. The analyses performed for Criterion 2 and Criterion 50 included instantaneous total penetration failure in coincidence with various accident scenarios inside the reactor building.

Brittle fracture phenomena does not exceed instantaneous total penetration failure.

\. l \..' .. t ZO "-l 3 S The staff has evaluated the lxltential offsite dose consequences for all of the ,., -;:; . above worst case scenariOs.

The results of these scenarios show that the worst case offsite dose projections at the exclusion area boundary are within the exposure guidelines of 10 CFR 20. The effects of a penetration failure and simultaneous seismic event have been analyzed by the staff as stated in the above discussions.

The result of these occurrences have been shown to be within 10 CFR 20 guidelines.

Therefore the staff concludes that there is no undue risk to the health and safety of the public resulting from a seismic induced penetration failure, and it is the staff's opinion that the licensee's exemption request is justified.

The staff has determined that the post-accident status of the THI-2 facility presents exceptional circumstances relative to the applicability of the Commission's regulations.

Because of the suspension of the licensee's authori ty to operate the facili ty in other than the present recovery mode as defined in the proposed technical specifications, certain of the lations, which are intended to apply to normal operating plants, are simply inappropriate and, more significantly, are unnecessary to protect the public health and safety. Indeed, given the unique status of the plant in terms of primary system temperature and pressure, available fission product inventory, the ability to cool the reactor without forced circulation (loss-to-ambient), and the low decay heat rate, maintenance of the facility with the exemptions granted and the alternate design approved hereby will provide an equivalent level of safety. Furthermore, because of the condition of the plant and

( PAGt Z( l:f 5 the need to proceed with cleanup activities, literal compliance with the .>" regul ations fran wt{1*h 1"el ief is sought woul d present an unwarranted impediment.

IV. Accordingly, the Commission has determined that, pursuant to 10 CFR 50.12, an exemption is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest.

The Commission hereby grants an exemption to the requirements of 10 CFR Part 50, Appendix A, Criterion 2, 50, and 51. It is further determined that the exemption does not a change l. in effluent types or total amounts nor an increase in poWer level and will not result in any significant environmental impact. In light of this mination and as reflected in the Environmental Assessment and Notice of Finding of No Significant Environmental Impact prepared pursuant to 10 CFR 51.21 and 51.30 through 51.32, issued concurrently herewith, it was concluded that the instant action is insignificant from the standpoint of environmental i.mpact and an environmental impact statement need not be prepared.

Effective Date: July 17, 1984 Dated at Bethesda.

Maryland Issuance Date: July 17,1984 FOR THE NUCLEAR REGULATORY COMMISSION Harold R. Denton, Director Office of Nuclear Reactor

\.-'-. THREE MILE ISLAND PROGRAM OFFICE SAFETY EVALUATION FOR THE REVIEW OF , PAGE 2 Z-Of 35 Enclosure 3 ALTERNATE DESIGN FOR 10 CFR 50, APPENDIX At CRITERIA 55 AND 56 INTRODUCTION In a letter dated April 24, 1984, GPUNC requested an exception to certain design criteria for containment penetrations.

These criteria are stated in 10 CFR Part 50, Appendix A. Criterion

56. During staff discussions on this request, GPUNC stated that what they actually were seeking was an approval of an alternate penetration design which differs from those suggested in Criterion

56. The staff also had numerous discussions with the licensee relative to the penetration design requirements of Criteria 55 and 57 and concluded that the approval of alternate design should be applicable to Criterion 55 and an exemption should be issued to Criterion 57 (see Exemption to 10 CFR 50. Appendix A, Criterion 57 issued concurrently herewith.

In their letter, the licensee also requested an exemption from the seismic design requirements of Criteria 2. 50,-and 51. That request is discussed in an Exemption to 10 CFR 50, Appendix A, Criteria 2, 50 and 51 also issued concurrently herewith.

Following the TI1I accident.

thousands of curies of fission gases and active particulates were released from the fuel to the containment atmosphere.

Because of the unique condition of the TI4I-2 core and the amount of contamination resulting the accident.

the NRC imposed the requirement to maintain containment integrity to ensure that radionuclides inside the containment would not be released to the environment.

( PAGE 23' Of 35 -. In October 1979 *. the;tfrst of several containment penetrations was modified " to probe the containment interior to evaluate the extent of damage and to gather data to begin the cleanup. The penetrations were modified in accordance with NRC approved procedures.

The TMI-2 Proposed Technical Specifications also required that penetrations and operations that could affect containment integrity could be modified only by NRC approved procedures.

Since the 1979 accident, fission gases that were to containment have either decayed or have been purged from the containment.

Decontamination activities have also reduced airborne particulate contamination to below maximum permissible concentrations listed in 10 CFR 20, Appendix S, Table 1. In an evaluation associated with a Modification of Order dated April 9, 1982, the staff concluded that the maximum credible containment building pressure was approximately 2 psig. Calculated offsite doses resulting from a failed penetration in conjunction with a 2 pSig driving head and the associated reactor building airborne contamination were well below the limits of 10 CFR 20 and within the scope of impacts assessed in the "Final Programmatic Environmental Impact Statement" dated March 1981. DISCUSSION AND EVALUATION Criterion 56 provides guidelines for isolation valve configurations for piping that penetrates containment.

Criterion 55 provides guidelines for a reactor coolant pressure boundary that penetrates containment.

These guidelines also state that the licensee can. propose other containment isolation provisions that may be acceptable on another defined basis. Paragraphs (1) through (4)

( \ of Criteria 55 and 56 describe configurations that are preferred by the staff -, for a normal nucle(r-plant.

They are as follows: (1) one locked closed isolation valve inside and one locked closed isolation valve outside containment; or (2) one automatic isolation valve inside and one locked closed isolation valve outside containment; or (3) one locked closed isolation valve inside and one automatic isolation valve outside containment (a simple check valve may not be used as the automatic isolation valve outside containment);

or (4) one automatic isolation valve inside and one automatic isolation valve outside containment (a simple check valve may not be used as the isolation valve outside containment).

Criteria 55 and 56 were written for operating plant conditions and are generally applicable whenever the plant is operating.

in startup, hot standby, or during core alteration.

Presently.

the conditions at Unit 2 most closely resemble the standard criteria for cold shutdown (K ff<O.99, T <:.200°F).

e ave-During the normal cold shutdown mode for typical plants, containment integrity is

• normally not required and Criteria 55 and 56 are not normally applicable.

As previously stated, the staff correlated the shutdown condition of TMI-2 to that of a normal reactor in "cold shutdown." The staff also approved on this basis various penetration designs on the premise that if the plant were to enter a mode that *when compared to a normal plant would require containment isolation, either an alternate design or an exemption to Criteria 55 and 56 would have to be approved by the NRC. The licensee proposed several alternate penetration designs to the NRC staff -to support specific recovery operations.

The isolation feature common to all of the alternate deSigns includes two isolation valves outside of containment.

/

l PAGE 2S-Of. 3S In most cases," isolation valves are manual. Manual valves were found acceptable . ., .. . .... ", ". for isolation since all conceivable accident scenarios still permit access to the isolation valves. Isolation valves in containment as stated in Criteria 55 and 56 have not been because of difficulties (e.g., high dose rate areas) associated with accessibility for repairs or testing. It is the staff's opinion that the benign status of the reactor did not warrant the increased worker dose which would be incurred during the installation and testing of interior isolation valves. Therefore penetration modifications containing two . manual valves outside containment will be acceptable in satisfying Criteria 55 and 56 for all future recovery operations.

CONSI DERATI ONS We have determined that the alternate design approvals do not authorize a change in effluent types or total amounts nor an increase in power level and will not otherwise result in any significant environmental impact. Having made this determination, and, as reflected in the Environmental Assessment and Notice of Finding of No Significant Environmental Impact prepared pursuant to 10 CFR 51.21 and 51.30 through 51.32. issued concurrently herewith, we have further concluded that the change involves an action which is insignificant from the standpoint of environmental impact and that an environmental impact statement need not be prepared in connection with the issuance of this action.

( \ CONCLUSION

-' ".: ,. The staff has therefore concluded that the licensee's proposed penetration configuration is acceptable when considering the present condition and anticipated recovery activities at TMI-2. We have also concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the implementation of this change will not be inimical to the common defense and security or to the health and safety of the publ ic.

PAGE Z 7 OF, 3 5 ENCLOSURE 4 UNITED STATES NUCLEAR REGULATORY COMMISSION In the of GENERAL PUBLIC UTILITIES NUCLEAR CORPORATION

'" , . (Three Mile Island Nuclear Station, Unit 2) I ) EXEMPTION

1. Docket No. 50-320 GPU Nuclear Corporation, Metropolitan Edison Company, Jersey Central Power and Light Company and Pennsylvania Electric Company (collectively, the licensee are the holders of Facility Operating License No. DPR-73, which had authorized operation of the Three Mile Island Nuclear Station, Unit 2 (TMI-2) at power levels up to 2772 megawatts thermal. The facility, which is located in Londonderry Township, Dauphin County, Pennsylvania, is a pressurized water reactor previously used for the commercial generation of electricity.

By Order for Modification of License, dated July 20, 1979, the licensee's authority to operate the facility was suspended and the licensee's authority was limited to maintenance of the facility in the present shutdown cooling mode (44 Fed. Reg. 45271). By further Order of the Director, Office of Nuclear Reactor Regulation, dated February 11, 1980, a new set of formal license requirements was imposed to reflect the post-accident condition of the facility and to assure the continued maintenance of the current safe, stable, long-term cooling condition of the facility (45 Fed. Reg. 1l292). The operating license provides.

among other that it is subject to all rules, regulations and Orders of the Commission now or hereafter in effect.

( l. " .. .."

., II. PAGE Z1>

3.5 By letter dated April 24, 1984, the licensee requested exemption from

• 10 CFR 50, Appendix A, Criteria 2, 50, 51 and 56 regarding the design of containment penetrations after the removal of the reactor vessel head. Based on subsequent conversations with the licensee, the staff also concluded that an exemption from the requirements of 10 CFR 50, Appendix A, Criterion 57 is also warranted.

This criterion states the requirements for closed system isolation valves. III. Following the TMI accident thousands of curies of fission gases and active particulates were released to the containment atmosphere.

Because t* of the unique condition of the core and the amount of contamination resulting from the accident, the NRC imposed the requirement to maintain containment integrity to ensure that radionuclides inside the containment would not be releaSed to the environment.

In October 1979, the first of several containment penetrations were modified to probe the containment interior to evaluate the extent of damage and gather data for the cleanup. The penetrations were modified in accordance with NRC approved procedures.

The TMI-2 Proposed Technical SpeCifications also required that penetrations and operations that could affect containment integrity could be modified only by NRC approved procedures.

l \ l PAGE 2,( OF, 3S""" Since the 1979 accident, fis:sion gases that were released to contairrnent have .. -' .. either decayed or have purged from the contairment.

Decontamination activities have also reduced ambient airborne particulate contamination to levels below maximum penmissible concentrations listed in 10 CFR Part 20, Appendix S, Table 1. In an evaluation associated with a Modification of Order dated April 9, 1982, the staff concluded that the maximum credible contairment building pressure was approximately 2 psig. Calculated offsite doses resulting from a failed penetration in conjunction with a 2 psig driving head and the associated reactor building airborne contamination were well below the limits of 10 CFR 20 and within the scope of impacts assessed in the -Final Programmatic Envirormental t* Impact Statement" dated March 1981. Criterion 57 requires that each line penetrating the primary containment that is neither a part of the reactor coolant pressure boundary nor directly connected to the contairment atmosphere have at least one containment isolation valve which shall be either automatic, or locked closed or capable of renotemanual operation.

This valve shall be outside contairment and located as close to the contairment as is practical.

A simple check valve may not be used as an automatic isolation valve. Criterion 57 was written for operating plant conditions and is generally applicable whenever the plant is operating, in startup, hot standby, or during core alteration.

Presently, the conditions at Unit 2 most closely resemble the standard criteria for cold shutdown (K 'ff <.. 0.99, T 200°F). e ave During the nonnal 'cold shutdown mode for typical plants, containnent , . "'., . ",...,.. .... -integrity is nonnally not required and Criterion 57 is not nonnally applicable.

As previously stated, the staff correlated the shutdown condition of TMI-2 to that of a nonnal reactor in "cold shutdown." The staff also approved on this basis various penetration designs on the premise that if the plant were to enter a mode that when compared to a nonnal plant, would require containment isolation, either an alternate design 'or an exemption to penetration criteria would have to be approved by the NRC. The licensee proposed several alternate penetration designs to the NRC staff t" to support specific recovery operations.

The isolation feature common to all the alternate designs includes two isolation valves outside of containment.

In most cases, isolation valves are manual. Manual valves were found able for isolation in lieu of the Criterion 57 requirements since all conceivable accident scenarios still pennit access to the isolation valves. Therefore, it is the staff's opinion that penetration modifications of the type described above will be acceptable for all future recovery operations.

The staff has detennined that the post-aCCident status of the TMI-2 facility presents exceptional circumstances relative to the applicability of the Commission's regulations.

Because of the suspension of the licensee's authority to operate the facility in other than the present recovery mode as defined in the proposed technical specifications, certain of the lations. which are intended to apply to nonnal operating plants, are simply -. inappropriate and, mor,e significantly, are unnecessary to protect the public .. '., "". ....... heal th and safety *. Indeed, given the unique status of the plant in terms of primary system temperature and pressure, available fission product inventory, the ability to cool the reactor without forced circulation (loss-to-ambient), and the low decay heat rate, maintenance of the faciliy with the exemptions

• granted and the alternate design approved hereby will provide an equivalent level of safety. Furthennore, because of the condition of the plant and the need to proceed with cleanup activities, compliance with the regulations from which relief is sought would present an unwarranted impediment.

IV. l . Accordingly, the Commission has determined that, pursuant to 10 CFR 50.12, an exemption is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest.

The Commission hereby grants an exemption to the requirements of 10 CFR Part 50, Appendix A, Criterion

57. It is further determined that this exemption does not authorize a change in effluent types or total amounts nor an increase in power level and will not otherwise result in any significant environmental impact. In light of this determination and as reflected in the Environmental Assessment and Notice of Finding of No Significant Environmental Impact prepared /

t* l.' #-_---::.-.

__ .-------.

_ .. _--_. ---PAGE 32-OF sS pursuant to 10 CFR 51.21 and:51.30 through 51.32, issued concurrently herewith, ., .. "",:.: it was concluded instant action is insignificant from the standpoint of environmental impact and that an environmental impact statement need not be prepa red

• Effective Date: July 17, 1984 Da ted at Bethesda, t1a ryl and Issuance Date: July 17, 1984 FOR THE NUCLEAR REGULATORY COMMISSION Harold R. Denton, Director Office of Nuclear Reactor Regulation PAGE 33 Ji 3S .. "".J'.II . .

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9. 1N2. Tf\ls 1s h bal1l far the current dIIiW I presSIJl'e criteria in the Tecnn1ca1 5CJecJ.flc.t1ana.

llsed an the .oove, provicl1ng c:ICU)le 1aolaUon outslde cantaJ,nllent tor teIIPOfary to existing priMry contaiNent penetrations provides acIIQllte protection to the healtn Ind .. tety at the p.cllc uing the recovery effort If\Lle COIIPlylng wltt\ the Jntn of 10 (FR 50 A Criterion these C-WUaUant 1ft not dllipd in .ccarGlnCe w1tn 10 Cf'R SO A criteria 2, 50. or 51 to wlhtand the affects of a Slfe (SSE) al dlacuateCI 1n CPUC Letter 4t10-8l-l.-G18S dIIted August 26, 1983. Tnerefcn, QFUtC requeltl an elllllPtion to tte M1.a1c nqu1nllents far ..c21tled cmtav--.t peretrat1ans.

UNITED STATES SEP 4., NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20155 Mr. &ale K. Hovey Vice President and Di rector of TMI-2 Metropolitan Edison Company P.O. Box 480 Middletown.

Pennsylvania 17057

Dear Mr. Hovey:

PAGE SEP 2 1931 The Nuclear Regulatory Commission hereby grants Metropolitan Company, et. al. an exemption from certain requirements of 10 CFR Part 50. Appendix J 11;r lire Three Mile Island Nuclear Station Unit 2. This exemption consiSts of relief from the requirement to perform Type A. B *. and C leakage tests on the TMI-2 reactor building and is in response to your request of May 11. 1981. This exemption does not provide relief from the requireJRents to leak test the air lock door seals in accordance with Appendix J. subsection III.D.2.b.iii within three days after the door has been opened. See Surveillance _nt 4.6.1.3.2.

By performing the air lock door seal test. air lock integrity can be verified without the radiation hazards applicable to performing Type A. B, and C tes ts

• We have determined that the granting of this exemption involves an action which is insignificant from the standpoint of environmental iapact and that there is reasonable assurance that the health and safety of the public will not be endangered by this action. Having made this deteNination, we have further concluded that pursuant to 10 CFR 151.5 (d) (4) an environmental iapact appraisal need not be prepared in connection with the granting of this exemption.

Copies of the related Safety Evaluation and the Notice of Issuance.

which has been forwarded to the Office of the Federal Register for publication.

are also enclosed.

Sincerely. , ....... , I*... , . I . J .1 'Si;y"r PrOgfuJ TMI Program Offi ee-Office of Nuclear Reactor Regulation

Enclosures:

1. Safety Evaluation

,. ' 2. Noti ce of Issuance cc w'encls: See attached

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_ ... ,_s..c __ .M UI., z: OF /4 SAfETy EVAlUATION IN SUPPORT OF EXEMPTIONS FROM CERTAIN REQUIREMENTS OF THE CCIMISSION I S RULES AND REGULATIONS BY THE OFFICE OF NUCLEAR REACTOR REGULATION U. S. NUCLEAR REGULATORY CCM1ISSION IN THE MAmR OF METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER I LIGHT PENNSYLVANIA ELECTRIC COMPANY THREE MILE ISLAND NUCLEAR STATION, UNIT 2 DOCKET NO. 50-320 PAGE Of. (4

---_. ._j.-* .-.*.. _....,.-...;.-.-.:.

.. . ' l I. INTRODUCTION SAFm EVAlUATION IN SUPPORT OF AN EXEMPTION F101 CERTAIN REQUIREMENTS OF APPENDIX J TO 10 CFR PART SO Metropolitan Edison Company has requested (reference

1) ex_tion from certain requirements of 10 CFR, Part 50, Appendix J, which states the criteria to be used for verifying primary reactor containment leak tight integrity.

The licensee has proposed the exemption based on the reactor and the containment's current and future status, and the lIinimal consequences per Met-Ed's lations for any containment press_urization accident.

The TMI Program Office staff has reviewed the licensee's technical justification and concludes that the request for exemption from Appendix J is justified and acceptable.

Our \ -basis for this conclusion follows. II. EVALUATION Per 10 CFR Part SO J. paragraph III.D.l.(a), after the preoperational . leakage rate tests, a set of three type A tests are requi red at approximate equal intervals during each 10 year service period. This required testing .. asures primary-reactor containment overall integrated leakage under design basis accident pressure conditions.

The applicable test pressure is discussed in paragraph III.A.4 of Appendix J. For Type B tests, paragraph III.D.2 of 10 CFR 50, Appendix J requires that air locks be tested at 6 IIDnth intervals.

Penetrations are also required to be tested every other reactor shutdown for refue11ng but in no case at '--: intervals greater that 3 years. These tests will detect local leaks and .asure leakage across each pressure containing or leakage limiting

• . ::: PAGE S OF 14---2 -boundary for a reactor containment penetration.

All of these tests are performed by local pneumatic pressurization of the containBInt penetration either individually or in groups at a pressure not less than the calculated peak contail1l8nt intemal pressure related to the design basis accident.

This pressure at TMI-2 is 56.2 psig. Type C tests .. asure containment isolation valve leakage and have requirements set forth in paragraph 111.0.3 of 10 CFR 50. Appendix J. Type C tests shall be performed during each reactor shutdown for refueling but in no case at i ntenal s greater than 2 years. In addition to the Type At B. and C tests discussed.

paragraph IV.A of Appendix J requires that any .ajor modification or replacement of a component which is part of the prilllry reactor containment boundary or resealing of a seal welded door. perfonled after the preoperational leakage rate test shall be followed by either a Type At B t or C test as applicable for the area affected by the modification tests. In reviewing the applicability of Appendix J. an analysis was performed by the licensee (reference

1) to estilllte the .xinun containment building pressure change in the event that intemal equipment or piping failed. The TMIPO staff perfor-.d a similar analYSis and confirmed the licensee's results. The worst case equiPleftt failure analysis was based on the loss of all Reactor Building Air Coolers which are located inside the reactor building.

Priurily because of the low decay heat in the reactor coolant system (less than 32.2 kw) the effects of the loss 0'1: the coolers has been 1IIinimized.

The analysis concluded that the pressure inside of the containment building would take several days

.. l \. PAGE ro {4--3 -to increase by one to t., psi, ass .. 1ng this scenario occurred duri.ng the summer months which would be the worst case .. bient condition.

Another analysis based on the worst case piping failure assumed the instantaneous release of all reactor coolant to containment.

The pressure of the reactor + coolant system is .. intained at 90-10 psig and the temperature of the coolant ranges from approx1 .. tely 120°F in the hot leg to 7S o F in the cold leg. At . these temperatures and pressures, the effects on the containment atmosphere is minimized.

Therefore, the LOCA analysis resulted in approximately 2 psi pressure increase in the containment building.

The only that M)uld cause the pressure to exceed approximately 2 psi would be a recrit1cality accident.

This event .. s discussed in the Final Programmatic Environmental Impact Statement (PElS) for TMI-2 issued in March 1981. Paragraph 4.1 of the PElS states that -the most probable (although very unlikely) cause of recritical1ty

.. s found to be boron dilution, which M)uld be a slow enough process that any approach to criticality can be detected and remedied." This statement is still validi therefore, the staff has concluded that this accident need not be designed against in reference to containment integrity.

The containment is a prestressed reinforced concrete structure that provides biological shielding for normal and accident conditions.

It is also constructed to contain the pressures associated with a loss of coolant or steam generator blowdown accident occurring at 1001 power. Since the conta1maent has been analyzed for capability to withstand such accidents, the accidents discussed in this safety evaluation are within the l1lrlts of those for which TMI-2 .as originally designed and evaluated as discussed in the safety evaluation report for operation (NUREG-G107, Suppl .... ts 1 and 2). _0._" _______ . ____________

• .._ ... " __ .... __

.. . Consequently, the granting of this exemption would not result in a significant increase in the probability or consequences of accidents previously" considered nor a significant reduction in a _rg1n of safety, and does not involve a significant hazards consideration.

In addition to the discussed analyses results, Type A, B, and C tests would require a considerable amount of work and operator time spent in high radiation areas resulting in significant exposure to personnel, which. would not be sistent with the ALARA concept. There has been no detectable leakage of radioactive materials from containment since the ",reh 28, 1979 accident, however, a pressure test of the structure and its penetrations at the desi9n pressure of 60.0 pSig could induce a leak resulting in an uncontrolled release of radioactivity.

This pressure would increase the potential for a containment leak and therefore not benefit the public interest.

Based on the analyses, the*ALARA

'linplications, no apparent leakage from the contli ... nt and the increased risk associated with performing the tests, the TMI ProgrUl Office staff concludes that the public interest is served by not imposing the applicable requirements of Appendix J to 10 CFR Part 50 since such imposition would result in hardship . . or unusual difficulties without a COIIIPensating increase in the level of quality and safety. However, if a subsequent decision is _de to restore TMI-2 to operation.

all of the requirements of Appendix J shall again be applicable.

III. CONCLUSIONS Based on the foregoing, we have deterlrined that, pursuant to 10 CFR Section 50.12, an exemption to the periodic leak rate testing requirements of Appendix J to 10 CFR Part 50 is authorized by law and can be granted without life or DroDerty or the common defense and security and is --, ...

---

" ..

... --...... ... -'-.

t. \. * . otherwise in the public inprest. In _king this detel'lrinat1on we have given due consideration to <the burden that would result if these requi .... nts were i..,sed on the facility.

The granting of this re11ef does not involve a significant hazards consideration.

We have detenrfned that the granting of this eXeqltion does not authorize a change in effluent types or total UIOunts nor an increase in power level and will not result in any significant lien tal We have concluded that this exemption would be insignificant from the standpoint of environmental impact and pursuant to Paragraph (d) (4) of Section 51.5 of 10 CFR Part 51 that an enviror.ntal illPact statement, or negative declaration and environmental appraisal, need not be prepared in connection with this action.

\ -------= -,.. PAGE CJ Of. (4--REFERENCE

1. Letter to Lake Barrett, NRC, from G. K. Hovey, Metropolitan Edison Company, -Request for an Exemption from the Testing Requirements of 10 CFR 50, Appendix J,-LL2-81-OO94, May II, 1981. * --

--

t. .. PAGE 10 OF. I 4-UNITES STATES NUCLEAR REGULATORY COMMISSION DOCKET NO. 50-320 METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER AND LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY GRANTING OF RELIEF FROM APPENDIX J REQUIREMENTS OF 10 CFR PART 50 The U.S. Nuclear Regulator,y Commission (the Commission) has granted relief from certain requirements of Appendix J to 10 CFR Part SO, *Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors", to Metropolitan Edison Company, Jersey Central PC*er and Light CoqJany, and Pennsylvania Electric CoaIpany.

The relief relates to the leakage testing requirements for tests in areas which are radiologically inaccessible.

The request for relief complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's . . rules and regulations.

The Commission has made appropriate findings as required by the Act and the Commission1s rules and regulation in 10 CFR Chapter I, which are set forth in the NRC Staff Safety Evaluation Report in thi s matter dated The Commiss*ion has determined that the granting of this relief will result in any significant environmental impact and that pursuant to 10 CFR 151.5 (d) (4) and enviro .. ntal illPlct stateMnt or negative declaration and environ.ntal impact appraisal need not be prepared in connection with this action.

-# ** , . For further details with respect to this action. see (1) the request for relief (2) dated May 11. 1981. and (3) the ec.nission*s letter to the licensee dated September 2.1981. These items are available for public inspection at the Commission's Public Document Room. 1717 H Street. N.W ** Washington.

D.C. 20555 and at the Government Publications Section. State Library of Pennsylvania, Education Building.

Commonwealth and Walnut Streets. Harrisburg.

Pennsylvania

.17126. A copy of item (2) _y be obtained upon request addressed to the U.S. Nuclear Regulatory Commission.

Washington.

D.C. 20555, Attention:

Di rector. TMI Program Offi ce. Dated at Bethesda.

Maryland this September

2. 1981. oj. FOR THE NUCLEAR REGULATORY COMMISSION -:) J"" , . .:;0. I 0': f. Bernard J. Snyder. Program Director Three Mile Island Program Office Office of Nuclear Reactor Regulation 0** *._0 __ .0_ *. _-,-..... __ _...... _. _

-.. t: .. PAGE fZ-OF 14-MetropaU.n Edison Company Post Office Box 480 Middletown.

Pennsylvania 17057 717 844-4041 lI&y n, 1981 LL2-81-0094

'IMl-2 D1atr1ltut1oD TMI Proaram Office Attn: Mr. Lake Barrett, Deputy Director U.S ** uclear Reculatory eo...illion c/o Three Mile Illand .ucl.ar Station Middl.town, P.nnlylvania 17057

Dear Sir:

Three Mile I.land Nuclear Station, Unit 2 (THI-2) . Operation Licenle No. DPR-73 Docket No. S0-320 Reque.t for an 'Exemption from the T ** tina lequiremenu of 10CFR SO Appendix J Paragraph (0) of 10 CFR 50.54 "Conditions of Lic.n ...... tates that "primary reactor containment for vater cool.d power r.actor. Ihall be lubject to the r.quirementsaet forth in Appendix J." Appendix J lpecifie.

l.ak t ** t r.quirement.

for verifyiDJ the leak-tight inte,rity of the primary r.actor contaiament.

Iow.v.r, perfoTmiDJ Appendix J t ** t. at Three Mil. Ialand Nucl.ar Station Unit 2 (THI-2) i. no longer appropriate vith the r.actor and the contaiament in their curr.nt condition.

Therefore an exemption from the requirement.

of Appendix J i. reque.ted.

The ba.i. of thia requeat il the fact that potential contaiament .ode. would only re.ult in very .1iCht incre.ental poaitive pr ** aure change., commencin, from a ne,ative prea.ure .a required by para,raph 3.6.1.4 of the Recovery Technical cation.. For example, an analyd. va. performed to bound contaiament building prea.ur.*chan,e in the event of failure of all the aeactor luildiDJ Air Cooler., Which are located in.ide the contaiament buildina.

1hia concluded that the pre ** ure in.ide the containment buildin, would take leveral daya to increa.e by one to two ,.i, .a.wainc tbi. Icenario occured duriDJ tbe lummer 8Onth ** Thia analy.i. va. ba.ed on tbe followin, heat input.: .) Solar aadiation

• 1.82 X 10 6 STU/Hr. b) Core Decay Heat

• 0.327 X 10 6 Btu/Hr. (Thi. value hal -aince decayed to approxiaately 0.11 X 10 6 BTU/Hr.a.

of May I, 1981.) .. ... :-... ), at. Id-AD.IC.

lartoa-AD.IC.

Clark-PAIl.

IleV1!ae-AD .IC. ElD-AD.Be

• FUti-Tl.259 , Fuller-AD.IC.

Bard11l1-Tl.68 lierbe1n-Tl.118 Heward-PAIl.

Hockley-HEAR

• Rolzvortb-EG&G Bovey-AD.IC

• HWt111-Tl.l84 ltu&uas-1I A1l. Uq-AD.Be

• luDder-AD.1C

• Lace,..JCP&L Ikftpuro-PAIl.

sa.aua.-PAIl.

!horpr'AR.

'Hptoa-PAIl.

Wallace-PAIl

• Wa18b-PAIl.

J VUllOD-JU03 a VUaoa-PAIl.

DDCC-'IMl DDCC-PAIl.

C A / .'

l.* Jk. Lake Barrett PAGE 13 . Of (4-c) CoDduct\on aDd coaveciion throuah CODcrete vallI (baled OD 90 0 F and 8001 outside aod iaaide air t_perature respectively)

• 0.142 X 10 6 BTU/Br. Another analYli. Which va. perfor.ed a ** u.ed in.tantaaeou.

release of all reactor coolant to the cODtaia.ent.

In this Icenario:

a) 'lbe averaae (bulk) lncon t_perature va ... su.ed to be about 12001, b) 'lbe reactor coolant Iy.tem fluid va. assu.ed to be instantaneoully and homeeneoully released, c) The heat contained in the fluid val .ssumed to homoaeneoully diltributed to all two million cubic feet of air in the contaiament d) 50 credit for heat tranlfer to any coaponentl or the concrete structure vas a .. The analYlil yielded an approximate two (2) Pli prellure increale in the reactor contaiament building.

The sole event Which could caule contaiament prellure to exceed thele low pre. lures i. a recriticality accident.

The probability of this accident val evaluated in the Programmatic Environmental Impact Statement Which .tated that "(dhe molt probable (although very unlikely) cau.e of a recriticality val found to be boron dilution, Which would be a .low enouah proce.s that any \.. . approach to criticality can be detected and remedied." Therefore ba.ed on the above, we believe there is adequate ju.tification for Iranting an exemption from Appendix J te.ting. MditionaUy, performance of this teltina hal ALARA implication.

in that performance of the required te.ting would involve a con.iderable amount of work in high radiation areas. Bence it would re.ult in lianificant radiation expo.ure to te.t perlonnel.

Specific reasonl for exemption from each type of Appendix J te.tina are dilcussed beleN. Type A te.ting i. intended to .ea.ure primary reactor contaiament overall intearated leakage under delian basil accident conditions.

HeNever, al discussed above, becau.e of the current condition.

at TMl-2 the contaiament i ** ubject to very low politive pre.lures in the event of an accident.

further, performance of Type A te.tina require. exten.ive preparation in. ide the contaiament prior to pre ** urizing the contaiament.

With .. ny of the area. of contaiament phy.ically and/or radiologically inacce.sible the.e preparation.

c.nnot be completed.

Additionally IRC approved .odification.

have been .. de to contaiament tion. R401, 1S61 and 2626, to alleN their ule during the recovery, Which have reduced their .e.iln pre ** ure. 'or the.e rea.ons it i. currently neither nece ** ary, DOl' po ** ible, to perform a Type A test.

.. _--. -.. --.. _-..

l.: 1Ir. L.ke Barrett 'lJpe

• t ** dlll b illteacJ.d to d.t.ct loc.l leak. a1lc! to *** ure l.akqe .croll eacb pr *** ure contaillilll or leakaae li.itiaa bouDcJar, for r.actor cont.iament pelletr.tion.

lucb a. tb. cont.i_nt .ir lock. acJ relUient .. ala. !'b.re pre.entl,i.

1lO purpo.e ill te.tilll th ** e penetr.tioll.

to 56.2 plil (,.) ** r.quired b, Appendix J "ec.u.e. a. dbcu ** ecJ pr."ioud,.

the COIlt.i_llt ia lubject to only very low poaidve pr *** ur ** ill the .. ellt of _ ,accident.

Addition.lly, .lthoulb the penetr.tion pre ** uri.ation connectioll.

for all cont.iament penetr.tion.

are acc *** ible.

h.lf of the penetr.tions' are loc.ted ill bilh radi.tion .r *** out.ide the collt.i ... nt. !berefore even if

• l.ak i. detected froa .ny of tbe.e penetr.tioll.

there i ** hilb prob.bility tb.t the .ource of tbe le.ule could Ilot be found acJ rep. ired

• Further. the fieeS penetr.tionl fora. double b.rrier vith a de.ian pre ** ure of 60 p.ia which i * .ore th.n adequate for the very low po.itive pre ** ur ** to vbicb the .ent .. y be .ubject. Therefore t,pe

• t ** tilll Ile.d Ilot be performed.

'lJpe C t ** t ** re illtended to .aaure cont.i_nt i.ol.tion v.lve le.k.ae r.tes. In tbi. telt tbe cont.iament i.ol.tion v.lve ** re te.ted for leakaae aa.inst

• te.t pre.sure (P.). of 56.2 p.il i. aucb are.ter th.n the very low pre.lures the cont.iament i ** ubject to in tbe ."ent of .n .ccident.

Furtber tbe Recovery Tecbnic.l .pecific.tion.

require u. to .. int.in tvo OPERABLE cont.iament tion v.lvel cloled vben not requir.d open in accord.nce vitb an .pproved procedure.

Thie foral

• double b.rrier vith

• deaian Fea.ure of 60 psil (for tbele ilol.tion v.lves) whicb il .isnific.ntl, are.ter tb.n tbe 8aXiaum pOlsible cont.iament prel.ure ** di.culled e.rlier. Therefore type C te.tinl need not be performed.

Addition.lly 64 of 67 .. cbanic.l penetr.tion.

require oper.tions to be performed in biab r.di.tion .re.1 botb inlide and out. ide of cont.iament Which vould relult ill .ianificant perloDDel expo.ure ** In conclu.ion:

'lJpe A te.tina i. Ileitber n.c **** ry. Ilor possible.

under current condition.

and Type ** C testina would .erve 110 u.eful purpo.e .s tbe cont.iament is .t80st .ubject to very low politive pressures Whicb are aucb lower tb.n the desian pressure of tbe lubject items. Addition.lly,there is

• very hilh ability tb.t any leak tb.t i. di.covered viiI be ill

• hilh radi.tion are. which i. Ilot .cces.ible and di **** embly of coaponent.

requir.d to fix tbe le.k could r.sult in an _ ** fe condition.

!berefore we reque.t th.t an Appendix J exemption be ar.nted. In tbe event

• deci.ion i ... de to re.tore THl-2 to .n oper.ble condition we underst.nd tb.t we viII then be required to coaply vitb the _nt. set forth in Appendix J. Sincerely, G. It. Bovey Vice 'resident Director.

tMl-2 CItH:JJB:be cc: Dr. B.J. Snyder, Program Director -n!l Office

, -

GPU Hue *** r Corpor.tion Post Office Box 480 Aoute 441 South Middletown, Pennsylvania 17057-01 e1 717 944-7621 I* ( 'lMI Program Office Attn: Dr. B. J. Snyder Program Director US Nuclear Regulatory Camnissioo Wasmngtoo, DC 20555

Dear Dr. Snyder:

Three Mile Island Nuclear Statioo, thit 2 ('lMI-2) Operating License No. DPR-73 Docket No. 50-320 Fire Protectioo for Safe Shutdown TELEX 84-2386 Writer's Direct 0181 Number: (717) 948-8461 44l0-84-Ir-0124 Dxl.Dent 10 0047A July 31, 1984 In a letter dated May 18, 1984, NRC granted GPO Nuclear a schedular exemptioo to 10 CPR 50, Appendix R. iJaplementatim requirements.

As an alternative GPU Nuclear was required to provide a S\ll!lllBry of present or prO,tX)Sed fire protectioo features for systems required to mintain or Iali tor cold shutdown and provide a Technical Speci ficatioo Olange Request (TSCR) to incorporate those fire protectioo features into the Technical Specificatims.

This letter is the respc:nse to the NRC request. As an initial step to respald to NRC' s request, GPO Nuclear utilized a recently initiated study to identify those systems necessary to mintain and lID'li tor the cold shutdown CXI'ldi tim. 'lhe study determined those functialS/systems are necessary to _intain and verify the Safe ShutcXJw.n (SSD) c:Xmditic;n of the 'lMI-2 core. 'lhe study then focused m . protectioo of the cold ahutciJwn state fran the effects of a Desi91 Basis Fire (Im') ,and was performed without regard to existing systan requirements, i.e., Technical Specificatioo requirements and previous analyses.

'lhe intent was to exclude, from the study, the broader based (Technical Specifioatioo) requir.ents which had DO effect m the _fe shutCkJwn requirements.

After identifying the systems necessary to "aintain and lD1itor" the SSD CXI'lditioo, a _fe shutciJwn analysis was performed to determine if a lBF in any fire area/Zale could prohibit aint.-mnce or .alitoring of the SSD CXI'lditim.

On the basis of the SSD analysis, GPO Nuclear CXI'lcludes that GPU Nuclear Corporation is a subsidiary of the General Public Utihties Corporatton

{ ( Dr. B. J. Snyder July 31, 1984 44l0-84-L-0124 the level of fire prdtectioo presently provided in the plant and specified in the Technical SpecificatialS is sufficient to protect the SSD CCI'lditioo.

Therefore.

a 'l'SCR is not being suanitted with this corresp::ndence.

Attadhment 1 provides a sUll11'llU'y of fire protectioo features in the fire areas identified as necessary for maintenance and lIOlitoring of the SSD CCIlditioo.

Attachment 2 provides a s\llllJlliU'y of the system identificatioo study and the SSD analysis.

This letter is being submitted after the 60 day period specified in your letter dated May 18, 1984, as cxx>rdinated between Mr. T. Poindexter of your staff and Mr. J. J. Byrne of my staff 00 July 12, 1984. If you have any questiCXlS CCIlceming this informatioo, please call Mr. J. J. Byrne of my staff. BYJ<./SOC/jep Attachments Sincerely, /s/ J. J. Barton for B. It. l<anga Director, 'lMI-2 cc: Acting Deputy Program Director -'lMI Program Office. Mr. P. J. Grant PAGE 3 OF b ( Attachment 1 FIRE DETECTION/SUPPRESSION AVAILABILITY FOR SAFE SHUTDOWN .!!:!!. -Detection SUppress ion CoIIaents Fuel Handling Building Area SIoke Detection Hose reel/portable Additional hose reel available (FA-OOn extinguisher in the Auxiliary Building Auxiliary Building Area SIoke De.tection Hose reels/portable HVAC filters in Fire Area (FA-D09) extinguishers lIose deluges Control Building Area SIoke Detection Hose reels/portable Fire Area is required for (FA-033) extinguishers rtIOte operations only Cable Chase Area SIoke Detection

• Hose reels/*portable Fire Area is required for (FA-035) extinguishers r..ote operations only HIV Duct and Cable Room Area s.oke Detection

• Hose reels/*portable Fire Area is required for (FA-041) extinguishers reaote operations only Cable Room (FA-045) Area SIoke Detection Halon/*hose reels/ Fire Area is required for *portable extinguishers rtIOte operations only Control Room (FA-046) Area SIoke Detection Portable extinguishers Fire Area 1s required for rtIOte operations only Service Building and Area Smoke Detection Hose reels/portable Fire Area is required for Control Building Area extinguishers remote operations only (FA-047) Reactor Building (FA-049) Area Smoke Detection Hose reels/portable

• • 281' Elevation exU ngui shers inaccessible

• This equipment located/available outside of Fire Area. ** The elevation is inaccessible due to high radiation levels. This area does not contain any SSD support' equiPMent.

The combustible loading level is low and sources of ignition are .ini .. l. NOTE 1: All Area Saoke Detection, Hose Reels, Ind Halon$ystels noted Ire TMI-2 Technical Specification 1te.s with the exception of the three (3) hose reels in the Service Building (part of Fire Area 047) which are controlled and serviced per the NatiOQll Fire Codes under approved procedures.

NOTE 2: This lfsting does not include Fire Areas whose functions ltave not changed due to head lift conditions.

All Fire Areas not included on this list contain detection and suppression equipment IS listed in the TMI-2 Fire Protection Progr .. Evaluation, Revision 1. ---._--<--_.-

-.. ....,,--____

--__ . _ ... __

( PAGE' 4 7' Attachment 2 A study was performed to identify trose systems required to mintain and JIO'li tor the shutCbwn .9Qldi tim of the plant and to perform an analysis of the impact of a Desi91 BaSis Fire (IEF) m the IIystems identified.

,;'he c:bjective was to ensure the capability to mintain and JIO'litor the reactor in a cold shutCbwn state is not lost as a result of a mF. 1.0 S'1'UDY APPRlAQi '!'he approach used for the study was as follows: Define Safe Shutcbm (SSD) relative to 'JMI-2's OCI'lfiguratim Identify cc:ntrolling mctioos necessary to mintain the SSD cc:nditioo Identify lII:X'litoring requirements to verify (ensure) the SSD cc:nditioo that portioo of the plant (i.e., plant systems, subsystems, etc.) that would be necessary to maintain/m::ni tor those functioos as identified above. Perform a SSD analysis to determine the impact of a reF. 2.0 Safe ShutCbwn '!be reactor at Three Mile Island Unit 2 is stable with the head renoved, core heat rejectim via loss to aDbient, the core is subcritical (i.e" keff less than 1.0), and reactivity is being cc:ntrolled through the cc:ncentratioo of dissolved poison (boroo) in the Reactor Coolant System (RCS) inventory.

For the p.1I'pOSe of this evaluatioo, maintenance of SSD can be qualitatively equated to maintenance of the core in a subcritical cc:ndi tim. 3.0 Controlling Parameters Maintenance of the core in a suberi tical cc:ndi tioo is entirely dependent m cc:ntrol of ,the core reactivity.

O::>re reactivity, for 'JMI-2. is a mctioo of core geanetry or OCI'lfisuratioo and the (presence and) OCI'loentratim of dissolved poi&alS (berm) in the RCS water. If a IEF is to bave an impact m the core, it must l)alter the core's cc:nfiguratioo, 2) affect RCS water inventory, or 3) affect RCS berm OCI'loentratim. Doc\Dent 10 0047A -----_.--.---

---_. -.--

4.0 Monitoring Reqqirements The JIDlitoring necessary to verify the oontrolling mctiCX'lS CCI'lSists of 1) cbtaining and analyzing water aamples for boroo ooncentratioo and 2) maintenance of water level indicatioos.

5.0 Analysis and Results 5.1 Configuratioo Changes in reactivity due to a core oonfiguratioo Change resulting fran the direct effect of a I8F are not CCI'lSidered credible.

A IeF, by itself, oould not produce a mechanical s'OOck sufficient to cause settling or shi fting of the core. The allyoonceivable indirect ..chanism by which a RB I8F would affect core oonfiguratioo is via a IeF induced polar crane load drop. This mechanism is also not oonsidered credible because the polar crane Cbes not have a single failure catIfClIlent subject to a DBF that can cause a load drop, and a. DBF occurring in the RB is not assumed to spring into being as a full-blown.

fire. Sufficient time can be assumed to allow J1Dvement of a load suspended over the reactor vessel to a safe area. 5.2 Water and Level Monitoring Q::nceivably, a DBF can adversely affect _intenance and mcnitoring of the RCS water inventory, i.e., water level, through three (3) di fferent mechanisms:

loss of an RCS or associated fluid boundary, loss of mkeup capability to the RCS, and by disrupting the ability to mc:nitor RCS water level. (teF induced additioo of inventory is discussecs under RCS boroo ooncentratioo.)

The systems which are available, at least in part, to maintain, makeup, and JD:X'litor RCS water level include: -Reactor Coolant System -Decay Heat Removal System (DHR) -Mini-Decay Heat Rerroval System (miR) -Makeup and Purificatioo System (MU&P) -StancJ::>y Pressure Control System (SPC) -Borated Water Storage Tank (!liST) and intercamecting piping with ReS -Reactor Vessel (RV) Standpipe -RC-LT-100 The necessary portialS of the above systems were examined in light of each mechanism by which a teF oould induce failure of the given aystem/C'CIDp'J'\ent such that it oould not perform its intended functioo.

A S1.DDla.rY of each follows. Document 10 0047A 5.2.1 Loss of RCS Piping or Vessel Boundary Integrity

'.rhe reactor vessel bo\;rldary, including the inoore guide tl.:bes, is not subject to mechanical failure due to a DBF ain£e no part of the bo\;rldary below the oold leg is subject to c::3atbustioo or melting \meter ISF conditioos.

This is generally true for the entire RCS fluid boundary.

except the IN Standpipe. " '.rhe IN Standpipe is subject to ocmbustioo and is assumed to lose its integrity.

This loss of integrity cannot result in lowering the RCS water level any lower than the oold leg ncEzle of the reactor vessel, i. e., the lowest tie-in p::>int to the vessel, excluding the incore guide tubes. The oold leg elevatioo provides an acceptable minimum water level above the oore to maintain the SSD conditioo.

This was evaluated in the Head Renoval Safety Evaluatioo Report which was approved by the NRC 00 July 17, 1984. 5.2.2 Loss of an RCS Support System Integrity The supp::>rt system bo\;rldaries (pipes, valves, and pump casings) are not subject to mechanical failure due to a JEF. H::lwever, if a loss of integrity did occur, it would not result in the IN water level g;:>ing below the oold leg rx::rz.zles.

5.2.3 Loss of Makeup Capability

'.rhe sources of water for the makeup of RCS inventory are tanks (SPC, RC:BT, and BtlST) that would be unaffected by a JEF. The path(s) of water makeup cxnsist of piping and valves that also would be unaffected except that rem:>te operatioo of valves is assumed to be lost. The pumps associated with water inventory makeup would not be operable, hcwever, they would maintain the system boundary.

Because of elevatioo differences, the SPC System and the BtlST could add water to the RCS witlx:>ut the use of electrical power. Therefore, a teF would not preclude manual operatioo of these two systems. 5.2.4-Loss of Level Malitoring Capability L8vel indicatioo is provided by two different systems. RC-LT-1OO provides RCS water level indieatioo to the 347' Elevatioo of the Fuel Handling Building (FHB) and to the Centrol 1tx1m. '.rhe level transmitter is located 00 the 280' Elevatioo of the FHB and taps into the Decay Heat System. The 'iN Standpipe is a visual water level reference which is contained entirely within the Reactor Building (Ra). Document 10 0047A

• . A t8F in either the !'HB or the RB would not cause a loss of all level indicatioo capability.

In s\DIIBIy, a t8F would not eliminate the ability to maintain or Jalitor . the RCS .water 5.3 RCS Boroo Concentratioo A t8F oould adversely impact the maintenance or Jalitoring of RCS boroo ocncentratioo by either the introduction to the RCS of water that has low or no boroo concentration or by the disruptioo of the capability to Jalitor boroo concentratioo.

5.3.1 Reductioo of Boroo Q:ncentratioo l..oi ccncentratioo and/or \rix)rated water oould be delivered to the RCS through nearly any of the support systems which ClCIDII1Il1'licate with the RCS. As a result, controls are implemented to isolate unnecessary CXIII1DUI'licatioo paths with the RCS from Ul'lCCXltrolled sources of water as well as requiring verification of that isolatioo on a regular basis. As a further precaution, tanks that are used for normal RCS makeup are sampled prior to transfer of their contents to the RCS. '1'hese controls are discussed in both the Head Renoval and Internals Indexing Fixture Processing Safety Evaluation Reports which have been approved by the NRC 00 July 17, 1984, and July 24, 1984, respectively.

A t8F would not establish a camnunication pathway between the RCS and an uncontrolled source of water. An additional mechanism by which unborated water could concei vably be introduced into the RCS is via use of fire suppression water from fire hose reels for a Reactor Building DBF (i.e., from fire fighting within the fuel transfer canal). However, the area around the open reactor vessel has minimal OOIIi:>ustible JlBterial loading and is, therefore, not subject to a fire of a magnitude to cause large quantities of fire suppression water to be used near the vessel. In addition, the Intema.ls Indexing Fixture and its work platform, :installed over the vessel, will keep 8)St, if not all, fire suppressien water out of the vessel *.in the event of a fire in the fuel transfer canal. 5.3.2 Illss of Boren Oxlcentratial Malitoring Capability Malitoring RCS boral concentratien is a functial of two acti vi ties: cbtaining an RCS auple and analyzing the sample. A t8F could p::>tentially interfere with the perfoI'JDi!!Ulce of either functial.

Document 10 0047A

. .. r '!'he sampling mechanisms ocnsist of the normal RCS sampling path and a DBnual grab sample directly fran the iN. RCS samples are normlly obtained through a Mtnpling statioo 00 the lOS' Elevaticn of the FHB which COIIIllI.lnicates through existing and temporary piping to the iN. 'l'he DBnual grab umple is obtained entirely within the Reaetor Building (through the IIF work platform) and would not depend 00 the FHB for any sU.P,lX)rt.

A IBF in the Reaetor Building will result in loss of the normal. aampling path. It my also impede DBnual grab samples. Scwever, mder emergency CXIlditicns, the RCS would be isolated thereby ensuring maintenance of the SSD CXIldi ticn. '!'his would include the securing of sampling, processing and/or any other processes in the affected fire area. At the end of the IBF aampling can be resumed either by restoring the normal sampling path or by initiaticn of a grab sample. Since sampling is ally required cnce per seven (7) days by the Technical Specificaticns and since the RCS would be effectively isolated during a aw, the RCS can be allowed to 9=> lZlSalIIpled during a IBF if As a result, a DBF in either building cbes not unacceptably impact the capability to obtain an RCS sample. Analysis of the sample is carried out in the Clemistry Laboratory that is outside the plant. Backup analysis capability is available in Unit 1. A IBF in any area of the plant would not result in a loss of analysis capability.

In SumJ[Bry, a DBF in any fire area would not adversely impact the ability to mintain or llICl'litor the RCS boroo CXIlcentraticn.

5.4 Cbnclusioos As shown above. a canplete loss of a fire area/zooe and its associated equipnent Cbes not result in loss of either the ability to maintain or llICl'litor the SSD CXIlditicns.

Given that, GPU Nuclear CXIlcludes that the present level of fire protecticn features existing in the facility and those specified in the Technical Specificatioos are sufficient to protect the SSD CXIlditioo of 'lMl-2. Document ID 0047A

, , UNITED STATES NUCLEAR REGULATORY COMMISSION Docket No. 50-320 Mr. B. K. kanga, Director Three Mile Island Unit 2 GPU Nuclear Corporation P.O. Box 480 Middletown, PA 17057

Dear Mr. Kanga:

WASHINGTON, D. C. 20555 May 18, 1984

Subject:

Three Mile Island Nuclear Station, Unit 2 Operating License No. DPR-73 ' Docket No. 50-320 Appendix R Exemption Request /)'-/0'<7 PAGE 9 The Fire Protection Rule, (10 CFR 50.48) published on November 19. 1980, became effective on February 17, 1981, and required the results of certain tasks to be submitted to the Nuclear Regulatory Commission (NRC) by Of

19. 1981. By letter dated March 24, 1981. you applied for exemption from some of the schedular requirements of 10 CFR 50.48(c) for the following items: (1) 10 CFR Part 50 Appendix R, Section III.G. --Fire Protection of Safe Shutdown Capabil ity" (2) 10 CFR Part 50 Appendix R, Section 111.0. -"Oil Collection System for Reactor Coolant Pump" You also requested relief from the schedular requirements contained in Section 9.0 of the TMI-2 Safety Evaluation Report, Supplement No.2, dated February 1978, for "Fi re Hose Stations Systems" and "Automatic Water sion in Diesel Room Basement." The staff responded to you in a letter dated May 7, 1981. which required that an updated Fire Hazards Analysis (FHA) be completed and submitted to the NRC, before a determination would be made on the exemption request. The revision to the FHA was made on June 15, 1982. With respect to items related to safe shutdown capability, the staff agrees with the licensee that the TMI-2 reactor is in a cold shutdown condition with no active systems, required for core cooling. However, certain mentation is required for monitoring various parameters such as reactor

. , Mr. B. K. Kanga coolant temperature and neutron flux level to insure that a cold shutdown condition is maintained.

Additionally.

several backup systems are required which can provide makeup and maintain pressurization for the reactor coolant system if necessary.

It is the staff's opinion that even though Appendix R requirements are not appropriate for the unique conditions at TMI-2. the Proposed Technical Specifications and the Recovery Operations Plan would be acceptable as an alternative location for specific fire protection ments for systems used to maintain and verify that cold shutdown.

Therefore it is our position that systems used for monitoring or maintaining the reactor in a stable cold shutdown condition (e.g., monitoring instrumentation, the Mini-Decay Heat Removal System and the Standby Pressure Control System) should have fire protection features.

Even though an exemption to the schedular requirements of 10 CFR 50.48(c) is granted relative to Section III.G of Appendix R, you should submit to the NRC a summary of present and proposed fire protection features for systems required to maintain or monitor cold shutdown.

You should also submit a change to your Technical Specifications to include the proposed features within 60 days of the date of the exemption.

With regard to the 011 Collection System for reactor coolant pumps, the staff finds that an exemption to the schedular requirements of 10 CFR 50.48(c) is warranted because of the shutdown condition of TMI-2 and the prohibition to operate the pumps per the technical specifications.

In summary, the Commission has granted your exemption request that the date for submittal of documents relative to 10 CFR 50.48(c) and 10 CFR 50, Appendix R. Sections III.G and 111.0 be extended for the remainder of the recovery mode as described 1n the enclosed exemption (Enclosure 1). A copy of this exemption is being filed with the Office of the Federal Register for publ ication. Enclosure 2 provides a rewording of the "request for information" included with Generic letter 81-12. This rewording is the result of meetings with representative licensees who felt that clarification of the request would help expedite responses.

It does not include any new requests and, therefore, will not adversely affect licensees' ability to respond to Generic letter 81-12. Enclosure 3 provides information regarding our criteria for evaluating exemption requests from the requirements of Section III.G.2 of Appendix R. In a letter dated March 13, 1984, the staff provided comments on your June 15, 1982 Fire Hazards Analysis *

'-. .' I Mr. B. K. Klnga . -The staff and its consultant have also reviewed the Fire HAzards Analysis and conclude that because of the shutdown condition of the TMI-2 reactor, the commitment made in the SER, Supplement No.2, dated February'1978, need not be met as long as the facility is maintained in its current mode.

Enclosures:

1. Exempt ion 2. Rewording of Request for Information

3. Criteria for Evaluating Requests 4. Notice of Issuance cc: J. Barton J. Byrne J. Larson Service Distribution List (see attached)

Sincerely,

10. J Bernard J.

Three Mile Island Program Office Office of Nuclear Reactor Regulation

( ::. ,. Or. Thomas Murley Regional Administrator.

Region 1 U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, PA 19406 John F. Wolfe, ESQ., Chairllan, Administrative .Judge 3409 Shepherd St. Chevy Chase, MD. Z0015 Or. Oscar H. Paris Adl\inistrative Judge Atollic Safety and licensing Board Panel U.S. Nuclear Regulatory Ca-llission waShington, D.C. Z0555 Or. Frederick H. Shon Administrative Judge AtOllic Safety and licensing Board Danel U.S. Nuclear Regulatory Commission waShington, D.C. 20555 Kari n W. Carter Assistant Attorney General 505 Executive House P.O. Box 2357 KerriSburg, PA 17120 Or. Judith H. Johnsrud Environmental Coalition on Nud ea r Power 433 Orlando Ave. State College, PA 16BOl Geor;e F. Trowbridge, ESQ. Shaw. Pittman. Potts and Trowbridge lBOO M. St., Washington, D.C. 20036 Atomic Safety and licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Atomic Safety and liCensing Appeal Panel U.S. Nuclear Regulatory Commission Washington, D.C. Z0555 Secretary

'J.S.

ATi'Il: C"ief, :Jocketing Service Sranch 20555 Mr. Larry Hochendoner Dauphin County Commissioner P.O. Sox 1295 Harrisbur;, PA 17108*1295 John E.

hairperso n , DaUPhin :ounty SOard of Commissioners Dauphin County .ront and Streets 17101

ounty Office Emergency Preo.redness Court 7 Front &

Streets HarriSburg, PA 17101 U.S. :nvironmenta1 Agenc:t Re.ion III EZS

Jrtis Dui1ding 6th &

Streets 7homas M. ,frettor DurtlU

Jf iI.O. Do" wess

Jf Er.*/ir:l"II!Itntai

'1!nnin; of

?.sources

'.:l. 30x 2::63 Willis Bixby, Site Kenager U.S. Department of Energy P.O. Box BB Middletown, PA 17057-0311 David J. Mc:Goff Division of Three Mile Island Programs NE-23 U. S. O.partlllent of En.rgy Washington, D.C. ZOS45 Williall lochst.t 104 Davey Laboratory Pennsylvania State University University Park, PA 16802 Randy Myers, Editorial The Patriot 812 Market St. Kerrisburg, PA 17105 Robert 8. Borsum Babcock & Wilcox Nuclear Power Generation DiviSion Suite ZZO 7910 Woodmount Ave. Bethesda, MO. Z0814 Michael Churchhi11, ESQ. PIlCOP 1315 walnut St., Suite 1632 Philadelphia, PA 19107 linda W. Little 5000 Herllitage 27612 Marvin 1. lewis 6504 Bradford Terrace Philadelphia,?A 19149 Jane lee 183 Valley Rd. Etters,PA 17319 J.B. Liberllln, Esquire Berlack,Israels, liberllan Z6 tlroadway New York, NY 10004 Walter W. COhen, ConSJmer of Justice Strawberry

Square, HarriSburg, PA 17127 Edward O. Swartz Board of Supervisors londonderry Township RFO Geyers Church Middletown, PA 17057 Robert l. Knupp, :SQuire Assistant Solicitor KnuDP Ind p.O. Sox D 407 N. Front St. HlrriSbur., DA John levin, Esquire Pennsylvanil C;)mm. P.O. Box 3265 Harrisburg, PA 17120 Honorable Cohen

aoital "120 Mr.

'fntner

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'ars; ::a t 'l.j ')::sj UNITED STATES NUCLEAR REGULATORY COMMISSION In the Matter of GENERAL PUBUC UTILITIES NUCLEAR CORPORATION (Three Mile Island Nuclear Station, Unit 2) ! ! ) EXEMPTION I. Docket No. 50-320 GPU Nuclear Corporation, Metropolitan Edison Company, Jersey Central Power and Light Company and Pennsylvania Electric (collectively, the licensee) are the holders of Facility Operating License No. DPR-73, which had authorized operation of the Three Mile Island Nuclear Station, Unit 2 (TMI-2) at power levels up to 2772 megawatts thermal. The facility, which is located in Londonderry Township, Dauphin County, Pennsylvania, is a pressurized water reactor previously used for the commercial generation of electricity.

By Order for Modification of License, dated July 20, 1979, the Licensee's authority to operate the facility was suspended and the Licensee's authority was limited to maintenance of the facility in the present shutdown cooling mode (44 Fed. Reg. 45271). By further Order of the Director, Office of Nuclear Reactor Regulation, dated February 11, 1980, a new set of formal license requirements was imposed to reflect the post-accident condition of the facility and to assure the continued maintenance of the current safe, stable, long-term cooling condition of 'the facility (45 Fed. Reg. 11292). This license provides, among other things, that it is subject to all rules, regulations and Orders of the Commission now or hereafter in effect. II. On November 19, 1980, the Commission published a revised Section 10 CFR 50.48 and a new Appendix R to 10 CFR 50 regarding fire protection features of nuclear power plants (45 FR 76602). The revised Section 50.48 and Appendix R became effective on February 17, 1981. Section 50.48(c) lishes the schedules for satisfying the provisions of Appendix R.Section III of Appendix R contains 15 subsections, lettered.A through 0, each of which specifies requirements for a particular aspect of the fire protection features at a nuclear power plant. Two of these 15 subsections, III.G and 111.0, are the subject of this Exemption.

Subsection III.G specifies detailed ments for fire protection of the equipment used for safe shutdown by means of separation and barriers (III.G.2).

If the requirements for separation and barriers cannot be met in an area, alternative safe shutdown capability, independent of that area and equipment in that area, is required (III.G.3).

Subsection 111.0 requires that the reactor coolant pump be equipped with an oil collection system if the containment is not inerted during normal operation.

The system has to be capable of collecting lube oil from all potential pressurized and'unpressurized leakage sites in the reactor coolant pump lube oil systems.

PAGE /s OF. G r Section 50.48(c) requires completion of all modifications to meet the provisions of Appendix R within a specified time from the effective date of this fire protection rule, February 17, 1981, except for modifications to provide alternative safe shutdown capability.

These latter modifications (111.G.3) require NRC review and approval and Section 50.48(c) requires their completion within a certain time after NRC approval.

The date for submittal of design descriptions of any modifications to provide alternative safe shutdown capability is specified as March 19, 1981. By letter dated March 24, 1981, the licensee requested exemptions from 10 CFR 50.48(c) with respect to the requirements of Section 111.G and 111.0 of Appendix R as follows: (1) Extend until the end of the Recovery Mode the date for filing additional exemptions or complying with the provisions of Section III.G as required by 10 CFR 50.48(c).

(2) Extend until the end of the Recovery Mode, the date for filing additional exemptions or complying with the provisions of Section 111.0 as required by 10 CFR With regard to the exemption requests, when this fire protection rule was approved by the Commission, it was understood that the time required for each licensee to reexamine those previously-approved configurations at its plant to determine whether they meet the requirements of Section III.G of Appendix R to 10 CFR 50 was not well known and would vary depending upon the degree of conformance.

For each item of nonconformance that was found, a fire hazards analysis had to be perionned to detennine

"'ether the ing configuration provided sufficient fire protection.

If it did not, fications to either meet the requirements of Appendix R or to provide some other acceptable configuration that could be justified had to be designed.

Where fire protection features alone could not ensure protection of safe shutdown capability, alternative safe shutdown capability had to be designed as required by Section III.G.3 of Appendix R. Qepend1ng upon the extensiveness and number of the areas involved, the time required for this reexamination, reanalysis and redesign could vary from a few months to a year or more. The Commission decided, however, to require one, short-tenn date for all licensees in the interest of ensuring a best-effort, expedited completion of conpliance with the fire protection rule, recognizing that there would be a number of licensees who could not meet these time restraints but who could then request appropriate relief through the exemption process. Because of the unique condition of the TMI-2 reactor, additional infonnation had to be obtained by the staff before a decision could be made regarding the applicability of Appendix R. This infonnation was requested in a letter dated May 7,1981. In this letter, the licensee was required to submit a revised Fire Hazards Analysis (FHA) before the exemption request would be considered further. The FHA was submitted by the licensee on June 15, 19B2.

• PAGE /70F J 69 III. Prior to the issuance of Appendix R. THI-2 had been reviewed the criteria of Appendix A to the Branch Technical Position 9.5-1 (BTP 9.5-1). The BTP 9.5-1 was developed to resolve the lessons learned from the fire at Browns Ferry Nuclear Plant. It is broader in scope than Appendix R. formed the nucleus of the criteria developed further in Appendix R and its present. revised form constitutes the section of the Standard Review Plan used for the review of applications for construction permits and operating licenses of new plants. The review of the Fire Hazards Analysis based on Appendix Rand BTP 9.5-1 was completed by the NRC staff and its fire protection consultant and a Fire Protection Safety Evaluation Report (FPSER) was provided to the staff by the consultant on February 28, 1983. Even though the fire hazards analysis was acceptable.

several suggestions were proposed by our contractor relative to the licensee's fire protection program. These suggestions were discussed in a letter to the licensee dated March 13. 1984. With respect to items relating to safe shutdown capability.

the staff agrees wi th the licensee that the THI-2 reactor is in a col d shutdown condi ti on wi th no active systems required for core cooling. However. certain instrumentation is required for monitoring various parameters such as reactor coolant temperature and neutron flux level to that a cold shutdown condition is maintained.

Additionally.

several backup systems are required which can provide makeup and maintain pressurization for the reactor coolant system if necessary.

It is t-the staff's opinion that even though Appendix R requirements are not priate for the unique conditions at TMI-2. the Proposed Technical SpeCifications and the Recove"ry Operations Plan would be acceptable as an alternative location for specific fire protection requirements for systems used to maintain and verify that cold shutdown.

Therefore, it is our position that systems used for monitoring or maintaining the reactor in a stable cold shutdown condition (e.g. monitoring instrumentation, the Mini-Decay Heat Removal System and the Standby Pressure Control System) should have fire protection features.

A summary of present and proposed fire protection features for systems required to maintain or monitor a cold shutdown as discussed above should be submitted to the NRC in addition to a change to your Technical Specifications to include these proposed features within 60 days of the date of this exemption. , . With regard to the Oil Collection System for reactor coolant pumps, the staff finds that an exemption to the schedular requirements of 10 CFR 50.48(c) is warranted because of the shutdown condition of TMI-2 and the prohibition to operate the pumps per the technical specifications.

IV. Accordingly, the Commission has detenmined that, pursuant to 10 CFR 50.12, an exemption is authorized by law and w111 not endanger life or property or the common defense and security and is otherw1se in the public interest.

The Commission hereby grants the following exemptions with respect to the requirements of 10 CFR Part 50.48{c): (1) Extend until the end of the Recovery Mode. the date for filing tional exemptions or complying with the provisions of Section lII.G as required by 50.48(c).

(2) Extend until the end of the Recovery Mode. the date for filing tions or complying with the provisions of Section 111.0 as required by 50.48(c);

The NRC staff has determined that the granting of this Exemption will not result in any significant environmental impact and that pursuant to 10 CFR 51.5(d)(4) an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with '-" this action. Dated at Bethesda, Maryland this 18th day of May, 1984. FOR THE NUCLEAR REGULATORY COMMISSION Harold R. Denton. Director Office of Nuclear Reactor Regulation ENCLOSURE 2 CLARIFICATION OF GENERIC LETTER On February 20, 1981, generic letter 81-12 was forwarded to all reactor licensees with plants licensed prior to January 1,1979. The letter restated the ment of Section 50.48 to 10 CFR Part 50 that each licensee would be required to reassess areas of the plant where cables or equipment including associated non-safety circuits of redundant trains of systems necessary to achieve and maintain hot shutdown conditions are located to determine whether the ments of Section III.G.2 of Appendix R to 10 CFR.50 were satisfied. ally, Enclosure 1 and Enclosure 2 of the generic letter requested additional information concerning those areas of the plant requiring alternative shutdown capability.

Section 8 of Enclosure 1 requested information for the systems, equipment and procedures of alternative shutdown capability and Enclosure 2 defined associated circuits and requested information concerning associated circuits for those areas requiring alternative shutdown.

In our review of licensee submittals and meetings with licensees, it has become apparent that the request for information should be clarified since a lack of clarity could result in the submission of either insufficient or excessive information.

Thus, the staff has rewritten Section 8 of Enclosure 1 and Enclosure 2 of th'e February 20, 1981 generic letter. Additionally, further clarification of the definition of associated circuits has been provided to aid in the reassessments to determine compliance with the requirements of Sections III.G.2 and IlI.G.3 of Appendix R. In developing this rewrite we have considered the comment of the Nuclear Utility Fire Protection Group. The attached rewrite of the Enclosures contains no new requirements but merely attempts to clarify the request for additional information.

c* Licensees who -have not responded to the February 20, 1981 generic letter, may choose to respond to the enclosed request for information.

Since the enclosed request for information 1s not new, but merely clarification of our previous letter, responding to it should not delay any submittals in progress that are based upon February 20, 1981 letter. Licensees whose response to the February 20, 1981 letter, has been found incomplete resulting in staff identification of a major unresolved item (i.e., ciated circuits), may choose to respond to pertinent sections of the enclosed request for information in order to close open items (i.e., open item for associated circuits, use rewrite to Enclosure 2). If additional clarification is needed, please contact the staff Project " . Manager for your plant .

• Attachment 1 to Enclosure 2 REWRITE OF SECTION 8 REQUEST FOR ADDITIONAL INFORMATION The following is a rewrite of the staff's request for additional information concerning design modifir.ation to meet the requirements of Section III.G.3 of Appendix R. The following contains no new requests but is merely a rewording of Section 8 of Enclosure 1 of the February 20, 1981 generic letter. 1. Identify those areas of the plant that will not meet the requirements of Section III.G.2 of Appendix R and. thus alternative shutdown will be vided or an exemption from the requirements'of Section III.G.2 of Appendix R will be provided.

Additionally provide a statement that all other areas of the plant are or will be in with Section III.G.2 of Appendix R. For each of those fire areas of the plant requiring an alternative shutdown system(s) provide a complete set of responses to the following requests for each fire area: a. List the system(s) or portions thereof used to provide the shutdown capability with the loss of offsite power. b. For those systens identified in "laM for which alternative or dedicated shutdown capability must be provided, list the equipment and cOO1ponents of the normal shutdown system in the fire area and identify the functions of the circuits of the normal shutdown system in the fire area (power to what equipment, control of what components and instrumentation).

Describe the system(s) or portions thereof used to provide the alternative shutdown capability for the fire area and provide a table that lists the equipment and components of the alternative shutdown system for the fire area.

,-c. For each alternative system identify the function of the new circuits being provided.

Identify the location (fire zone) of the alternative shutdown equipment and/or circuits that bypass the fire area and verify that the alternative shutdown equipment and/or circuits are separated from the fire area in with Section III.G.2. Provide drawings of the alternative shutdown system(s) which highlight any connections to the normal shutdown systems (P&IDs for piping and components, elementary wiring diagrams of electrical cabling).

Show the electrical location of all breakers for power cables, and isolation devices for control and instrumentation circuits for the alternative shutdown systems for the fire area. d. Verify that changes to safety systems will not degrade safety systems; (e.g., new isolation switches and control switches should meet design criteria and standards in the FSAR for electrical equipment in the system mounted in should also meet the same criteria (FSAR) as other safety related cabinets and panels; to avoid inadvertent isolation from the control room, the isolation switches should be keylocked or alarmed in the control room if in the "local" or "isolated" position; periodic checks should be made to verify that the switch is in the proper position for normal operation; and a single transfer switch or other new device should not be a source of a failure which causes loss of redundant safety systems).

,,-. e. Verify that licensee procedures have been or will be developed which describe the tasks to be performed to effect the shutdown method. vide a summary of these procedures outlining operator actions. f. Verify that the manpower required to perform the shutdown functions using the procedures of e. as well as to provide fire brigade members to fight the fire is available as required by the fire brigade technical specifications.

g. Provide a commitment to perform adequate acceptance tests of the alternative shutdown capability.

These tests should verify that: equipment operates from the local control station when the transfer or isolation switch is placed in the ulocalu position and that the equipment cannot be operated from the control room; and that equipment operates from the control room but cannot be operated at the local control station when the transfer isolation switch is in the "remote" position.

h. Provide Technical Specifications of the surveillance requirements and limiting conditions for operation for that equipment not already covered by existing Technical Specifications.

For example, if new isolation and control switches are added to a shutdown system, the existing Technical Specification surveillance requirements should be supplemented to verify system/equipment functions from the ndte shutdown station at testing intervals consistent with the lines of Regulatory Guide 1.22 and IEEE 338. Credit may be taken for other existing tests using group overlap test concepts.

l PAGE ZS-OF. Co 'i i. For new equipment comprising the alternative shutdown capability, verify that the systems available are adequate to perform the sary shutdown function.

The functions required should be based on previous analyses, if possible (e.g., in the FSAR), such as a loss of normal ac power or shutdown on Group 1 isolation (BWR). The equipment required for the alternative capability should be the same or lent to that relied on in the above anlaysis.

j. Verify that repair procedures for cold systems are developed and material for repairs is maintained on site. Provide a summary of these procedures and a list of the material needed for repairs *

• PAGE 2.(,. _ ut (p ( Attachment 2 to Enclosure 2 SAFE SHUTDOWN CAPABILITY The following discusses the requirements for protecting redundant and/or alternative equipment needed for safe shutdown in the event of a fire. The requirements of Appendix R address hot shutdown equipment which must be free of fire damage. The following requirements also apply to cold shutdown equipment if the licensee elects to demonstrate that the equipment is to be free of fire damage. Appendix R does allow repairable damage to cold shutdown equipment.

Using the requirements of Section III.G. and III.L of Appendix R, the bility to achieve hot shutdown must exist given a fire in any area of the plant in conjunction with a loss of offsite power for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.Section III.G \. . of Appendix R provides four methods for ensuring that the hot shutdown bility is protected from fires. The first three options as defined in Section III.G.2 provides methods for protection from fires of equipment needed for hot shutdown:

1. Redundant systems including cables, equipment, and associated circuits may be separated by a three-hour fire rated barrier; or, 2. Redundance systems including cables, equipment and associated circuits may be separated by a horizontal distance of more than 20 feet with no vening combustibles.

In addition, fire detection and an automatic fire suppression system are required; or, l* PAGE 2-7 OF 0 '( 3. Redundant systems including cables, equipment and associated circuits may be enclosed by a one-hour fire rated barrier. In addition, 'fire detectors and an automatic fire suppression system are required.

The last option as defined by Section III.G.3 provides an alternative shutdown capability to the redundant trains damaged by a fire. 4. Alternative shutdown equipment must be of the cables, ment and associated circuits of the redundant systems damaged by the fire. Associated Circuits of Concern The following discussion provides; A) a definition of associated circuits for Appendix R consideration, B) the guidelines for protecting the safe shutdown capability from the fire-induced failures of associated circuits and C) the information required by the staff to review associated circuits.

The nition of associated circuits has not changed from the February 20, 1981 generic letter; but is merely clarified.

It is important to note that our interest is only with those circuit (cables) whose fire-induced failure could effect snutdown.

The guidelines for protecting the safe shutdown capability from the fire-induced failures of associated circuits are nQ! reqUirements.

These guidelines should be used only as guidance when needed. These guidelines do not limit the alternatives available to the licensee for protecting the shutdown capability.

All proposed methods for protection of the shutdown capability from fire-induced failures will be evaluated by the staff for acceptability.

I PAGE OF. Co '1 A. Our concern is that circuits within the fire area will receive fire damage which can affect shutdown capability and thereby prevent Post-fire safe shutdown.

Associated Circuits*

of Concern are defined as those cables (safety related, non-safety related, Class IE, and non-Class IE) that: 1. Have a physical separation less than that required by Section 111.G.2 of Appendix R, and; 2. Have one of the following:

a. a common power source with the shutdown equipment (redundant or alternative) and the power source is not electrically protected from the circuit of concern by coordinated breakers, fuses, or similar devices (see diagram 2a), or b. a connection to circuits of equipment whose spurious operation would adversely affect the shutdown capability (e.g., RHR/RCS isolation vavles, ADS valves, PORVs, steam generator atmospheric dump valves, instrumentation, steam bypass, etc.) (see diagram 2b), or c. a common enclosure (e.g., raceway, panel, junction) with the shutdown cables (redundant and alternative) and, *The definition for associated circuits is not exactly the same as the definition presented in IEEE-384-1977.

l PAGE zy OF &1 (1) are not electrically protected by circuit breakers, fuses or similar devices, or (2) will allow propagation of the fire into the common enclosure, (see diagram 2c). B. The following guidelines are for protecting the shutdown capability from fire-induced failures of circuits (cables) io'the fire area. The guidance provided below for interrupting devices applies only to new devices stalled to provide electrical isolation of associated circuits of concern, or as part of the aiternative or dedicated shutdown system. The shutdown capability may be protected from the adverse effect of damage to associated circuits of concern by the following methods: 1. Provide protection between the associated circuits of concern and the shutdown circuits as per Section III.G.2 of Appendix R, or 2. a. For a common power source case of associated circuit: Provide load fUse/breaker (interrupting devices) to feeder fuse/breaker coordination to prevent loss of the redundant or alternative shutdown power source. To ensure that the following coordination criteria are met the following should apply:

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• I ! . I I . F.I\E, t* I I cC 'L... 1 *l -. II ( -----a: I r [! C"""c The area barrters shown above meet th& appropriate sub-paragraphs' (a-1 of sectton of Appendtx R * ; Dtagrall 2C * , (1) The associated circuit of concern interrupting devices (breakers or fuses) time-overcurrent trip characteristic for all circuits faults should cause the interrupting device to interrupt the fault current prior to initiation of a trip of any upstream interrupting device which will cause a loss of the common power source, (2) The power source shall supply the necessary fault current for sufficient time to ensure the proper coordination without loss of function of the shutdown loads. The acceptability of a particular interrupting device is considered demonstrated if the following criteria are met: (i) The interrupting device design shall be factory tested to verify overcurrent protection as designed in accordance with the applicable UL, ANSI, or NEMA standards. (ii) For and medium voltage switchgear (480 V and above) circuit breaker/protective relay periodic testing shall demonstrate that the overall coordination scheme remains within the limits specified in the design criteria.

This testing may be performed as a series of overlapping tests. PAGE 2-OF. fo '7 (iii) Molded case circuit breakers shall periodically be manually exercised and inspected to insure ease of operation.

On a rotating refueling outage basis a sample of these breakers shall be tested to detenmine that breaker drift is within that allowed by the design criteria.

Breakers should be tested in accordance with an accepted QC testing methodology such as MIL STD 10 5 D. {iv} Fuses when used as interrupting devices do not require periodic testing, due to their stability, lack of drift, and high reliability.

Administrative controls must insure that replacement fuses with ratings other than those selected for proper coordinating are not accidentally used. b. For of equipment and/or components whose spurious operation would affect the capability to safely shutdown:

(1) provide a means to isolate the equipment and/or components from. the fire area prior to the fire (i.e., remove power cables, open circuit breakers);

or (2) provide electrical isolation that prevents spurious operation.

Potential isolation devices include breakers, fuses, fiers, control switches, current XFR$, fiber optiC couplers, relays and transducers; or

, " (3) provide a means to detect spurious operations and then dures to defeat the maloperation of equipment

{i.e., closure of the block valve if PORV spuriously operates, opening of the breakers to remove spurious operation of safety injection};

c. For common enclosure cases of associated circuits: (l) provide appropriate measures to prevent propagation of the fire; and (2) provide electrical protection (i.e., breakers, fuses or similar devices) c. We recognize that there are different approaches which may be used to reach the*same objective of determining the interaction of associated circuits with shutdown systems. One approach is to start with the fire area, identify what is in the fire area, and determine the interaction between what is in the fire area and the shutdown systems which are outside the fire area. Ue have entitled this approach, uThe Fire Area Approach.

II A second approach which we have named liThe Systems Approachll would be to define the shutdown systems around a fire area and then determine those circuits that are located in the fire area that are associated with the shutdown system. We have prepared two sets of requests for information, one for each approach.

The licensee may choose to respond to either set of requests depending on the approach selected by the licensee.

\. FIRE AREA APPROACH 1. For each fire area where an alternative or dedicated shutdown method, in accordance with Section III.G.3 of Appendix R is provided, the following infonmation is required to demonstrate that associated circuits will not prevent operation or cause maloperation of the alternative or dedicated shutdown method: a. Provide a table that lists all the power cables in the fire area that connect to the same power supply of the alternative or dedicated shutdown method and the function of each power cable listed (i.e., power for RHR pump). b. Provide a table that lists all the cables in the fire area that were considered for possible spurious operation which would adversely affect shutdown and the function of each cable listed. c. Provide a table that lists all the cables in the fire area that share a common enclosure with circuits of the alternative or dedicated shutdown systems and the function of each cable listed. d. Show that fire-induced failures (hot shorts, open circuits or shorts to ground) of each of the cables 11stes in at b t and c will not prevent operation or cause malcperat10n of the alternative or dedicated shutdown method. , I e. For each cable listed in a, b, and c where new electrical isolation has been provided or modification to existing electrical isolation has been made, provide detailed electrical schematic drawings that show how each cable is isolated from the fire area. SYSTEMS APPROACH 1. For each area where a"n al ternative or dedicated shutdown method, in accordance with Section III.G.3 of Appendix provided, the following information is required to demonstrate that associated circuits will not prevent operation or cause maloperation of the alternative or dedicated shutdown method: a. Describe the methodology used to assess the potential of associated circuit adversely affecting the alternative or dedicated shutdown.

The description of the methodology should include the methods used to identify the circuits which share a common power supply or a common enclosure with the alternative or dedicated shutdown system and the circuits whose spurious operation would affect shutdown.

Additionally, the description should include the methods used to identify if these circuits are associated circuits of concern due to their location in the fire area. b. Provide a table that lists all associated circuits of concern located in the fire area.

\.. c. Show that fire-induced failures (hot shorts, open circuits or shorts to ground) of each of the cables listed in b will not prevent operation or cause maloperation of the alternative or dedicated shutdown method. d. For each cable listed in b where new electrical isolation has been provided, provide detailed electrical schematic drawings that show how each cable is isolated from the fire area; e. Provide a location at the site or other offices where all the tables and drawings generated by this methodology approach for the associated circuits review may be audited to verify the information provided above. HIGH-LOW PRESSURE INTERFACE For either approach chosen the following concern dealing with high-low pressure interface should be addressed.

2. The residual heat removal system is generally a low pressure system that interfaces with the high pressure primary coolant system. To preclude a LOCA through this interface, we require compliance with the recommendations of Branch Technical Position RSB 5-1. Thus, the interface most likely consists of two redundant and independent motor operated valves. These two motor operated valves and their associated cables may be subject to a single fire hazard. It is our concern that this single fire cause the two valves to open resulting in a fire initiated LOCA through the high-low pressure system interface.

To assure that this interface ad other high-low pressure interfaces are adequately protected from the effects of a single fire, we require the following information:

a. Identify each high-low pressure interface that uses redundant electrically controlled devices (such as two series motor operated valves) to isolate or preclude rupture of any primary coolant. b. For each set of redundant valves identified in a., verify the redundant cabling (power and control) have adequate physical separation as required by Section III.G.2 of Appendix R. c. For each case where adequate separation is not provided, show that fire induced failures (hot short, open circuits or short to ground) of the cables will not cause maloperation and result in a LOCA.

CRITERIA FOR EVALUATING EXEMPTIONS TO SECTION III G OF APPENDIX R OF 10 CFR PART 50 Enclosure 3 Paragraph 50.48 Fire Protection of 10 CFR Part 50 requires that all nuclear power plants licensed prior to January 1,1979 satisfy the requirements of Section III.G of Appendix R to 10 CFR Part 50. It also requires that native fire protection configurations, previously approved by an SER be reexamined for compliance with the requirements of Section III.G.Section III.G is related to fire protection features for ensuring that systems and associated circuits used to achieve and maintain safe shutdown are free of fire damage. Fire protection configurations must either meet the specific requirements of Section III.G or an alternative fire protection configuration.

must be justified by a fire hazard analysiS. . The general criteria for acceting an alternativ.e fire protection configurations are the fol lowing: o The alternative assures that one train of equipment necessary to achieve hot shutdown from either the control room or emergency control stations is free of fire damage. . o The alternative assures that fire damage to at least one train of ment necessary to achieve cold shutdown is limited such that it can be \. repaired within a reasonable time (minor repairs with components stored on-site).

o Fire retardant coatings are not used as fire barriers. , o Modifications required to meet Section III.G would not enhance fire protection safety above that provided by either existing or proposed alternatives.

o Modifications required to meet Section III.G would be detrimental to overall facility safety. Because of the broad spectrum of potential configurations for which exemptions may be requested, specific criteria that account for all of the parameters that are important to fire protection and consistent with safety requirements of all plant-unique configurations have not been developed.

However, our evaluations of deviations from these requirements in our previous reviews and in the requests for III.G exemptions received to date have identified some recurring configurations for which specific criteria have been developed.

\ Section III.S.2 accepts three methods of fire protection.

A passive 3-hour fire barrier. should be used where possible.

Where a fixed barrier cannot be installed, an automatic suppression system in combination with a fire barrier or a separation distance free of combustibles is used if the configurations of systems to be protected and in-situ combustibles are such that there is reasonable assurance that the protected systems will survive. If this latter condition is not met, alternative shutdown capability is required and a fixed suppression system installed in the fire area of concern, if it contains a large concentration of cables. It is essential to remember that these al ternative requirements are not deemed to be However. they provide adequate protection for those urations in they are accepted.

When the fire protection features of each fire area are evaluated, the whole system of such features must be kept in perspective.

The defense-in-depth principle of fire protection program is aimed at achieving an adequate balance between the different features.

Strengthening anyone can compensate in some measure for weaknesses, known or unknown in others. The adequacy of fire protection for any particular plant safety system or area is determined by analysis of the effects of postulated fire relative to taining the ability to safely shutdown the plant and minimize radioactive releases to the environment in the event of a fire. During these tions it is necessary to consider the two-edged nature of fire protection features recognized in General Design Criterion 3 namely, fire protection should be provided consistent with other safety considerations.

An evaluation must be made for each fire area for which an exemption is requested.

During these evaluations, the staff considers the following pa rameters : A. Area Description

-walls, floor, and ceiling construction

-ceil ing height -room vol urne -ventilation

-congestion B. Safe Shutdown Capability

-number of redundant systems in area -whether or not system or equipment is required for hot shutdown -type of equipment/cables involved -repair time for cold shutdown equipment within this area -separation between redundant components and in-situ concentration of combustibles

-alternative shutdown capability

. ,--3-C. Fire Hazard Analysis -type and configuration of combustibles in area -quantity of combustibles

-ease of ignition and propagation

-heat release rate potential

,-transient and installed combustibles

-suppression damage to equipment

-whether the area is continuously manned -traffic through the area -accessibility of the area D. Fire Protection Existing or Committed

-fire detection systems -fire extinguishing systems -hose station/extinguisher

-radiant heat shields A specific description of the fire protection features of the configuration is required to justify the compensating features of the alternative.

Low fire loading is not a sufficient basis for granting an exemption in areas where there are cables. If necessary, a team of experts, including a fire protection engineer, will visit the site to determine the existing circumstances.

This visual inspection is also considered in the review process. The majority of the 111.G exemption requests received to date are being denied because they lack specificity.

Licensees have not identified the extent of the exemption requested, have not provided a technical basis for the request and/or have not provided a specific description of the alternative.

We expect to receive requests for exemption of the following nature: 1. Fixed fire barriers less than 3-hour rating. 2. Fire barrier without an automatic fire suppression system. 3. Less than 20 jeet separation of cables with fire propagation retardants (e.g., coatings, blankets, covered trays) and an automatic suppression system. 4. For large open areas with few components to be protected and few in-situ combustibles, no automatic suppression system with separation as in Item 3 above. 5. No fixed suppression in the control room

• p\GE 4( t*Ot &; 1 6. No fixed suppression in areas without a large concentration of cables for which alternative shutdown capability has been provided.

Our fire research test program is conducting tests to provide information that will be useful to determine the of acceptable conditions for fire protection configurations which do not include a fire rated barrier. Based on deviations recently approved, specific criteria for certain recurring configurations are as follows: Fire Barrier Less than Three Hours This barrier is a wall, floor, ceiling or an enclosure which separates one fire area from another. Exemptions may be granted for a lower rating (e.g., one hour or two hours) where the fire loading is no more than 1/2 of the barrier rating. The fire rating of the barrier shall be no less than one hour. Exemptions may be granted for a fixed barrier with a lower fix rating supplemented by a water curtain. An Automatic Suepression System With Either One Hour Fire Barrier or 20-Foot This barrier is an enclosure which separates those portions of one division which are within 20 feet of the redundant division.

The suppressant may be water or gas. Exemptions may be granted for configurations of redundant systems which have compensating features.

For example: A. Separation distances less than 20 feet may be deemed acceptable where: 1. Fire propagation retardants (i.e., cable coatings, covered trays, conduits, or mineral wool blankets) assure that fire propagation through in-situ combustibles will not occur or will be delayed suffic.i ently to ensure adequate time for detection and suppress ion. 2. Distance *above a floor level exposure fire and below ceiling assures that redundant systems will not be simultaneously subject to an unacceptable temperature or heat flux. B. The ommission of an automatic suppression system may be deemed acceptable where: 1. Distance above a floor level exposure fire and below ceiling assures that redundant systems will not be simultaneously subject to an unacceptable temperature or heat flux. 2. The fire area is required to be manned continuously by the visions in the Technical Specifications.

\ " .. Enclosure 4 UNITED STATES NUCLEAR REGULATORY COMMISSION DOCKET NO. 50-320 GENERAl PUBLIC UTILITIES NUCLEAR CORPORATION NOTICE OF ISSUANCE OF EXEMPTION 10 CFR The U.S. Nuclear Regulatory Commission (the Commission) has issued an Exemption from some of the schedular requirements of 10 CFR 50.48(c) relative to the fire protection requirements of 10 CFR 50 Appendix R. GPU Nuclear Corporation, Metropolitan Edison Company, Jersey Central Power and Light Company and Pennsylvania Electric Company (collectively, the licensee) are the holders of Facility Operating License No. DPR-73, which had authorized operation of the Three Mile Island Nuclear Station, Unit 2 (TMI-2) at power levels up to 2772 megawatts thermal. The facility, which is located in Londonderry Township, Dauphin County, Pennsylvania, is a pressurized water reactor previously used for the commercial generation of electricity.

By Order for Modification of License, dated July 20, 1979, the licensee's authority to operate the facility was suspended and the licensee's authority was limited to maintenance of the facility in the present shutdown cooling mode (44 Fed. Reg. 45271). By further Order of the Director, Office of Nuclear Reacto.r Regulation, dated February 11, 1980, a new set of fonnal license requirements was imposed to reflect the post-accident condition of the facility and to assure the continued maintenance of the current safe, stable. long-tenn cooling condition of the facility (45 Fed. Reg. 11292). This license provides, among other things, that it is subject to all rules, regulations and Orders of the Commission now or hereafter in effect.

\. On November 19. 1980. the Commission published a revised Section 10 CFR 50.48 and a new Appendix R to 10 CFR 50 regarding fire protection features of nuclear power plants (45 FR 76602). The revised Section 50.48 and Appendix R became effective on February 17. 1981. Section 50.48(c) lished the schedules for satisfying the provisions of Appendix R.Section III of Appendix R contains 15 subsections.

lettered A through 0. each of which specifies requirements for a particular aspect of the fire protection features at a nuclear power plant. Two of these 15 subsections.

III.G and 111.0, are the subject of this Exemption.

Subsection III.G specifies detailed ments for fire protection of the equipment used for safe shutdown by means of separation and barriers (III.G.2).

If the requirements for separation and barriers could not be met in an area. alternative safe shutdown capability, independent of that area and equipment in that area, was required (III.G.3).

Subsection 111.0 required that the reactor coolant pump be equipped with an oil collection system if the containment is not inerted during nonnal operation.

The system had to be capable of collecting lube oil from all potential pressurized and unpressurized leakage sites in the reactor coolant pump lube oil systems. Section 50.48(c) required completion of all modifications to meet the provisions of Appendix R within a specified time from the effective date of this fire protection rule. February 17. 1981. except for modifications to provide alternative safe shutdown capability.

These latter modifications

\, (III.G.3) require NRC review and approval and Section 50.48(c) requires their completion within a certain time after NRC approval.

The date for submittal of design descriptions of any modifications to provide alternative safe shutdown capability was specified as March 19.1981. By letter dated March 24, 1981. the licensee requested exemptions from 10 CFR 50.48(c) with respect to the requirements of Section III.G and 111.0 of Appendix R as follows: " (l) Extend until the end of the Recovery Mode the date for filing additional exemptions or complying to the requirements, plans and schedule to achieve compliance with Section lIl.G as required by 50.48(c).

(2) Extend until the end of the Recovery Mode, the date for filing additional exemptions or complying to the requirements, plans and schedule to achieve compliance with Section 111.0 as required by 50.48(c).

Prior to the issuance of Appendix R. TMI-2 had been reviewed against the criteria of .Appendix A to the Branch Technical Position 9.5-1 (BTP 9.5-1). The BTP 9.5-1 was developed to resolve the lessons learned from the fire at Browns Ferry Nuclear Plant. It is braoder in scope than Appendix R, fonned the nucleus of the criteria developed further in Appendix R and its present. revised form constitutes the section of the Standard Review Plan used for the review of applications for construction pennits and operating licenses of new plants. The review of the Fire Hazards Analysis based on Appendix Rand BTP 9.5-1 was completed by the NRC staff and its fire protection

• . " consultant and a Fire Protection Safety Evaluation Report (FPSER) was provided to the staff by the consultant on February 28. 1983. Even though the fire hazards analysis was acceptable.

several suggestions were proposed by our contractor relative to the licensee's fire protection program. These suggestions are discussed in separate correspondencei-. With respect to items relating to safe shutdown capability, the staff agrees with the licensee that the TMI-2 reactor is in a cold shutdown condition with no active systems required for cooling. However, certain instrumentation is required for monitoring various parameters such as reactor coolant temperature and neutron flux level to insure that a cold shutdown condition is maintained.

Additionally, several backup systems are required which can provide makeup and maintain pressurization for the reactor coolant system if necessary.

It is the staff's opinion that even though Appendix R requirements are not appropriate for the unique conditions at lt11-2, the Proposed Technical Specifications and the Recovery Operations Ptan would be acceptable as an alternative location for specific fire protection requirements for systems used to maintain and verify that cold shutdown.

Therefore.

it is our position that systems used for monitoring or maintaining the ,reactor in a stable cold shutdown condition (e.g., monitoring instrumentation, the Mini-Decay Heat Removal System and the Standby Pressure Control System) should have fire protection features.

A summary of present and proposed fire protection features for systems required to maintain or I!1Onitor a cold shutdown as discussed above should be submitted to the NRC in addition to a change to your Technical Specifications to include these features within 60 days of the date of the exemption. With regard to the Oil Collection System for reactor coolant pumps, the staff finds that an exemption to the schedular requirements of 10 CFR 50.48(c) is warranted because of the shutdown condition of TMI-2 and the prohibition to operate the pumps per the technical speCifications.

Accordingly, the Commission has determined that, pursutant to 10 CFR 50.12, an exemption is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest and hereby grants the following exemptions with respect to the requirements of 10 CFR Part 50.48(c).

(1) Extend until the end of the Recovery Mode. the date for filing tional exemptions or complying to the requirements.

plans and schedule to achieve compliance with Section III.G as required by 50.48(c);

(2) Extend until the end of the Recovery Mode. the date for filing tions or complying to the requirements.

plans and schedule to achieve compliance with Section 111.0 as required by 50.48(c).

The NRC staff has determined that the granting of this Exemption will not result in any significant environmental impact and that pursuant to 10 CFR 5l.5(d)(4) an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with this action. The exemption complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act). and the Commission's rules and regulations.

The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter 1, which are set forth in the license amendment.

Prior public notice of this -exemption was not requi red since it does not invol ve a significant hazards, consideration.

For further details with respect to this action, see the exemption request dated March 24, 1981. This item is available for public inspection at the Commission's Document Room, 1717 H Street, N.W., Washington, D.C. 20555 and at the Government Publications Section, State Library of Pennsylvania 17126. A copy may be obtained upon request addressed to the U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, Attention:

Program Director, TMI Program Office, Office of Nuclear Reactor Regulation.

Dated at Bethesda, Maryland this 18th day of May , 1984. FOR THE NUCLEAR REGULATORY Cor-mISSION Bernard J.

Director Three Mile Island Program Office Office of Nuclear Reactor Regulation

" '-. / ,. .. 11j .... ij .... j J Nuclear D'\r'.E I \ l:' Memorandum

Subject:

MEETING WITH NRC RE 10 CFR 50 R EXEMPTION REQUEST Date: May 7, 1984 4410-84-M-0280 From: TMI-2 Licensing Engineer S. D. Chaplin Location:

TMI-2 Licensing To: Director, Licensing and Nuclear Safety R. E. Rogan At 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> on May 4, 1984, Tom Poindexter (TMI Program Office NRC), Jim Quinnette (TMI-2 Fire Protection Engineer), and myself met to discuss NRC's action on GPU Nuclear's request for partial exemption from the implementation schedule for Appendix R Fire Protection requirements.

Basically, NRC will issue a full exemption in the near future. However, NRC will require GPU Nuclear to provide a list of fire protection provided for equipment necessary to "maintain and monitor cold shutdown", and to incorporate said equipment into the fire protection Technical Specifications.

Those systems necessary for "maintenance and monitoring" the current condition of the core are identified.

However, after headlift, there is a significant potential that the list will change and a risk analysis is needed to define this list so that GPU Nuclear can identify which post lift systems are needed. Once identified, these systems will form the basis for modifying the fire protection Technical Specifications and updating the Fire Protection Program Evaluation (FPPE). NRC agrees that the FPPE is not necessarily a real-time document, i.e., it is not mandatory to update it prior to headlift.However, NRC emphasized that those systems necessary for*maintaining and monitoring cold shutdown must be adequately protected, regardless of FPPE status. If you have any questions, please let me know . ..4'-# S. D. Cb:aplin Extension 8693 SDC/jep CC: See Page 2 AOOO0648 8*83

.. \ R. E. Rogan cc: Deputy Director, TMI-2, J. J. Barton May 7, 1984 4410-84-M-0280 p.....= 4Cf Co OJ \\3 ... I ( ..... -( TMI-2 Licensing and Nuclear Safety Director, J. E. Larson Site Operations Director, S. Levin Manager, TMI-2 Licensing, J. J. Byrne Lead Systems Engineering Supervisor, R. P. Warren Fire Protection Engineer, J. W. Quinnette CARIRS -TMI UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 Docket No. 50-320 Mr. B. K. Klnga, Director Three Mile Island Unit 2 GPU Nuclear Corporation P.O. Box 480 Middletown, PA 17057

Dear Mr. Kanga:

March 13, 1984 ? \ 3':: S(L .... Qfl (

Subject:

Three Mile Island Nuclear Station, Unit 2 Operating License No. DPR-73 ' Docket No. 50-320 Fire Hazards Analysis The staff has reviewed your Fire Hazards Analysis (FHA) dated June 15, 1982 which was in support of your request for en exemption from the schedular requirements of 10 CFR 50.48{c) and 10 CFR 50. Appendix R,Section III.G and 111.0. During the period November 29 through December 3, 1982, the NRC's fire protection consultant inspected your facility and provided the staff with several suggestions that should be incorporated into your program. Enclosed is a list of modifications that should be made to fire protection systems or to your discussion of those systems in your FHA. Since our inspection was over a year ago, you may have already corrected many items on the enclosed list. Therefore, for items that have already been corrected, you should provide the staff with information describing how the corrections were made. For those items that have not yet been addressed by your staff, please vide a schedule for corrective actions. You should provide this information within 30 days of receipt of this letter.

Enclosure:

As stated cc: .J.. Barton V{). Byrne J. Larson Service Distribution List (see attached)

Sincerely, Bernard J. Snyder, Program Di rector Three Mile Island Program Office Office of Nuclear Reactor Regulation

...... l* ENCLOSURE GPU shall the fire protection program evaluation to include the following general requirements:

1. The analysis that you provided for the turbine building, oil-drum storage rooms, and 305 foot elevation of the fuel handling building was not -date in your Fire Hazards submittal.

Therefore the analysis for these areas should be revised. 2. A revision of the fuel loading calculations to correct the errors in heats of combustion.

3. Complete descriptions of the manner in which the licensee considers that each requirement of Appendix A to BTP-APCSB 9.5-1 has been met. 4. A discussion of potential releases of radioactive materials from the Solid Waste Staging Building resulting from the use of fire suppression water. 5. Additional information on any safety-related cable trays that may be in the vicinity of Tygon tubing. . 6. Resolution of the apparent conflict between Sections E.l (b) and F.14 of your Fire Hazards Analysis regarding alarms for the Solid Waste Staging Building detection system. \ 7. A compilation of a comprehens ive list of fire protection systems. periodic testing requirements and the parties responsible for carrying out this testing. 8. Provision for the operation of detection and suppression systems in the Chemical Cleaning Building complex to alarm and annunciate at the main fire protection panel. The following is a list of additional specific requirements that shall be met by GPUNC: 1. The leak over the interior hose near column N-63 in Zone 2 of Fire Area 9 should be stopped. 2. The fire hose at the base of the stairs in Zone 1 of Fire Area 9. near pump WG-P-l. should be properly installed on a hose reel in a readily accessible location. . 3. The main discharge valve for the fire pump in Fire Area 25 should be electrically supervised.

locked, or sealed open, and visually inspected periodically.

4. In Fire Area 47. the broken ceraf1ber board on cable tray 3226 should be replaced with an unbroken board. the untreated wood scaffolding should be removed or treated. and the stair door should be arranged to close easily.

.... .. ' .. " / . 5. The new fire hose in Fire Area 49 should be hydr9statically tested at the required pressures.

6. The fire detection cable in the filter unit of Fire Area 59 should be repaired or replaced as soon as possible.

7. All miscellaneous wood and metal scaffolding that is no longer being used should be moved to storage locations.

8. The sprinkler system in the fonner paint shed should be restored to service. 9. Hose stretch tests should be performed in all plant areas in which the arrangement of equipment or materials has been changed Significantly since the previous hose stretch test. . 10. The sprinkler system water supply connections for the DOE trailers should be hydrostatically tested to the required pressures, and tected from physical damage by a sturdy enclosure.

As an alternative, the hose may be replaced by piping installed in accordance with the provisions of NFPA 13-1980. 11. Your combustible loading inventory should be updated to reflect current conditions in the plant, including Fire Areas 1, 2, 7, 8, 20, 47 and 49. 12. The water-tight doors between Fire Areas 1 and 47 should be closed, or the two areas should be considered together in a revised Fire Hazards Analysis.

13. The folding steel fire door between the east end of the 305' elevation of the Turbine Building and the adjoining corridor should be closed completely, or the two areas should be considered together in a revised Fire Hazards Analysis.: . 14. The untreated plywood shoul d be removed from the 280'-6" elevation of Fire Area 7. 15. Compressed gas cylinders on the 347' elevation of Fire Area 7 should be properly secured in an upri9ht position.

16. A fire detection system should be provided in the decontamination room on the 328' elevation of Fire Area 9. 17. The adequacy of fire detection and suppreSSion equipment servin9 the protective clothing storage area on the 305' elevation of Fire Area 9 should be reevaluated.

18. Incorrectly hung fluorescent lights in Fire Area 27 should be corrected.

19. The use of the 280'-6* elevation of Fire Area 36 as a janitor's closet should be discontinued.

... Elyl] Nuclear ....., ....... P.O. Box 480 Middletown, Pennsylvania 17057 717-944-7621 THl Proaram Office Attn: Dr. B. J. Snyder. Proaram Director U. S. Nuclear "lu1atory Commission Washinaton, D.C. 20555

Dear Sir:

Writer's Direct Dial Number: June 15. 1982 4400-82-1.-0102 Three Mile Island Nuclear Station, Unit 2 (tKI-2) Operatina License No. DPR-73 Docket No. 50-320 Fire Protection Proaram Evaluation This letter supercedes GlU letter dated Kay 4, 1982 (440o-82-L-0073) which submitted the containment portion of the revised Fire Hazards Analysis (FHA) and a clarification of GlU's request for exemption from the implementation schedule of Appendix R to 10CFRSO specified in 10CFRSO. 48c. . Accordingly, the enclosed Revision 1 to the TMI-2 Fire Protection Program Evaluation (FPPE) is being submitted in response to your letter of May 7, 1981 which requested that GPU update the FHA (aee Attachment 3). This document is also beina provided to support NRC's evaluation of GlU's Appendix R schedule exemption request. (The FPPE consists of three basic parts, General Information, the BTl APCSB 9.5-1 Appendix A Comparison, and the FHA). The revision to the program evaluation reflects the preaent1y existing conditions at TMI-2. In addition, a brief of the imp1ementina procedures and the backaround used in determinina conformance to Branch Technical Position APCSB 9.5-1 is provided as Attachment

1. This information is provided to reaffirm GPU Nuclear's previous submittals on this subject and to provide an outline of the tKI-2 Fire Protection Proaram ation reference

• • Aa part of the evaluation effort, GPU bas determined that certain DeW .odifications are required to .. intain and .enitor the plant in its current aode. These aodification.

are discussed in Sectiona &.l(d), 1.4, and F.14. Bowever, a achedule for implementation of the .. cations bas not been completed.

GPU will provide the implementation for these items by July 15, 1982. GPU has also determined that certain SEa GPU Nuclear is a part of the General Public Utilities System

\. of Dr. B. J. Snyder 4400-82-L-ol02 modifications are no longer required for the aaintenance and monitoring of the cold shutdown reactor at tMI-2. A summary table of these cations and their disposition is provided as Attachment

2. The table also addresses the justification for not completing the SEa modifications with the present plant condition.

One area of TMI-2 fire protection not specifically identified in the attachments or the FPPE is the on-going fire protection requirements for TMI-2 and the methods used for continuing to meet those requirements.

The fire protection features of TMI-2 are essentially the same now as they were prior to the accident.

The installed fire protection equipment is still in place along with the inspection and testing requirements.

The fire protection organization is still in place. including the required QA/ QC support. These areas are described in both Attachment 1 and in the FPPE, although not under one specific requirement or section. One area not fully described in the FPPE is the fire protection input to the recovery modification and work activity review process. The recovery modification and work activity process will aake changes to TMI-2 that affect fire protection and/or the fire hazards within the plant. The recovery modification and work activity is governed by two basic mechanisms, the Engineering Change Memorandum and the Technical Evaluation Report. Both these mechanisms have provisions to ensure fire protection is addressed during the review process. That ensures the ongoing recovery effort fire protection requirements are met. In a letter dated March 24, 1981, an exemption was requested from the completion time requirements of lOCFRSO.48c for certain parts of Section III of Appendix a to lOCFR50 as well as open SEa items. Specifically.

the intention of this letter was to request an exemption under lOCFRSO.48c(6) for Section III G "Fire Protection of Safe Shutdown Capability".

Section III 0 "Oil Collection System for Reactor Coolant Pump", and the open SEa items outlined in the subject letter. Based additional review it has been determined that: o GPU's exemption request for Appendix a Section III G remains unchanged from the March 24, 1981 request. o Although the existing emergency lighting units are acceptable per the SEa. they do not satisfy the new Appendix a Section III J requirements.

However. GPU believes the installed units in the facility, as described in the FPPE, are of sufficient capacity for current plant shutdown conditions.

In addition, also described in the FPPE, there are portable, . {

..... -., Dr. B. J. Snyder 4400-82-L-0102 battery powered lights available to back up the installed lighting as necessary.

Therefore.

this request for exemption from the schedule implementation requirements of Appendix a J. is submitted as provided for in 10CFB.50.48c6 where "required modification would not enhance fire protection safety in the facility.

• .". This request supercedes the status specified in both Met-Ed letter dated March 24, 1981 (LL2-81-OO84) and GPU letter dated May 4. 1982 (440o-82-L-0083).

o The existing acp Oil Collection System satisfies the new Appendix a Section III 0 requirements.

Therefore.

the exemption requested for Section III 0 in Met-Ed letter dated March 24. 1981. is drawn. Sincerely.

lsI J. J. Barton J. J. Barton Acting Director.

TKI-2 JJB:SDC:djb Attachments:

1. Fire Protection Program Evaluation.

Implementing and Background Information List 2. Fire Protection Program Evaluation.

Table of Modifications and Disposition of Modifications

3. TKI-2 Fire Protection and Program Evaluation, aevision 1 cc: L. H. Barrett, Deputy Program Director -TKI Program Office

• 4400-82-L-ol02 ATTAcmmHT 1 'l'MI-2 Fire Protection Proar-Evaluation I!!p1ement:1Dg aDd Backaround Information L1st 1. BackgroUDd/Imp1eaent:1Dg Requireaents on file with the DC: 1. TKI-2 Final Safety ADa1ys1.a leport 2. TMI-2 Fire Protection Program Plan ILL-322 3. 'l'MI-2 Fire Protection Program Evaluation (bv. 0) 4. TKI-2 Emergency Plan 5. TMI-2 Technical Specifications

2. Background/Implementing Requirements not on file with the NRC: 1. Burns and Roe, Inc. Appendix A to BTP APCSB 9.5-1 Comparison dated 3/17/77. 2. Burns and Roe, Inc. TKI-2 Fire Hazards Analysis Update. W.O. 3682-01 3. tMI-2 Administrative.

Operations, Emergency and Maintenance Procedures

4. Emergency Plan Implementing Procedures

'o' leport Section It.l(d) E.3(c) ATTACHMENT 2 4400-82-L-OI02 Pa,e 1 of 4 PAGE r;7 OF. '( THI-2 Fire Protection Program Evaluation Table of MOdifications and Disposition of MOdifications Subject/Modification Fire Detection/

add detection to Chemical Cleaning Bldgs. Disposition Detection will be added. Bose Standpipe System/ 1. Control Building 351' elev. -not presently required install 14 hose stations required by the SER *. * -see notes 1,2,3,4,5

& 7 2. Control Building 351' elev. -not presently required -see notes 1,2,3,4,5

& 7 3. Control Building 280' elev. -not presently required -see notes 1,3,4,5,6

& 7 4. Control Building 280' elev. -not presently required -see notes 1,3,4,5,6

& 7 5. Auxiliary Building 328' elev. -not presently required -see notes 1,2,3,4,5,6 & 7 6. Auxiliary Building 305' elev. -not presently required -see notes 1,2,3,4,5.6

& 7 7. Auxiliary Building 305' elev. -not presently required -see notes 1,3,4,5,6

& 7 8. Auxiliary Buidling 280' elev. -not presently required -see notes 1,3,4,5,6

& 7 9. Control Building Area 282' elev. -not presently

-see notes 1.4,5 & 7

,--. ... port Section E.4 r.l(a) F.6 F.7 F.9

• Page 2 of 4 ATrACHMENT 2 (Cont'd) PAGE ( dif1catioD Halon Systems! address corrosion characteristics of Halon Containment!

hose reels in service during maintenance activities.

Remote Safety Related Panels! bose reel protection Station Battery Rooms! bose reel protection Diesel Cenerator Area! required sprinkler system from SER It sition 10. Diesel Ceuerator Bldg. West -Dot preseDtly required -see DOtes 1,3,4,5,7,8

& 9 11. Diesel Generator Bldg. East -Dot presently required -see Dotes 1,3,4,5,7,8

& 9 12. River Water Pump House -Dot presently required -see DOtes 1,3,4,5,7

& 8 13. River Water Pump House -not presently required -see notes 1,3,4,5,7

& 8 14. Control Building 305' elev. -not presently required -aee notes 1,5 & 7 -the safety related room on this elevation has a Halon suppression system installed.

For fire areas with installed Halon Systems the corrosion characteristics will be addressed.

The 280' elevation hose reels are not presently in service. When Reactor Building conditions permit these hose reels will be put in aervice. See E.3(c) See E.3(c) The SER required systems in the Dieael Cenerator Building Baaements, however, theae aprinkler systems are Dot required pres.ently for the reasons:

\ *: Report Section E.4 F.1(a) F.6 F.7 F.9

• 4400-82-L-OIU2 Page 2 of 4 I q tJA3E: S '1 r \0 I ATTACHMENT 2 (Cont'd) Halon Systems/ address corrosion characteristics of Halon Containmentl hose reels in service during maintenance activities.

Remote Safety Related Panels/ hos"e reel protection Station Battery Rooms/ hose reel protection Diesel Generator Areal required sprinkler system from SER D1s osition 10. Diesel Generator Bldg. West -not presently required -see notes 1.3.4,5,7,8

& 9 11. Diesel Generator Bldg. East -not presently required -see notes 1.3,4,5,7,8

& 9 12. River Water Pump House -not presently required -see notes 1.3,4,5,7

& 8 13. River Water Pump House -not presently required -see notes 1,3,4,5,7

& 8 14. Control Building 305' elev. -not presently required -see notes 1,5 & 7 -the safety related room on this elevation has a Halon suppression system installed.

For fire areas with installed Halon Systems the corrosion characteristics will be addressed.

The 280' elevation hose reels are not presently in service. When Reactor Building conditions permit these hose reels will be put in service. See E.3(c) See E.3ec) The SER required sprinkler systems in the Diesel Generator Building Basements, however, these sprinkler systems are not required presently for the reasons:

• .-Page 1 of 4 (g 1 ATTACHMENT 2 (Cont'd) Report Section D1s osition F.ll F.14 FA-026 FA-027 FA-05S FA-059 Safety Related Pumps/ hose reel protection Buildings/

1. Chemical Cleaning Building Modifications

2. Address fire related tion releases Diesel Generator Bldg. 12/ sprinkler Diesel Generator Bldg. 11/ sprinkler Chemical Cleaning Building/

detection Chemical Cleaning Air Filtration Room/detection

1. The Diesel Generator Bldg. 305' elevation has an installed deluge. 2. The-basements are separated by a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> rated wall. 3. Outside hydrants with stocked hose houses are available for additional fire suppression.

4. A detection system is installed.

5. Loss of one Diesel Generator does not affect the ability to maintain safe shutdown conditions.

See E.3(c) 1. Chemical Cleaning Building, Chemical Cleaning Air Filtration Room and TV Monitor Control Building will have a Detection System installed.

2. Radiation releases due to fire in Radwaste Buildings will be addressed.

See F.9 See F.9 See F.14 See F.14 *

-. ,. Report Sect:S:on FA-060 FA-045 FA-058 FA-060 .... uu-o.t.-.... -U.Lu.t. Paae 4 of 4 ATl'ACHMENT 2 PAGE 0 ( Of &, 1 THI-2 Fire Protection Prosram Evaluation Table of Modifications and Disposition of Modifications Subject/Modification TV Monitor Control Building/

detection Cable Room/ Alternate Shutdown Modifications Chemical Cleaning Building/

1. De tee tion 2. Radiation Releases Chemical Cleaning Air Filtration Room/ 1. Detection

2. Radiation Releases TV Monitor Control Building/

1. De tee tion 2. Radiation Releases Review of Ventilation Systems/ Radiation Releases Review of Drainage Systems/ Radiation Releases

• Disposition See F.14 Alternate Shutdown Modifications are not presently required since local control of the SPC system is available to maintain plant safe shutdown conditions.

1. See F.14 2. See F.14 1. See F.14 2. See F.14 1. See F.14 2. See F.14 See F.14 See F.14

,,-4000-82-L-Ol02 ATl'ACHMENT 2 NOTES 1. General area fire detection installed.

2. Nearby filter banks have deluge aystems. 3. Nearby electrical equipment is in aeparate 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> rated encloaures.

4. Loss of an individual fire area will not affect plant conditions.

5. Time is available for running additional hoses. fire extinguishers are immediately available.

6. The hose stations are back up only. 7. Fire Brigade preplans and strategies will reflect necessary additional equipment and techniques.

8. Outside hydrants with stocked hose houses available.

9. Deluge protection installed.

L** ......... -Metropolitan Edison Company Post Office Box 480 Middletown, Pennsylvania 17057 -.. -. TMI Program Office Attn: Dr. B. J. Snyder, U. S. Nuclear Regulatory Washington, D.C. 20555

Dear Sir:

Program Director Commission Write'" DiNet Dia. N.umber November 9, 1981-LL2-81-0260 Three Mile Island Nuclear Station, Unit 2 (TMI-2) Operating License No. DPR-73 Docket No. 50-320 10CFR50 Appendix R As a result of our request for exemption from 10CFR50 Appendix R, Section 111 G, J, and 0 completion time requirements on March 24, 1981 (LL2-8l-0084) you requested us to submit an updated fire hazards analysis by November 7, 1981. We are currently proceeding with an update to the TMI-2 Fire Hazards Analysis.

but have not progressed to the point where we are able to supply you a firm completion and/or submittal date. We will, however. inform you of our submittal date by December 7. 1981. Sincerely, lsI J. J. Barton J. J. Barton Acting Director, nll-2 JJB:SDC:djb cc: L. B. Barrett, Deputy Program Director Metropolitan EdIson Company IS a Member Of the General Pubhc S.'Stem AI'.

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON.

D. C. Z055S -S\ -tHo\ PAG£ (Q 1:, ,Qt to cr ..... it .r. .... , ... -Distribution MAY '1 1981 .. AI's 'i/OJf:! Subjec t:

No. 50-320 dU.tB U*!!¥lt"1f.1fI

.-Mr. Gale K. Hovey Vice President and Directcr of TMI-2 Metropolitan Edison Company P.O. Box 480 Middletown, Pennsylvania 17057

Dear Mr. Hovey:

The NRC Staff has reviewed your request of March 24. 1981 (LL2-81-0084) for exemption from the completion time requirements of 10 CFR Part 50.48(c) for the following items: . , '-(1) 10 CFR Part 50 Appendix R.Section III.G. -"Fire Protection of Safe Shutdown Capability" (2) 10 CFR Part 50 Appendix R.Section III.J. "Emergency Lighting" (3) 10 CFR Part 50 Appendix R.Section III.O. . "Oil Collection System for Reactor Coolant Pump" (4) Fire Protection Safety Evaluation Report -Open Items -"Fire Hose Stations Systems" and "Automatic Water Suppression in Diesel Room Basement" &fllft(.tI:

Assigned To: Due Date: '-O?-IJ Dist. Arnold-AD.BG.

Barton-AD.BG.

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Iv. Zlt:/!<<.JII t:A. 04t'WA-I P II. D"",;"'h/rrr Your request for exemption is based upon a fire hazards analysis which was formed prior to the TMI-2 accident of March 28. 1979. Since that fire hazards analysis does not include the facility modifications made following the accident, it is our position that prior to our taking any further action on this exemption request you should update your fire hazards analysis to reflect these facility modifications and then also update your exemption request accordingly.

The updated fire hazards analysis should be completed and submitted to us no later than six months from the date of this letter. i I would also note that the basis you provided for relief from the completion time I requirements of Section III.G. of Appendix R to 10 eFR Part 50 does not appear sufficient.

Specifically.

you state that Asince the accident in March 1979, the Unit 2 facility has progressed to its present cold shutdown condition which requires no active support for adequate core cooling. i.e ** maintenance of the /' cold shutdown condition." Although the reactor is being maintained in cold :shutdown without active core cooH.ng, certain actions including continued _. toring of various parameters (temperatures, pressures and neutron flux level) C , I I I I J J I J ./ I I I I i , I " -I A 4' . ";-,1. ' . ' " PAGE 5: Or: f..o; Mr. Gale K. Hovey -2:. are necessary to assure that the reactor is maintained in cold shutdown.

Paragraph l.b. of Section III.G. requires that in addition to achieving cold shutdown, fire protection features also be provided for systems necessary to maintain cold shutdown.

Therefore, it is our position that systems used for maintaining the reactor in a stable cold shutdowf.

condition (e.g ** monitoring instrumentation and systems for maintaining Reactor Coolant System pressure) should also be provided with fire protection features.

The availability of these fire protection features should also be addressed in your request for exemption from the completion time requirements of Section III.G. If you have any further questions regarding this matter you may contact Donald Brinkman of my staff at (301) 492-4857.

cc: See attached Sincerely.

--zSc:,..,., ** --J /.. ;4)-A -Bernard J. Snyd ,Program Director TMI Program Office Office of Nuclear Reactor Regulation

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r:P'a ijJ) L. Metropoli.n Edison Company Post Office Box 4BO Middletown, Pennsylvania 1'057 \. . Wrher'. Diract Dill Number Karch 24, 1981 LL2-8l-0084 Division of Operating Reactors Attn: Darrell G. Eisenhut, Director U. S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Sir:

Three Mile Island Nuclear Station, Unit Operating License No. DPR-73 Docket No. 50-320 10CFRSO Appendix R 2 (nlI-2) 1'MI-2 l>>1atribut1on

-a

• Arnold-AD.Be.

Arnold-PAR.

Barton-AD.Be.

Clark-PAR.

Devine-AD.Be.

El __ AD.Be. Fenti-TR.

259 , Fuller-AD .Be. This letter is submitted subject to 10CFRSO.48c6 concerning the recent1YHard1n&-TR.

6 implemented Appendix R Fire Protection Program requirements.

Specifi-Herhein-TR.ll cally, with regards to Section III G, J, and 0 and open SIR items, the licensee is requesting exemption from the completion time requirements Holawortb-EG&

of 10CFRSO.48c based on the following.

Hovey-AD.Be.

luklll-TR.I84 lOCFRSO Appendix R Section 111 G -"Fire Protection of Safe Shutdown Capability" This section requires that "Fire protection features shall be provided for structures, systems and components important to safe shutdown." The TMI Unit 2 facility is currently and will remain in a cold shutdown conditianfor some time to come. Since the accident in March 1979, the Unit 2 facility has progressed to its present cold shutdown condition which requires no active support for adequate core cooling, i.e. maintenance of the cold shutdown condition.

Passive heat removal via loss to ambient, if not used exclusively dUring recovery, will remain a viable option for core cooling until the eventual removal of the fuel from the core. For this reason, coupled with the fact that the facility will buana.-PAR.

J Xin&-AD.Be.

Kunder-AD.Be. .J Lacey-JCPU lfaD,anaro-PAR Scbuua.-PAR.

Thorpe-PAR.

Tipton-PAR. "allace-PAR.

J Webb-PAR.

J "11.on-ll.1.03 it "l1.on-PAR.

&!! Dot experience power levels during the recovery period auch that Section III G(l)a could apply (Bot Shutdown), the licensee . requests exemption from the requirement of the above referenced fire protection regulation for the remainder of the reco,very .ade.1IC._ This request is made as provided for in lOCFRSO.48c6 "required modification would not enhance fire .afety -SI-C ....

in the facility *** ". . Wt-tltl lOCFR50 Appendix R Section III J -Emergency Lighting DIS. tcril The capacity of the existing emergency lighting units installed in the Unit 2 facility is acceptable per the SER. In addition, Metropolitan Ed'son Company IS a Member of tl"o& Genera Put C Ul'I:a:

, Darrell G. Ei.enhut LL2-S1-00S4 the exi.tiDg units .atisfy the DeW Appendix R Section III J Tequirements.

Therefore, DO upgrading is Tequired in Unit 2 to .atisfy Appendix R Section 111 J except as may be Decessary as a result of Tesulting from Section 111 G .odifications performed .ometime after the recovery period. 10CFRSO Appendix R Section 111 0 "Oil Collection System for Reactor Coolant Pump" This section requires that "the reactor coolant pump .hall-be equipped with an oil collection system if the containment is not entered during normal operation." "Such collection .ystems shall be capable of collecting lube oil from all potential pressurized and unpressurized leakage sites in the reactor coo lent pump lube oil systems." As stated in regards to Section 111 G above, the Unit 2 Facility is in a cold shutdown mode requiring no active support, in this case reactor coolant pump (RCP) operation, for maintenance of proper decay heat removal. Furthermore, the RCP's cannot be operated while the ReS parameters are .0 low as to not meet the RCP's minimum NPSH requirements.

This condition will remain for the duration of the recovery mode. Therefore, the licensee requests exemption from the requirement of the above regulation for the remainder of the recovery mode. This request is made as provided for in lOCFRSO.4Sc6 where "required modification would not enhance safety in the facility.

• .". Fire Protection Safety Evaluation Report -Open Items Additional hose stations and sprinklers .ystems committed to in the SER section noted below are incomplete as of the present time. Section 9.0 Auxiliary System Paragraph 9.S Fire Protection Systems Subsections

-Fire Hose Statlons Systems and -Automatic.ater in Diesel Room .. sement" Owing to the fact that these open items originated from aafe concepts preceeding Appendix R Section 111 G, the licen.ee believes the aame reasoning expressed relevant to aection G above applies here also. Therefore, the licensee requests exemption from the time limit i;

.. t No. 50-320 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 JUl 2 (\ 1381 '. JUL 2'7 E'I \-0 f"l ' . I "-") ,4--TI'1l-2 Dl.trlbuUoo AI'.

Aadgned To: Due Date: Mr. Gale K. Hovey Vice President and .rAN: iQa Director of TMI-2 Metropolitan Edison Company P.O. Box 480 Dist. Middletown, Pennsylvania 17057

Dear Mr. Hovey:

The Nuclear Regulatory Commission has issued an exemption to the requirements of 10 CFR Part 50.71(e) to License No. DPR-73. This exemption deletes the requirements to periodically update the TMI-2 final safety analysis report (FSAR) to reflect facility changes made during the cleanup of TMI-2 and is in response to your request of May *6, 1981 (LL2-81-0114).

We have concluded your proposal to use Technical Evaluation Reports (TERs) for documenting e changes and associated safety evaluations is an acceptable alternative Jpdating the FSAR. provided the TERs are kept updated. Additionally, we will require that the System Descriptions for the major post-accident recovery systerrs(e.g

• • EPICOR-II, Mini Decay Heat Removal, Standby Pressure Control, Long Term B Cooling, Tank Farm, Solid Waste Staging Facility, etc.) are kept updated since there are no TERs for these systems. Therefore, as a condition of this exemption, we will require that at least once per six months, you review the TERs which have been issued and the System Descriptions for the major post-accident recovery systems and make any necessary updating revisions.

For those System Descriptions which have not been docketed, we will require their submission on the docket within six months of this letter. Furthermore, any changes to the facility described in the TERs and System Descriptions, Amold-AD.IC.

kUard-TR.175 krton-AD.aG.

araaher-AD.IC.

Clark-PAR.

Deltttte-AD.IC.

DeVine-AD.

loG. fuller-AD.IC

• Cischel-AD.IoC.

Herbe1n-TR.1l8 Hobvorth-EG"C liovey-AD.BC.

Hukill-TIL 184 Kanuna .. -PAR. Ung-AD.BG.

Xunaer-AD.aC.

Lacey-JCPl>.l.

Hanganaro-PAR.

Schmauu-PAR.

Skl1lun-AD.IC.

SIIyth-TR.68 Thorpe-PAR.

Wallace-PAR.

Walah-PAR.

J Wl1aoo-1.l03 I Wl1aon-PAR.

DOCC -nu DOCC -PAR. changes in the procedures described in the TERs and System Descriptions, and conducting of tests or experiments not described in the TERs and System Descriptions shall be subject to the requirements of 10 CFR Section 50.59. We have determined that the granting of this exemption involves an action which is insignificant from the standpoint of environmental impact and that there is reasonable assurance that the health and safety of the public will not be dangered by this action. Having made this determination, we have further cluded that pursuant to 10 CFR 151.5 (d) (4) an environmental impact appraisal need not be prepared in connection with the granting of this exemption.

C I ./ A / I ./ J ./ .; I .; '" " ./ J * * * , , I

( ( ( rnuL.. vr, I I Mr. Gale K. Hovey Copies of the related Safety Evaluation and the Notice of I.ssuance.

which has been forwarded to the Office of the ,Federal Register for publication, are also enclosed.

Enclosures:

1. Safety Evaluation

2. Notice of Issuance cc w/enclosures:

See attached Sincerely.

i I '; (. . -\'-u* ... j

.'

Program Di rector TM! Program Office Office of Nuclear Reactor Regulation

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" ( f SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION METROPOLITAN EDISON COMPANY Introduction JERSEY CENTRAL POWER AND LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY DOCKET NO. 50-320 THREE MILE ISLAND NUCLEAR STATION, UNIT NO, 2 By letter dated May 6, 1981 (reference I), the Metropolitan Edison Company (licensee) requested an exemption from the requirements of 10 CFR Part 50.71(e) to periodically update the TMI-2 final safety analysis report (FSAR). The exemption would be for the duration of the TMI-2 cleanup. In lieu of periodically updating the FSAR, the licensee has committed to submit a System Description (SO) and a Technical Evaluation Report (TER) for each major step of the cleanup. Evaluation The purpose of the requirements contained in 10 CFR Part 50.7l(e) is to provide an updated reference document to be used in recurring safety analyses.

As a result of the March 28, 1979, accident at TMI-2, power operation is no longer possible with TMI-2 in its present status but :ather cleanup operations not anticipated in the design of the TMI facility nor described and analyzed in the TMI-2 FSAR must now be performed.

The facility modifications and panying safety analyses for the operations are unique to the cleanup operations and such facility modifications would probably have to be removed prior to restoring TMI-2 to operation if such a decision is made at some future date. Therefore, the licensee has proposed that rather than modifying and dating the FSAR to describe and analyze the facility modifications associated . with these cleanup operations, SDs and TERs be prepared and submitted to the NRC for each major step of the cleanup.

.' ( The SOs and TERs will include system descriptions and safety evaluations

'. of the planned cleanup actions and will therefore provide the necessary information to describe and assess the cleanup Operations as well as providing a record of the facility modifications necessary to perform the cleanup. Since the SOs and TERs will provide the same type of information that would be added to an updated FSAR and since the SOs and TERs will provide this information for the entire cleanup operation, we have concluded that the SOs and TERs will be an acceptable alternative to the requirements of 10 CFR Section 50.71(e) provided they are kept updated. To ensure that these documents are kept updated, we will require as a condition of granting this exemption that the licensee review them at least once per six months and make any necessary updating revisions. ( This will require updates more frequently than the annual updates required by Section 50.71 (e){4) for FSARs. This augmented requirement is necessary because of the rapid pace at which some of the cleanup activities may be conducted.

Furthermore, if a subsequent decision is made to restore TMI-2 to operation, the FSAR will then require updating in accordance with the requirements of 10 CFR Part 50.71(e).

Any changes in the facility described in the SOs and TERs, changes in the procedures describea in the SOs and TERs, and conduct of tests or experiments not described in the SOs and TERs shall be subject to the requirements of 10 CFR Section 50.59. Public Interest Considerations Under 10 CFR Section 50.12{a).

the Conrnission may grant exemptions from the requirements set forth in Part 50 if it determines that the exemptions are "authorized by law and will not endanger life or property or the.conrnon defense and security and are otherwise in the public interest".

As analyzed above,

( the SOs and TERs are a more appropriate vehicle than the FSAR for achieving the updating requirement of 50.71(e) during the cleanup period. It is consistent, therefore, with the purpose of Section 50.71(e) to allow this requested exemption.

See Statement of Consideration, "Periodic Updating of Final Safety Analysis Reports", 45 F.R. 30614, May 9, 1980. The exemption is authorized by law since it is consistent with the purpose of Section 50.71(e) and will not endanger life or property or the common defense and security since it does not relax any Commission requirement.

Significant Hazards Considerations The granting of this exemption does not entail any significant hazards considerations since it merely permits an alternative form for the filing of required documentation with the Commission.

The time interval for update of this information will be more frequent than required under the Commission's regulation.

The granting of the exemption does not involve any increase in the probability or consequences of accidents previously evaluated nor the creation of the possibility of a different type of accident.

nor does it reduce the margin of safety defined in the basis of any license requirements.

Conclusions Based on the foregoing.

we have determined that. pursuant to Section 50.12 of 10 CFR Part 50, a specific exemption for the duration of the cleanup operations as discussed above is authorized by law and can be granted without endangering life or property or the common defense and security and is otherwise in the public interest.

IT Furthermore, we have determined that the granting of this exemption does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. We have concluded that this exemption would be insignificant from the standpOint of environmental impact and pursuant to Paragraph (d)(4) of Section 51.5 of 10 CFR Part 51 that an environmental impact statement, or negative declaration and environmental impact appraisal, need not be prepared in connection with this action.

" ' ( ( References

1. Letter to B. J. Snyder. NRC, from G. K. Hovey. Met Ed/GPO. dated May 6. 1981. (LL2-81-0114).

.. ( 7590-01 FArr 9 If UNITED STATES NUCLEAR REGULATORY COMMISSION DOCKET NO. 50-320 METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER AND LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY '. GRANTING OF RELIEF FROM REQUIREMENTS FOR UPDATING FINAL SAFETY ANALYSIS REPORT The U.S. Nuclear Regulatory Commission (the Commission) has granted an exemption from certain requirements of 10 CFR Part 50.7l(e) to Metropolitan Edison Company, Jersey Central Power and Light Company, and Pennsylvania Electric Company. The exemption relates to the requirement for periodically updating the Final Safety Analysis Report (FSAR) for Three Mile Island Nuclear Station, Unit 2, located in Dauphin County, Pennsylvania.

The exemption is effective as of its date of issuance.

The exemption deletes the requirement to periodically update the TMI-2 FSAR to reflect facility changes made during the cleanup of TMI-2 and provides for the use of System Descriptions (SOs) and Technical Evaluation Reports (TERs) for documenting these changes and associated safety evaluations.

The exemption also requires that any changes to the facility described in the SDs and TERs, changes to the procedures described in the SDs and TERs, and conduct of tests or experiments not described in the SDs and TERs shall be subject to the provisions of 10 CFR Section 50.59. The reQuest for relief complies with the standards and requirements of tile Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations.

The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the letter granting relief.

-' . ( '-. --( PAGl: (0, II 7590-01 .. The Commission has determined that the granting of this relief will not result in any significant environmental impact and that pursuant to 10 CFR 151.5(d)(4) an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with this action. For further details with respect to this action, see (1) the request for relief dated May 6, 1981, (2) the Commission's letter to the licensee dated July 20, 1981, and (3) the Corrrnission's related Safety Evaluation.

These items are available for public inspection at the Commission's Public Document Room, 1717 H Street, N.W ** Washington, D.C. 20555 and at the Government Publications Section, State Library of Pennsylvania, Education Building, Corrrnonwealth and Walnut Streets, Harrisburg, Pennsylvania 17126. A copy of items (2) and (3) may be obtained upon request addressed to the U.S. Nuclear Regulatory Corrrnission, Washington.

D.C. 20555, Attention:

Director.

1MI Program Office. Dated at Bethesda, Maryland this 20th day of July, 1981 FOR THE NUCLEAR REGULATORY COMMISSION I

.. I I -'!. * ,. ... .. ""'-_ .-:.'-* J. :' . '0 Bernard J.

Program Di rector Three Mile Island Program Office Office of Nuclear Reactor Regulation I rnl :{!1711j '-, I { I I Metropolitan Edison Company Post Office Box 480 Middletown, Pennsylvania 17057 Writer', Direct Dial Number TKI-2 Diatr1bution

!AI's -* , May 6, LL2-Sl-0ll4 c: .........

.&.\0..,.., TMI Program Office Attn: Dr. B. J. Snyder, U. S. Nuclear Regulatory Washington, D.C. 20555

Dear Sir:

Director Commission

.--Three Mile Island Nuclear Station, Unit 2 (TMI-2) Operating License No. DPR-73 Docket No. 50-320 10CFRSO.7l(e)

Exemption Request This letter is written to request an exemption from the ments of 10CFRSO.7l(e) to periodically update the final safety analysis report (FSAR) originally submitted as part of the application for the operating license. The stated purpose of this requirement is to provide an updated reference document to be used in recurring safety analyses.

The FSAR was written to provide a basis to allow TMI-2 to operate at power. Operation at power is no longer possible with TMI-2 in its present state. Additionally, the TMI-2 FSAR, although a valuable reference document, does not completely scope the recover effort. Each major step of the recovery, such as reactor building sump water processing, will be presented in a Technical Evaluation Report so that a safety evaluation of the entire recovery effort will eventually be developed, therefore, updating the FSAR itself is not necessary.

Additionally, the NRC's Office of Nuclear ." Reactor Regulation (NRR) has established a separate progtam office for TMI and maintains a permanent onsite staff. Through other transmittals, the staff is kept continuously aware of conditions at TMI, thus is little need' to make a separate transm1ttal to the of an updated FSAR to keep it informed of our status. -. Metropolitan Edison Company is a Member of the General Public Ulihtles System Rr .. ne.""\-[Ass1ghed To: I Due Date: -; D1st. C A Arnold-AD.BG.

.. Barton-AD.BG.

-Clark-PAR.

DeV1ne-AD.BG.

Ela_AD.BG.

Fenti-TR.259 Fuller-AD.BG.

Harding-TR.68 Herbein-TR.118 Hevard-PAR.

Hockley-HEAR.

Holzvorth-EG&G Hovey-AD.BG.

Hukill-TR

.184 Kanzanas-PAR.

King-AD. BG. , Kunder-AD.BG.

Lacey-JCP&L Manganaro-PAR.

Schmauss-PAR.

Thorpe-PAR.

Tipton-PAR.

Wallace-PAR.

Walsh-PAR.

J Wilson-RlO3 R Wilson-PAR.

DDCC-TMI v' DDCC-PAR.

t/ ---liEVIEWS --i.F .<:t. I ;.. , OPS. d?, LIC. J * /"j/f A. . i ..x.r'-jN'(' ,

( Dr. B. J. Snyder LL2-S1-0114 Therefore, since we are developing additional safety evaluations to proceed with the recovery effort, which are reviewed and approved by TMI Office, we believe there is little purpose in updating the present TMI-2 FSAR and request exemption from the requirements lOCFR, Part 50.7l(e).

In the event TMI-2 is restored to an operable cqpdition, that documentation will be required to FSAR format to address system changes. GKH:JJB:djb cc: L. H. Barrett, Deputy Program Director ."" Sincerely, I SIC;: K. . . -"-' ._-G. K. Hovey Vice-President and Director, TMI-2 * . P

.... g.a.. ...__.__ r-p .. * '.0' ** "(' -t . ......;..,....

..... . ........ *.,****x* . PA.:;r OF / T' "Suelear III ( .pu Nucl ** r P.o. Box 4S0 Middletown, Pennsylvania 17057 717-944-7621 . WrHer's Direct Dial Number:

tKl-I January 20, 1982 4400-82-L-0009

'-cr Program Office ; tn: Dr. B. J. Snyder, Progru' Director S. Nuclear Regulatory ,shington, D.C. 20555 :ar Sir: Recovery Three Mile Island Nuclear Station, Unit 2 (TMI-2) Operating License No. DPR-73 Docket No. 50-320 . System Description and Technical Evaluation Report Update ur letter of July 20, 1981 requested that we provide periodic updates of chnical Evaluation Reports (TER's) and System Descriptions for major cident recovery systems. Enclosed are updated System Descriptions for the Reactor Coolant Pressure Control (SPC) System and the Fuel Pool Waste 0l >ystem (Tank Farm) and the Technical Evaluation Report for the Interim' 1 .. aste Staging Facility.

discussed with Dr. R. R. Bellamy, of your staff, System Descriptions for e Mini Decay Heat Removal (MOHR) System, the Auxiliary Building Emergency quid Clean-up System (EPlCOR-II), the Submerged Demineralizer System (SDS), the Interim Solid Waste Staging Modules are undergoing internal review approval and will be provided to you as they become available, with all IUUiMdIa ilt. . Ar'Dolcl-AD.IC.

."1 JallarcS-n.175 . lartora-AD.IC

• lraaber-AD.IC. .llChaUD-AD.IC.

DeViDe-AD.IC.

rWrock-PAR.

G1acbel-AD.IC.

Berba1D-Tl..1l8 Bu1r.1l1-n.184 UDI-AD.IC.

luDclar-AD.IC.

Lacey-JCPiL LaraoD-AD.1C

• MaqUaro-PAR.

Paator-AD.IC.

ScbMuaa-PAl.

S1r.11l:uD-AD.IC.

Sllytboon.6S Wallace-PAR.

Wabb-PAl.

J WUaora-U03 DDCC -'IKI DCC the System Descriptions listed above provided by March I, 1982. Ly, the Technical Evaluation Report for the SDS i, being updated to include I lctor Coolant System Processing and will be provided to you by April I, 1982.: . I litionally, the requirement of updating these System Descriptions and :hnical Evaluation Reports every six(6) months as atated in your letter of Ly 20, 1981 is inconsistent with the ooe(l) year frequency for PSAR updates luired by 10CFRSO.71.

Therefore, we request that the frequency for the item Descriptions and Technical Evaluation Report updates be reduced to once I year *. If you have any further questions, please contact Mr. J. E. Larson, \ my staff. Sincerely, :achment L. B. Barrett, Deputy Program Director lsi J *. J. Barton J. J. Barton Acting Director; tMI-2 GPU Nuclear is a part of the General Public Utilities System I o **

v-, r a.a. : .... J \ . -

I UNITED STATES NUCLEAR REGULATORY COMMISSION WASHING10N.

o. C. 20555 Mr. J. J. Barton Acting Director, TMI-2 GPU Nuclear Corporation P.O. Box 480 Middletown, PA 17057

Dear Mr. Barton:

February 4, 1982 This letter is in response to your letter (4400-82-l-0009) of January 20, 1982 regarding the updating of recovery system descriptions (SO's) and technical evaluation reports (TER's). In your letter you request that the frequency for updating SO's and TER's be extended from 6 months to 1 year, noting that the frequency for required (i.e., by 10 CFR 50.71) updating of Final Safety Analysis Reports (FSAR's) by other licensees is 1 year. We note that your TER's and SO's are your method for documenting changes to /" the facility and providing associated safety evaluations and constitute an acceptable alternative to updating the FSAR. We agree that updating the SO's and TER's at a 1 year frequency is a more reasonable time period and would be consistent with the requirements of 10 CFR 50.71. We further note that we maintain a staff, both onsite and in headquarters, who keep abreast of facility changes and review the associated engineering and safety documentation (engineering change memos and safety evaluations required by 10 eFR 50.59), as necessary.

For the above reasons, we approve your request to submit updated SO's and TER's at a 1 year frequency.

cc: l. Barrett O. lynch R. Weller Sincerely, ll/d6-. Bernard J. Snyder, Program Director TMI Program Office Office of Nuclear Reactor Regulation

_ ... III ... \

¢.;2,8.1 A. N. SHINPOCH ...,.. CO E' f.a I Secretary ll8 s: ; .-c1O: R.l?1+I , STATE OF WASHINGTON U III'IEI ,..It'4. DEPARTMENT OF SOCIAL AND HEALTH SERVICES r Olympia, Washington 98504-()()95

""Pal ant roc' ltd .. or_ December 11, 1985 ...... * "tntr r , . " , t/)"\ \ 1QIn, t.ft ; two:r If MlR5* L* COI,( --;< .. i Russ Wells Licensing Manager GPU Nuclear Corporation P.O. Box 480 ! ...... DEC 198r i, : , .,'J RECEIVED; I

Route 441 South Middletown, Pennsylvania

Dear Mr. Wells:

K,M,I!I:I, "ill'" . ? IIWW

,-. ." \:",{' , .. -, -... , .. --..0.-. This letter is in response to the request received from E. K. Kanga, Director, TM! II, for a variance from the provisions of the license issued by the state of Washington to US Ecology. As I understand your request the variance was to allow the use of a high integrity container to bury Class Band C wastes. The high integrity container would also be used as a processing vessel to dewater resin materials.

It is my understanding that the Topical Report process and the Staff Evaluation Report dated October 1985, issued by the Nuclear Regulatory Commission briefly considered the use of the Ferrallium 50 cubic foot container with an internal dewatering mechanism.

Although the exact configurations of the dewatering equipment may differ slightly from that proposed by TMI, Nuclear Packaging indicated that the use of internal dewatering mechanisms did not represent a threat to the structural integrity, nor to the corrosion resistance of the material.

The internals would be sacrificial to the container and the rigid container design would preclude puncturing by the internal fixtures.

The completed review by the Nuclear Regulatory Commission did not take exception to the manufacturer's claim. Because of this tation it does not appear that a variance is necessary to allow EA/FL-50 to be used for dewatering power plant resin materials.

The container, however, must be used in conjunction with a documented process control program which verifies that the free standing liquid requirements, and all the provisions of the current quarterly authorization letter or Certificate of Compliance, when issued, are met. Also, the conditions of use referenced in the Staff Evaluation Report (copy enclosed) must be met.

me directly if you have further questions concerning this issue. We anticipate our Certificate of Compliance will be issued shortly for the EA/FL-SO container.

Sincerely, NPK:pm c,'c c:

• I ....... ...-0::

NOV 2 5 1985' Mrs. Nafley P. Kirner. Supervisor Waste Management Unit Department of Social and Health Services Heil Stop IF-13 Olympia. Washington 98504

Dear Mrs. Kirner:

Ref: SA/K!lS Enclosed is a copy of the Staff Evaluation Report related to thp Topical Report covering the Fl-50/EA-50 High Integrity Container manufacturEd by Packaging Inc. If you have any questions please contact Mrs. Schneider at 301-492-9893.

Enclosure:

As stated Distribution:

SA R/F Di r RIF KNSchneider Topical Report file (fc) w/encl lHi 99; nbotham JGreeves Sincerely.

Or1s1.Dal 11.-4 ",. >>.Ju.aa .....

• Donald A. Nussbaumer Assistant Director for State Program Office of State Programs _

_ Identical letter sent to: H. Shealy, SC

...... !d.]I**.***

.... dl I I 111.221 85 1 .1l/rI8,S

,! . -t .. '"' .", ", .. , ...........

,..1.. ------.. : ,:: ',' 31B (lC'BOI "'Ret.' C240 OFFICIAL RECORD COpy WM45/LBH/85/10/17 Docket No. WH-45 Mr. Richard T. Haelsig, Nuclear Packaging, Inc. 1010 South 336th Street Federal Way, Washington

Dear Mr. H3elsig:

President 98003 "ov .... . . ... DISTRIBUTION NMSS rlt WM s/f: WM-4S WMLU rlf MTokar TCJohnson JTGreeves DEMartin RJStanner MNataraja WGalllTlill.

NRR AGrella, IE FCostanzi, RES CMacDona1d, NMSS DNussbaumer, SP

SUBJECT:

NUPAC TOPICAL REPORT ON FL-50/EA-50 HIGH INTEGRITY CONTAINER The Nuclear Regulatory Commission (NRC) has completed its review of the Nuclear Packaging, Inc. (NUPAC) topical report on the Fl-50/EA-50 High Integrity Container (HIC) for low-level radioactive waste. The technical review included information contained in the draft topical report as well as further information that was submitted as a result of the review. The evaluation report for this review is enclosed.

We have concluded that the topical report, as supplemented by additional information that was provided in response to staff comments and questions, adequately describes the FL-SO/EA-50 HIC and that, as described, the HIC meets the structural stability requirements of 10 CFR 61 for the disposal of Class B and Class C wastes. These conclusions are predicated on completion of the final revised topical report (proprietary and non-proprietary versions) to include all appropriate information that was developed during the course of the technical review and the following conditions:

1. The FL-50/EA-50 HIC shall be used in accordance with the Operating Procedure restrictions outlined in the Appendix to this TER and all additional restrictions and requirements specified by the burial site operators and governing State agencies.

2. Users of the Fl-SO/EA-SO HIC shall certify that all restrictions and required procedures have been adhered to and that the HICs do not contain proscribed chemicals or waste materials.

It is our understanding that NuPac will retain and provide upon request appropriate specimens of container construction material for use in possible future surveillance programs.

For example, these specimens could be used as corrosion samples buried in an "archival trench" at a llW burial site and retrieved and inspected at periodic intervals.

The enclosed evaluation report is being forwarded to the States of South Carolina and Washington for their information and use. C .WMUJ :WMEG :WMEG :WHEG .. _-:------------:------------:------------:------------:-----------

TCJohnson

JTGreeves

____ .e. __________

.-___________

-. __________

._. __________

.-. __________

._. ______________________

_ . . . . . . . DATE : 84/10/17 :851101 :85/101 WM45/LBH/85/10/17 MOV .-. * **** If NRC criteria or regulations change such that the acceptability of the topical report is invalidated, NuPac or the applicants referencing the topical report will be expected to revise and resubmit their respective documentation or otherwise justify the continued use of the topical report without revised documentation.

Enclosure:

Evaluation Report for NuPac HIe Sincerely, Ori,ra.l .... ., Leo B. ILa:i nbcH.b..am leo B. Higginbotham, Chief low-Level Waste and Uranium Recovery Projects Branch Division of Waste Management Office of Nuclear Material Safety and Safeguards United States Nuclear Regulatory Commission Office of Nuclear Material Safety and Safeguards Washington.

D.C. 20555 ---===============================================

STAFF EVALUATION REPORT related to the Topical Report covering the FL-SO/EA-SO High Integrity Container .anufactured by Nuclear Packaging.

Inc. Docket No. WM-45 Prepared by: Engineering Branch Division of Waste Management October 1985 ABSTRACT This Staff Evaluation Report has been prepared by the Office of Nuclear Material Safety and Safeguards of the U. S. Nuclear Regulatory Comission for the Topical Report filed by Nuclear Packaging, Inc. covering its FL-SO/EA-SO High Integrity Container.

The container is proposed for use as a lDeans of containing low-level radioactive waste and .eeting the structural stability requirell1ents for waste in 10 CFR Part 61. The staff concludes that the Fl-SO/EA-SO high integrity container

_eets the structural stability of Part 61 and may be used for the disposal of low-level radioactive waste that requires disposal in a stable form. limiting conditions for use of the container aay be specified by the regulating authority for a particular disposal site. i i i TABLE OF CONTENTS Page 1.0 BACKGROUND

.................................................

1 1.1 Regulations

.............*.............................

1 1.2 Topical Report Submittals

.............................

1 1.3 Fl-SO/EA-50 HIC Description

.........................

,. 2 2.0

SUMMARY

OF TOPICAL REPORT ................................

,. 3 3.0

SUMMARY

OF REGULATORY EVALUATION

...........................

4 3.1 Major Artas of

.................................

4 3.2 Corrosion

.............................................

4 3.2.1 Background

...*......................*..........

4 3.2.2 Corrosion-Related Metallurgical Factors ........ 5 3.2.2.1 Corrosion Performance of F255 Welds *.............*.......*..

5 3.2.2.2 Pitting Corrosion Repassivation

..........................

6 3.2.2.3 Field Experience with Comparative Alloys .....................

7 3.2.2.4 Crevice Corrosion

....*..................

B 3.2.2.5 Effects of localized Corrosion on Structural Integrity

................

B 3.2.3 Environmentally-Related Corrosion Factors ...... 9 3.2.3.1 General .................................

9 3.2.3.2 Review Areas Concerning Environmentally-Related Corrosion Factors ...............................

10 3.3 Structural Analyses ..................................

11 3.3.1 Burial Loads ..................................

12 3.3.2 Drop Test Load Analyses .......................

13 3.3.3 Thermal Stresses ..............................

13 3.4 Prototype Testing ....................................

13 ii i WM-45 STAFF EVALUATION REPORT 3.4.1 Drop Tests ....................................

14 3.4.2 Type A Package Criteria .......................

14 3.5 Gas Generation and Internal Pressurization

.......*...

16 3.6 Radiation and Ultra-Violet Stability

.................

17 3.7 Quality Assurance and Inspection

....................*

18 3.8 Miscellaneous Requirements

.*............*............

19 3.8.1 Free liquid ...................................

19 3.8.2 Creep .........................................

19 3.8.3 Biodegradation

... '" ..........................

19 3.8.4 Top Surface Water Retention

...................

21 3.8.5 Cold Weather Testing ..........................

21 3.9 Surveillance

.........................................

22 4.0 REGULATORY POSITION .......................................

22

5.0 REFERENCES

................................................

24 6.0 APPENDIX (Operating Procedure)

........................... . iv WM-45 STAFF EVALUATION REPORT 1.0 BACKGROUND 1.1 Regulations By Federal Register Notice dated December 27. 1982 (47 FR 57Q6). the United States Nuclear Regulatory Commission amended its regulations to provide specific requirements for licensing of facilities for the land disposal of low-level radioactive waste. The .ajority of these requirements are now contained in Part 61 to Title 10 of the Code of Federal Regulations (10 CFR 61) entitled "Licensing Requirements for land Disposal of Radioactive Waste" (Ref. 1). Minor lIodifications.

lIostly of a procedural nature, have been lIade to other parts of the Commission's regulations, such a5 10 CFR 20 ("Standards for Protection Against Radiation").

These regulations are the of a set of prescribed procedures for low-level radioactive waste disposal that were proposed in the Federal Register on July 24, 1981. The effective date for the illplementation of 10 CFR 20.311. which requires waste generators to Meet the waste classification and waste form requirements in 10 CFR 61, was December 27. 1983. As set forth in 10 CFR 61.55, Class Band Class C waste Must .eet structural stability requirements that are established under 10 eFR 61.56(b).

In May 1983, the NRC provided additional guidance by lIeans of a Technical Position on Waste FoT'll {Ref. 2} that indicated that structural stability could be provided by processing (i.e., solidification of) the waste form itself (as with large activated steel components) or by emplacing the waste in a container or structure that provides stability (that is, a high integrity container (HIC>>. 1.2 Topical Report Submittals By letter. dated November 3. 1983 (R,f. 3) Nuclear Packaging, Inc. (NuPac) requested consideration by the State of Washington for approval of a Ferralium 255 (F255) liner System (the NuPac Fl-SQl high integrity container) for use in the disposal of Class Band C filters from Arkansas Nuclear One to Hanford, Washington at the U.S. Ecology low-level radioactive waste disposal site. At the time, Arkansas Power and light (AP&L) was contracting with NuPac for the supply of carbon steel liners for packaging these filters for burial at Hanford. With the i.inent hlplelllentation (on December 27. 1983) of the requirements for HICs as specified in 10 CFR 61, as well as site specific requirements dictated by the State of Washington, NuPac requested an early review of the request for approval of their Fl-50/EA-SO HIe, as described in the topical report. The State of Washington, in turn, requested assistance (Ref. 4) in the review 1 During the course of this technical review, HuPac renamed the Fl-SO HIe as the Enviral loy SO (EA-SO) HIC. From this point on in this Topical Report Evaluation the HIC is referred to as the Fl-50/EA-SO HIC. 1 WM-45 STAFF EVALUATION REPORT of the topical report through NRC's Office of State Programs.

A preliminary technical review, involving primarily .embers of (a) the Engineering Section of NRC's Waste Managelllent Engineering Branch, Division of Waste Management. (b) Brookhaven National laboratory, (c) the Waste Technology Section of NRC's Waste Management Branch, Office of Research.

and (d) the Transportation and Certification Branch of NRC's Division of Fuel C."cle and Material Sa1.:ty. resulted in the generation of several cOINIIents (Ref. 5) on the AP&L related FL-50/EA-50 report. These comments focussed principally on the need for further inforsation on the corrosion behavior of the Ferralium 255 alloy. because corrosion was believed to be a controlling factor in the performance of a HIC. At about the SUM! tiae that the corrosion coments were being transmitted to the State of Washington for consideration, NuPac submitted (Refs. 6 and 7) a second topical report on the FL-50/EA-SO HIC. Whereas the first report had dealt with a specific application of the HIC for AP&l filter cartridge waste to be sent to Hanford, the second topical was intended to be generic, to apply to a broad spectrum of waste streams, and to allow for disposal at Barnwell, South Carolina as well as Hanford, Washington.

InasmUCh as the generic report encompassed and bounded the information contained within the AP&l-related document, the review effort was consolidated, and further review activity focussed on the generic topical. A request for further infoMllat ion (Ref. 8) that incorporated relevant infoMllation on soil analyses by an NRC contractor (Ref. 9) and which consolidated questions on the generic report was transmitted to NuPac in October 1984. 1.3 FL-50/EA-50 HIC Description The NuPac Fl-SO/EA-SO high integrity container is a simple right angle cylinder with a flat top and bottom entirely of Ferralium 255. The HIC is approximately 47 inches in diameter by S1 inches tall. The top, bottom, and sides of the container are fabricated from 3/8 inch thick material.

The top head has a 24 inch diameter gasketed opening for loading. Closure of this opening is accomplished with a 3/8 inch Ferralium Alloy 255 plate held in place by eight wedge shaped retainer blocks. Four internal l-shaped vertical supports, welded to the inside surfaces of the top and bottom plates, are provided as stiffeners for the top and bottom plates. A seal is provided between the lid and top of the HIC by a silicone rubber gasket (an optional lead gasket is available for highly permeable wastes such as tritium gas). A vent system is located in the lid and allows relief of internal pressure that could result from gas generation caused by biodegradation or radiolytic decay, while preventing significant groundwater movement into or out of the container.

The vented lid is not to be used with wastes that contain highly mobile or transient gases such as tritium. Lifting of the container is accomplished using a cable sling that is provided.

The sling consists of a single 3/8 inch steel cable that is attached to two lifting eyes on the container with anchor shackles.

2 WM-45 STAFF EVALUATION REPORT 2.0

SUMMARY

OF TOPICAL REPORT The generic topical report on the NuPac FL-SO/EA-SO high integrity container is intended to demonstrate that the HIC lIeets (a) 111 the applicable stabil ity requirements Ind criteria of 10 CFR 61 (using guidance provided in the May 1983 Technical Position on Waste Form). (b) 10 CfR 71 sections dea,ing with Type A Packaging (as the Part 71 requirements Ipply to HICs). (c) 49 CFR 173 Type A Packaging related Ireas, and (d) special testing Ind design conditions requested by the Agreement States. rL-SO/*A-SO HIC designed to be certified as I DOT Type A container that would pass all U.S. DOT and U.S. NRC transportation requirements for a Type A container.

The HIC is intended to contain the following types of wastes from light water reactors:

(1) dewatered bead resins, powdered resins and diatomaceous earth; (2) compressible solid waste; (3) non-compressible solid waste; (4) filter elements Ind cartridges; (S) solidified resins. sludges, and liquid wastes. The .aterial from which the FL-SO/EA-SO HIC is fabricated is Ferral ium 255 (F2SS), which is I patented ferritic-austentic, duplex stainless steel that reputedly combines high lIechanical strength.

hardness and ductility with excellent corrosion properties.

As acknowledged in the report, lithe lIost critical Irea associated with long isolation is considered to be corrosion resistence." A lIajor portion of the report therefore.

addresses.

the predicted external corrosion behavior of the F2SS HIC under expected disposal site environments and In Inalysis of the interna' corrosion of the HIC. taking dewatered bead resin IS the expected worst case. The rest of the report. IS submitted, focussed on structural analyses (including results of finite-element

.. calculations using the ANSYS computer code). Ina lyses of closures Ind seals. Inalyses of internal gas generation and Issociated gasketing requirements.

Ina lyses of radiation Ind ultra-violet stability.

prototype testing. Type A package testing, heat transfer.

inspection.

Ind quality Issurance.

Much of the information Iddressing these subjects is contained in several Ippendices.

The final Ipproved report will contain this technical evaluation Ilong with additional information submitted in response to NRC review cOlllllents and questions.

The additional information will be included in the revised report IS a second volume. 3.0

SUMMARY

OF REGULATORY EVALUATION 3.1 Major Areas of Review The basic objective of this staff technical evaluation of the topical report was to confirm that the NuPac FL-SO/EA-SO HIe lIeets the structural stability requirements of 10 eFR 61. The NRC's TeChnical Position on Waste Form (May 1983). which Iddresses vlrious details including certain transportation and testing requirements that Ire presented in 10 eFR 71 Ind 49 eFR 173. provides guidance on how to satisfy Part 61. Major Ireas of review that are Iddressed 3

WM-4S STAFF EVALUATION REPORT in the Technical Position and which received particular attention in this review included the following:

3.2 3.2.1 1. Corrosion

2. Structural Analyses 3. Prototype Testing 4. Gas Generation and Internal Pressurization S. Radiation and Ultra-violet Stability

6. Type A Packaging Requirements

7. Quality Assurance and Inspection

8. Remaining Technical Position and Other Considerations Corrosion Background Because of its reputed high resistance to stress corrosion cracking.

crevice corrosion.

Ind chloride-induced pitting. when cocpared with austenitic 5tainle55 steels such IS Types 304 and 316. Ferralillll 2SS is used in urine applications, the oil and gas (and petrochemical) industries, for pollution control equiplllent, and other applications where the cOilbination of corrosion resistance and high strength are especially needed. There is little field experience, however, with F2SS in long-term underground applications.

Nor is there .uch information available in the open literature regarding the corrosion of F2SS weldaents and the potential for long-range pitting corrosion (for welded, as well as base. alterial).

Concern existed regarding the potential effects of localized corrosion on the structural integrity of the Fl-SO/EA-SO container and the corrosion effects of various waste stream products, including sulfonated resins. organic liquids. and chlorides; though these utters were addressed indirectly in the report an analysis that was intended to be bounding, that analysis did not provide adequate assurance that every possible corrosive cheMical was accounted for. Certain administrative procedures were to be

.. nted to identify and preclude incorporation of undesirable chemicals.

but the procedural details were not provided.

Substantive information on these .atters was needed before it could be confil"'lDed that the NuPac FL-SO/EA-SO HIC .eets the 300-year structural stability requirHlent.

Accordingly, HuPac was asked (Ref. 8) for considerably information concerning

<a) the metallurgical aspects of F2S5 corrosion, as well as (b) waste stream or other effects. The following discussion of F2SS corrosion addresses the review in the context of these two groups of concerns.

3.2.2 Corrosion-related Metallurgical Factors 3.2.2.1 Corrosion Performance of F2SS Welds In addressing the corrosion behavior of welded F2SS. HuPac (Ref. 10) cited (a) certain Metallurgical characteristics of the alloy that rendered it less susceptible than other stainless steels to intergranular and pitting attack and 4 WM-45 STA(f EVALUATION RtPORT (b) welding procedures that would be followed to lessen the 1 ikel ihood of corrosion problems with weldments.

With regard to advantageous aetallurgical characteristics.

HuPac pointed out that the reason that austenitic stainless steels Ire susceptible to heat-affected-zone (HAZ) stress/corrosion cracking (SCC) is that chromium-rich carbides Ire formed at the grain boundaries during welding. lew-carbon versions of the austenitic stainless steels (e.g .* 316L) have been developed to lessen the HAl problem in those Il10ys. Ferralium 255. however, has a typical carbon content of only O.Oll. which is even lower than the carbon content max.) used in the low carbon version of austenitic steels suer. as 316L. According to HuPac. microstructural examinations of HAls in Ferraiium have failed to reveal "sensitization" (i.e .* grain boundary carbide formation)

IS encountered in 316 S5 weldments.

It was 11so Isserted by HuPac that the Electro Slag Remelting process, which is used to produce the Ferralium F2SS alloy. greatly reduces or eliminates the types of non-aetallic inclusions that Ict IS preferential sites for localized Ittack in Icid chloride solutions.

Therefore.

superior performance under conditions conducive to localized corrosion would be expected.

This would be true for weldments as well as parent materill.

To provide assurance that the intrinsic corrosion-resistant nature of as-manufactured F255 would be preserved in welded metal, NuPac affirmed that all welding procedures utilized in the FL-SO/EA-SO HIC fabrication would be developed and qualified in strict accordance with ASME Section IX requirements.

Specific details regarding welding specifications.

required tests. and inspect ions were provi ded in the response (Ref. 10) to HRC staff corrwnents.

Typical drawing. planning.

Ind procurement documentation was also provided . .. During the course of the review of the topical report it became apparent that there was some conflicting information in the literature regarding the recorrwnended welding parameters (e.g., heat input and rate of cooling) for F255. As explained in HuPac's response (Ref. 10) to the staff's questions.

the apparent inconsistency stemed from differences in the wrought versus cast versions of F255. Recent work on welding parameters for F255 has been documented (Refs. 11. 12, 13) by Cabot, Ind HuPac will follow Cabot's recorrwnendations in welding F2SS HICs. Intercomparative data 2 on the Ferralium 2SS duplex stainless steel and 316 austenitic stainless steel were also used as supporting evidence for the 2 Austenitic stainless steels are I class of corrosion resistant alloys for which there is a considerable body of test data and substantial experience (some of which involves underground applications).

Hence, an intercomparison of the FL255 alloy (which is relatively new) with an established older alloy such as 316 stainless steel provides a Measure of the relative merit of the newer lIaterial.

5 WM-4S STAFF EVALUATION REPORT expected satisfactory service performance of F2SS weldments.

In laboratory tests involving the use of (a) potentio-dynamic polarization curves to determine pitting potential in various environments and (b) chloride pitting and crevice corrosion tests, it was shown that while there were instances where the performance of F2SS and 316L 55 was there was no case where the performance of F2SS was inferior to 316L. In S% NaC1, 316L SS welded samples pitted in the weld, whereas no pitting was observed in F255 in the welded or state. Hence, the test results showed that F2SS weldments generally were superior to 3l6L SS weldments.

This demonstrates that F25S welds Should provide even greater assurance of structural integrity and a higher safety .argin regarding the required HIC design life of 300 years than would 316L stainless steel. The performance of austenitic stainless steels in soil environments is discussed in Section 3.2.2.3 of this evaluation report. Based upon the totality of evidence regarding the performance of F2SS weldments and NuPac's procedures for assuring satisfactory performance, the staff concludes that there is reasonable assurance that welding of NuPac FL-SO/EA-SO F255 HICs will not i!llpair the uniform or stress/corrosion cracking resistance of the HIes. 3.2.2.2 Pitting Corrosion Repassivation As noted earlier, F2SS corrosion test results reported in the open literature suggested that uniform and pitting corrosion rates would both be low. F255 .icrostructural considerations, discussed in the previous section, also suggested that f2SS was quite resistant to pitting corrosion, even in the welded state. There was a concern, however, about the potential for non-passivation of corrosion pits, should corrosion pits ever be initiated.

NuPac was, therefore, asked to perform cyclic vol tamrnetry tests on F2S5 to assure that pitting corrosion, if initiated, would not progress to premature loss of structural integrity of the HI'. The cyclic polarization tests, which were performed (using simulated solutions) on base as well as weldments of both the f2SS and 316L SS, showed that there was a lack of hysteresis in all the polarization curves obtained with F255. This result, coupled with the lack of any visible pitting, confirmed the expected high resistance to pitting in F255. In contrast.

significant visible pitting and significant hysteresis of welded 316L SS occurred, thereby demonstrating both the superior pitting corrosion resistance of F2SS as well as the efficacy of the cyclic voltammetry test. 3.2.2.3 Field Experience with Comparative Alloys Due to the relatively short time (less than 20 years) that duplex stainless steels such as F2S5 have been in existence, there is limited field experience with such alloys in soil environments.

SOlIe experience does exist, however, wi th other common corros i on res i stant alloys such as the 300-sed es austenitic stainless steels. NuPac was, therefore, asked to document such field experience (in a variety of soils with the comparative alloys) that would demonstrate reasonably satisfactory performance of the comparative alloys in 6 WH-4S STAFF EVALUATION REPORT those applications.

That experience would serve as indirect evidence that the F2SS alloy would serve adequately 1n the proposed application inasmuch as the F2SS exhibits superior corrosion resistance to the austenitic alloys in laboratory tests. In response, NuPac pointed out that stainless steels have not generally been used in underground applications because of cost considerations and the availability of other less expensive corrosion prevention techniques.

Where stainless steel pipelines hive been installed.

there have been .ixed results, primarily because pipelines cross a variety of soils with varying resistivities that result in the creation of "long-Hne currents" that, in the absence of cathodic protection, will cause corrosion.

Pipelines installed a few feet below the surface of the ground also are subject to corrosion associated with bacterial decay of organic a.terial.

While pipeline experience with lustenitic stainless steels has not been totally satisfactory, HuPac contends that such experience .ay not be completely applicable to HIe burial because HIe*s are buried deeper than normal pipelines and are more isolated electrically.

On the other hand. where stainless steels have been used in small amounts for fasteners, hose cllMps, couplings, and the like in underground applications, the results reportedly (Ref. 10) have been excellent.

Tests performed with 300-series stainless steels in soil environments have generally been good, although in some samples taken from the .ore acidic and harsher soils, some pitting corrosion has been noted. These studies indicate that the common stainless steels, while they show substantial resistance to corrosion in long-term burial applications, also have weaknesses such as pitting. For a given thickness of 8etal. they thus appear to have less .argin to meet the 300-year service life required for HIes. Inasmuch as F2S5 has been delllonstrated to have significantly higher pitting resistance than the common 300-series stainless steels, particularly when considering attack by chloride. (and taking into consideration the expected Chloride concentrations, lIOisture content, and pH levels at the Barnwell and Hanford sites). the staff concludes that the F25S FL-SO/EA-SO HIes will perform better than the 300-series stainless steels would be expected to at those 5 ites. 3.2.2.4 Crevice Corrosion Hypothetically.

there is a potential for crevice corrosion in the area of the HIe between the container and the lid/gask.et.

As noted (Ref. 10) by HuPac, however. crevice corrosion testing performed with 10% ferric chloride and other solutions has shown that the temperature required for crevice corrosion is much higher than the temperatures that would be encountered at low level radioactive waste burial locations.

The burial site environment would, of course, be much less severe than the conditions imposed in laboratory corrosion testing. The staff, therefore, concludes that there is reasonable assurance 7

WM-4S STAFF EVALUATION REPORT that crevice corrosion will not be a significant problem with the HuPac Fl-SO/EA-SO HIC. l.2.2.5 Effects of localized Corrosion on Structural Integrity In the analysis of the structural adequacy of the Fl-SO/EA-SO HIe (discussed in aore detail in Section 4 of this staff evaluation).

a wastage allowance approach is applied to account for unifonn corrosion of the container.

That is, it is assumed that a portion of the total liB inch thickness of the F255 SS is ccrrl)ded away by uniform corrosion, and the stresses developed in the HIe to burial loads are then compared to the allowable stresses.

For reasons discussed elsewhere in this Staff Evaluation, staff considers it unlikely that unifoNn corrosion would result in this lIagnitude of HIC wall thickness loss; rather, it appears lIore likely for the F2SS container to be attacked by localized corrosion.

HuPac was, therefore, asked to provide a structural analysis that would address the potential effects of localized corrosion on structural integrity.

To calculate the .inillun!

weld thickness (the welded areas would be most susceptible to localized corrosion) required to prevent structural instability, the highest stressed element was identified, and an estimate of the allowable pitting damage was obtained by calculating the .axillum allowable uniform weld reduction.

That value (based on a BO,OOO psi y.s. for F2SS) is greater than the wastage allowance for unifo"" corrosion of the HIe wall. The reduction in weld thickness would reduce the welds' aoment carrying capability.

but if a weld were pitted. the remaining non-pitted portion of the weld would still not be reduced in thickness (neglecting uniform corrosion) and would thus lIainta;n a 1I0ment carrying capability.

It would. therefore.

require a gross amount of pitting to achieve a condition of structural instability.

Thus, in view of the inherent superior localized corrosion resistance of F2SS, and taking into account the environmental conditions expected at the Hanford and Barnwell burial sites. staff concludes there is reasonable assurance that localized external corrosion will not threaten the structural integrity of the HIC over its 300 year design life. More information on environmental factors is presented in the following subsection of this staff evaluation.

3.2.3 Environmentally-Related Corrosion Factors 3.2.3.1 General The discussion presented in Section 3.2.2 of this Staff Evaluation centers prillarily on aetallurgica1 factors that govern the corrosion resistance of the Ferralium HIe. In Section l.2.3 the focus is on environmental factors (internal as well as external) that were considered in assessing the 300 year corrosion performance of the HIC. As noted earlier, a wastage allowance (i.e., thickness of lIaterial allocated for corrosion) approach was used in the Fl-50/EA-SO HIC design; that ;s, a portion of the total lIB inch wall thickness

s allocated for uniform B WM-45 STAFF EVALUATION REPORT corrosion.

In assuring that the allowable uniform corrosion rite would not be exceeded, NuPac considered the possible externa' environments of the burial trench as well as the internal environment that would be provided by various waste streams. With regard to the external environment, NuPac Isserted that data on soils and their corrosive characteristics (Ref. 9) indicate that the soils in the current disposal sites are not necessar; ly lIore corrosive than other soi ls where austentic stainless steels have been tested and demonstrated to be highly resistant to both pitting and genera' attack (Ref. 14). While the possibility eXlsts that the burial trench groundwater could, in fact, be considerably agressive than would be encountered in native virgin soils (due to contamination chloride or organic compound-bearing chemicals), NuPac contended that the expected soil contamination levels are well below those that would Iffect the F2SS alloy. Based upon comparison of the burial site soil Ina lyses with corrosion test results Ind field experience with various stainless alloys, the staff would not expect the external (soil) environment to pose a threat to the structural integrity of the FL-SO/EA-SO HIe. (See the following subsections for details.)

With regard to waste stream effects on the internal environment of the HIe, the situation is considerably IIOre complicated because it is I function of lIany factors, including the type of waste, temperature, oxygen concentration, the history of the waste stream, Ind the waste stream itself. It was Icknowledged by NuPac that some detrilllental environments could exist. The analyses and Idminstrative procedures that were developed to address the potential environmental parameters are summarized in the following subsection, 3.2.3.2. 3.2.3.2 Review Areas Concerning Environmentally Related Corrosion Factors In the topical report, the analyses of environmentally rellted corrosion factors focussed primarily on two .ajor areas: (a) soil characteristics (e.g., pH, chloride concentration, water content, organics) and (b) I "worst case" analysis of bead resin corrosion effects. A series of questions concerning these subject Ireas were raised by the staff. The subject lIatter and the responses to the Staffls questions Ire too lengthy Ind complex to cover in detail here, but the following points summarize the situation.

(1) Several pH ranges Ire addressed in the topical report. They deal with the pH range for soils (4.0 to 11.0), the pH rlnge for ion exchange resins (taken IS 0 to 14), the minimum pH for trench sump liquid (assumed to be 2.4) Ind a limiting pH of 3 on liquid bearing waste containing 1II0re than free halogens.

The latter is used to establish a so-called "corrosion criterion" IS follows: liThe liquid portion of the waste lIust have a pH greater than 3. If not, then the waste stream must have less than 2% by weight of ionic halogens." This criterion was developed by considering (a) the lIaxilllum acceptable (uniform and pitting) corrosion rate compatible with preserving structural 9

WM-45 STAFF EVALUATION REPORT integrity; (b) the corrosion rates associated with possible waste streams and (e) practical liMitations imposed on the container by the potential waste fo,..s. (2) The oractical application of the corrosion Hmitations placed on the container is provided in a section of the report that contains the responses to Staff questions that deal with I proposed container operating procedure.

It is intended by NuPac that the procedure should be followed by III users of the Fl-SO/EA-SO HIe. Included with the operating procedure is I cheMical coapatibility flow and check off procedure.

Waste streass that would contain liquids with pH less than 3 or halides (chloride or fluoride) greater than 2% by weight would have to be neutralized.

diluted or excluded from the container.

Other provisions Ire aade for the use of I vent (to Iccomodate potential gas generation due to biodegradation) and short-teT'll tHlperature excursions (to Illow filling of the HIe with lIaterials It greater than ambient temperature).

Users of the fl-SO/EA-SO HIe will be required to certify that they have complied with all the operating procedures Ind that the HIes do not contain proscribed chemicals.

A copy of the Operlting Procedure required for fl-SO/EA-SO HIe users is provided IS an appendix to thh evaluation report. (3) Regarding the chemical COIIPatibility of ion exchange resins with the HIe, a theoreotical "worst case tl analysis was presented in Appendix Q of the Is-submitted report. Rather than rely solely on that Inalysis.

the NRC staff asked NuP.c to (I) propose the wlste strea.ms that the Fl-SO/EA-SO HIe would see the products of. (b) examine the Ipplicable test data, Ind (c) show by Inalysis that the environment that the HIe will be subjected to would not be unacceptable.

In response, NuPac presented In Inalysis that centered around data concerning the titration of ion exchange resins Ind the pH of contacting wlter. It was shown, that even with very low pHs (sillulating radiation damage effects), corrosion rates were well within the uniform corrosion 11.it for the HIe. A revised Appendix Q was submitted as I theoretical backup analysis for an analytical clse. The results of the Appendix Q revision indicated that dewatered resins could sillulate 10-20% sulfuric acid, which while it was considered excessive for 316 stainless steel, would not result in violation of the uniform corrosion lillit for F2S5. (4) In Iddition to the above points, NuPac also addressed (a) the potential need for organic solvents exclusion and pre-treatment, (b) the potential for growth of .icro-organislls. (c) effects of sulfur compounds, (d) trench and organic liquid chemical corrosion resistance, (e) chloride content of soils, and (f) effects of radiation on pH. In all cases, the Ferralium container was shown, on the basis of Inalyses coupled with appl icable 10

STAFF EVALUATION REPORT data, not to be significantly affected by the postulated plausible environ.ental condition.

The staff conc' udes, on the bas is of the Ina lyses and data presented in the Fl-50/EA-50 report ,..,d responses to Staff questions that there is reasonable Issurance that the fL-SO/EA-50 HIe, if used within the bounds prescribed by the proposed operating procedures, will not suffer I loss of structural integrity over its 300 year design life due to corrosion effects. Verification of acceptable perfoY'llance can be provided by means of J)trlodic surveillance of archival specimens (see Section 3.9 of this Staff Report). It should be noted that users of the Fl-SO/EA-50 HIe will hl'le tC' comply with III state requirements and criteria for a particular LLW burial facility.

For example, South Carolina requires waste forms to be within a pH range of 4 to 11. That requirement wi 11 thus apply to any Fl-50/EA-50 HIes that are buried at Barnwell, regardless of the pH <3 "corrosion criterion" proposed by NuPac. 3.3 Structural Analyses Burial depths at the Hanford, Washington site do not exceed 45 feet, which corresponds to an external pressure of 37.5 psi on the container, while the 25 feet maximum burial depth at Barnwell, South Carolina corresponds to a container external pressure of 20.8 psi. In the original design of the fL-SO/EA-50 HIe, the side walls were 1/4 inch Ferraliu., and the HIe had only two internal supports.

Reanalyses by NuPac, however, led to two .ajor design changes that were related to the structural analyses of other lIembers of NuPac's Enviralloy HIe family: (1) an increase in the HIe wall thickness to 3/8 inch, and (Z) the use of four internal supports.

These changes were intended to improve the structural design aargin_for the HICs. In examining the February 1985 responses to NRC Staff questions, however, it was discovered that there were sOlie areas that required further clarification and elaboration.

These included, in addition to some aspects of the structural analysis, they included sOlie aspects of the special vent design, proposed short term temperature lillits for the loaded Enviralloy (FZ5S) HICs, and the need for a clearer commitment to provide survei llance specilnens.

These concerns were transmitted to NuPac both orally and in writing (Ref. 15), and resulted in substantial revisions to the topical report and in responses to questions that were resubmitted (Ref. 16) in May 1985. 3.3.1 Burial loads One of the areas in the HIe structural analysis that required further attention was the effects of burial loads. Basically, the Staff concluded that it had not been Idequately demonstrated that the HIe could withstand the predicted burial loads. Specifically, additional information was required (Ref. 15) concerning (a) the calculation of a critical buckling stress, (b) applied loads resulting from placement of the HIe in a non-vertical position ;n the burial trench, (c) the determination of an allowable stress intensity value, and (d) 11 WH-4S STAFF EVALUATION REPORT various details of the structural Ina1ysis of the internal vertical angle supports.

In I telecopied response (Ref. 16(a>>, which was later a fonna1 submittal (Ref. 16(b>>. NuPac satisfactorily addressed the staff's concerns.

In brief. it was demonstrated that (1) the HIe did not have a stability problem due to buckling (2) there was significant lIargin for loading due to side burials of the HIes Ind (3) the stability of the internal vertical supports adequate.

While the staff did not accept NuPac I s approach for deri vi ng an stress intensity for the prilllary lIembrane plus bending stress, the of opinion was .oot inasmuch IS none of the burial stresses in the conta i ner, whether in the as fabri cated or "corroded" (mi nus the was tage allowance) state. exceeded the published yield stress of 80,000 psi for Ferralium 255. It should be noted that NuPac analyzed the FL-SO/EA-50 HIe for displacement and stresses utilizing a general purpose finite element code called ANSYS (Revision 3, Update 67L). ANSYS is I widely used and accepted finite-element analysis tool that has undergone extensive benchmarking to demonstrate its reliability for structural analysis.

The assumptions used in Ipplying the ANSYS lIodel to analyze the behavior of the FL-50/EA-SO HIe under various loadings are described in the structural analysis section of the topical report. A discussion of the .lements used and the output generated by the code are provided in various appendices of the topical report. The staff concludes, on the basis of the inforaation provided, that there is reasonable assurance that the FL-SO/EA-SO HIe is adequately designed for all conceivable burial loads. 3.3.2 Drop Test Load Analyses In addition to the analyses of burial .. loads, NuPac Ittempted to estimate the loads that would be incurred on various components of the HIe during the drop testing of HIe prototypes.

Those calculations, presented in Section 3 of the topical report, addressed such things as the load on the lid during flat-ended and corner drop tests. Several questions were raised by the staff concerning these analyses.

Most of the questions dealt with the need for clarification of portions of the report text. A couple of the questions concerned the values used for the maximum payload and gross weight of the container.

In response, NuPac stated that the drop analyses were performed to provide an approximation of the conditions that would be imposed on the HIe during the drop tests and that the actual qualification of the container was based on the drop test results (see Section 3.4). Clarification of the report text provided where needed, and certain typographical errors were corrected.

With regard to the container gross weight, NuPac stated that the lIIaximum gross weight of the FL-SO/EA-SO HIe is 4200 pounds and that the user will be required to limit the HIe contents such that this gross weight is not exceeded.

The 4200 pound limit .eets shipping container licensing requirements.

12 WM-45 STAFF EVALUATION REPORT 3.3.3.

Stresses The HIe will be subjected to some the",al loads due to solar heating during transportation.

Differential the naa 1 txpansion between the container and the lifting straps, for tUllple, could occur, and a "worst case" or bonding value was calculated.

A quantitative analysh of the re5ultant stresses in the straps or surface of the HIe. requested by the staff, showed that there was a significant safety factor, based on the difference between the thermal stress and the yield stress of the aaterial.

With regard to burial the,..a' loads. the relatively low burial temperature envelope at Barnwell and Hanford (68°Ftl8°F) would not be expected to be a factor. Mechanical strength properties of f2SS decline gradually with increasing temperature (e.g., strength properties at 200°f and 400°f are reportedly 8.6% and 12.6% less. respectively, than roOflll temperature values). Therefore.

Iny increase in temperature of the HIC that .ight ensue due to soil insulating effects or the near proximity of other heat-generating wastes would not be expected to significantly affect the HIe. likewise.

temporary storage Ibove ground in a storage facility would not be expected to be a significant factor. 3.4 Prototype Testing 3.4.1 Drop Tests The HIe should be capable of eeeting the requirements for I Type A package as specified in 49 eFR 173 Ind 10 eFR 71, as applicable to lIetallic containers (Ref. 2). With regard to drop test requirements.

the applicable criteria are provided in 10 efR 71.71. For the fl-SO/[A-SO HIe, which will have a gross weight under 4250 pounds, free drop tests (with the HIC loaded to the maximum gross weight) onto an unyielding surface, from I variety of orientations (i.e .* flat Ind corner drops) were perfor.ed.

Except for a dent about 1/4 inch deep in the side wall (of. HIe with the original 1/4 inch wall) after I corner drop test, no visible dllllage ensued. I.portantly, there was no loss of contents from the container due to cracks or rupture of the seal.

results were obtained fro. a full series of drop tests performed from 25 feet onto compacted lind. In this series of tests, the container included a lead gasket. The lead gasket llaintained a positive seal. The only visible dlll\age that ensued frOil the 25 foot drop tests cons isted of I denting (about 5/8 inch .axiaum) of the iapacted side between the two end plltes following a side drop. There WIS no loss of contents resulting from any of the 2S foot drop tests, nor did a eagnetic particle test performed on the closure welds indiclte any loss of structural integrity.

Angles welded to the lid that serve as hlndles were broken .t the welds .fter the 2S foot top down drop test. but these Ire non-structural components of the container and their failure did not affect container integrity.

After one drop test, which was .n early test conducted on I container with a gross weight of only 3000 pounds, I crack was detected in one of the welds. 13 WM-45 STAFF EVALUATION REPORT That crack was determined to be due to

• weld defect, however, and was not the result of I design deficiency.

NuPac has provided assurance that future inspection procedures, to be used on production containers, will preclude the presence of sil,; lar weld defects. The staff concludes, on the basis of the submitted that the Fl-SO/EA-SO HIe has satisfied the criteria for free drop tests for high integrity containers specified by NRC staff and the Statn. 3.4.2 Type A Package Criteria A higrl container for low-level radioactive waste should be capable of lIIeeting the "normal conditions of transport" criteria for Type A packages in 49 CFR 173 and 10 efR 71, as applicable to .etallic containers (Ref. 2). Criteria used are those contained in Section 71.71(c), 10 CFR Part 71. Of the Type A package test criteria, the results of drop tests are addressed in Sect ion 3.4.1, above. Other tests, or analyses perfoT'lled in lieu of tests, are addressed in the following sections.

Penetration Test A penetration test was performed using the criteria in 10 CfR 71.71(c)(10).

In this test a vertical steel cylinder 1-1/4 inch in diameter, weighing 13 pounds, and with a hemispherical end, was dropped from a height of 40 inches onto an exposed surface of the container with no Measurable effect. Water Spray Test Since the FL-SO/EA-SO HIe is fabricated from a duplex alloy steel, the water spray test (which simulates exposure to rainfall) described in 10 CFR 71.71 (c)(6) was not performed.

The staff concurs with NuPac's position that aetallic stainless steel packages will undergo no lIIeasurable physical change when exposed to the equivalent of two inches of rainfall for one hour. Vibration Testing The test criterion for vibration normally incident to transport is contained in 10 CFR 71. 71(c)(S).

Inasmuch IS the FL-SO/EA-SO HIC is a welded lIIetallic structure with which closure is accomplished by 8 retaining blocks that lock positively into the structure of the container, there is no credible physical way for shock and vibration nOrlllally incident to transportation to affect the integrity of the HIe. Also, inasmuch as the F255 alloy exhibits low temperature toughness characteristics similar to the commonly used ASTM A516 fine grain practice steels, vibration effects would not be expected to be a prob 1 em even at low telllperatures that IIi ght be encountered duri ng wi nter transport.

Consequently, staff concurs in NuPac's decision not to conduct vibration testing. 14 WM-45 STAFF EVALUATION REPORT Compression Testing Criteria for compression tests are addressed in 10 CFR 71.71(c)(9).

The compressive load to be applied to the HIes during these tests IIUSt be either the equivalent of five tilles the weight of the package or 1.85 ps: lIultiplied by the vertically projected area of the packages, Whichever is greater. As noted in Section 3.3.1 of this staff evaluation, however, the Fl-SO/EA-SO HIe is designed to withstand burial loads of at least 37.5 psi (corresponding to the 45 foot burial depth at Hanford).

This corresponds to a projected load that is lIore than three tillles the 21,000 pound load that is obtained by lIultip1ying the 4200 pound gross weight of the container by a factor of five. Therefore, the compression test was not conducted on the Fl-SO/EA-SO HIe. The staff agrees with NuPac's contention that the test is not warranted for this particular HIe. Pressure Testing The criterion for a "reduced external pressure" test, corresponding to an external pressure of 3.5 psia, is contained in 10 CFR 71.71(c)(3).

This corresponds to a pressure differential of 11.2 psi (that is, 14.7 psia internal pressure at sea level atmosphere at tille of tid closure, .;nus 3.S psia). The Fl-SO/EA-SO HIC was pressure tested with a silicone rubber gasket, using water as the pressurization leakage past the gasket occurred at 7S psig. A separate test with a lead gasket, following a drop test, resulted in a positive seal until 20 psig pressure was achieved.

The Fl-SO/EA-SO HIC thus was demonstrated to .eet the reduced external pressure requirements.

No increased external pressure tests were conducted, inasmuch as the HIC, as discussed in Section 3.3.1 of this report, was shown by analysis to be able to withstand the 37.S psi burial loads with aargin. 3.S Gas Generation and Interna' Pressurization One of the design changes .. de to the Fl-SO/EA-50 HIC involves the incorporation of a passive vent system (to be used for non-tritium wastes) to allow relief of pressure generated by gases resulting from possible biodegradation or radiolytic decay. The concern about internal gas generation originated from experience with a few polyethelene containers that exhibited symptoms of excessive gas generation (for example, had become stuck in their transportation casks due to the swelling resulting from generation and internal pressurization).

This had resulted in a request (Ref. 17) by the State of South Carolina Department of Health and Environmental Control for consideration of a passive ventilation system IS I design feature that would alleviate the problem. After due deliberation, The NRC Staff concluded that the installation of vents, in all HICs, not just polyethylene ones, would be a prudent way to address the potential symptoms of the problem with gas generation.

The approach thus provides a lIeans to .inilllize the effects of gas generation (e. g. , over-pressurization of the HIC) on handling, personnel safety, and long-term integrity of the container.

The use of vents is intended to be an interim 15 WM-45 STAFF EVALUATION REPORT Measure, which would Iddress the symptoms Ind preclude Iny serious effects of gas generltion, while l110wing a long-term solution to be Irrived at via a study that would identify the specific cause of the gls generation.

Accordi ngly, the pass ive vtnt system that NuPac currently proposes to use i r the Fl-SO/EA-SO HIC would be basically comprised of I pertlltable plug of poly.eric aaterial placed in the l1d of the container in a aanner that wi 11 .ini.ize Iny effects on the structure of the container Ind the possibility of damage exterior objects. The vent aaterial was chosen on the basis of its radiation resistance, lack of influence on corrosion, chemical resistance and hydrophobic nature. The vent wi 11 per,dt the relief of internal pressure by allowing the passage of gas while still minladzing the ingress of water as recommended by the Technical Position on Waste (Ref. 2). Samples of the polymeric aaterial have been tested (Ref. l6(b>> for both air and water at various pressures, and have demonstrated satisfactory performance.

The staff concludes that there is reasonable assurance that the passive vent system coupled with the back-up capabiHty provided by the silicone rubber gasket, will provide an Idequate aeans to allow for the release of pressure due to gas generation resulting from biodegradation or radiolytic decay. It should be noted that the passive vent system, though it has been designated "optional" by NuPac, is in fact aandatory because it is the current prillary pressure-relieving system for 111 the Fl-SO/EA-SO HICs except those that will be used for tritiUM containing wastes. In the latter case the HIC will have a lead gasket with no passive vent. This lead gasket/no vent design provides reasonable Issurance of the containaent of the tritium gas. 3.6 Radiation and Ultra-Violet Stability The radiation stability of proposed cpntainer aaterials as well as radiation degradation effects of the waste itself, should be considered in the design of the HIC. No significant changes in aaterial design properties should result following exposure to a total accumulated dose of 10 8 Rads. (Ref. 2) For the Fl-SO/EA-SO HIC, the basic .aterial of construction, Ferralium 255, would not be expected to be .ffected by radiation from low-level wastes. This is so because radiation daaage, in the form of swelling and embrittlement, is caused in .etals by neutron radiation, but these HICs will not contain detectable levels of neutron radiation producing aaterials.

The only components not .ade out of the F2SS alloy are the gasket and the vent. Neither one of these items affect the structural integrity or stability of the container.

However, because the topical report contained information indicating that the silicone rubber gasket material had a 2at compression set after exposure to 1 x 10 7 Rads, further information was requested regarding the testing and capabilities of the gasket. I n response (Re f , 10), NuPac noted that i nfor.at ion in the open 1 i terature (Ref. 18) indicated that a compression capability of about 1at was obtained in testing to radiation exposures of 10 8 Rads. Although this .ight not be 16 WM-45 STAFF EVALUATION REPORT considered sufficient for applications where the gasket .ight be subjected to impact loading (as eight be encountered during transportation), we agree with NuPac's assertion that under burial conditions there is no IIItchanism for the gasket .aterial to Move. The staff concludes that there is reasonable assurance that the silicone rubber gasket will perform as In effective barrier. The optional lead gasket is not affected by gamma radiation at the 10 8 Rad level and is thus also acceptable from a radiation stability standpoint.

Another component of the HIe outer wall that is not constructed of .etal is the passive vent. The vent is basically comprised of a permeable plug of polymeric which reportedly (Ref. 19) has good resistance to gamma radiation in excess of 10 8 rads. Inasmuch as the vent does not clrry any significant load, Iny reduction in .echanical properties that eight occur as a result of radiation will not affect the performance of the HIC. In regard to the effect of radiation on the contents of HIes, NuPac indicated (Ref. 10) that only the demineralization resin media have the potential to be affected by radiation in such a .anner that they May affect the container.

The resin Media May undergo radiolysis to produce gas within the container.

The slow build-up of gas could be I potential problem (with regard to over pressurization effects) only if there were no provision for pressure relief. Inasmuch IS the passive vent will pennit the Illeviation of the pressure, however, the radio1ysis of wastes i5 not expected to result in over pressurization of the HIe. The potential effect of ultra-violet (UV) radiation on the silicone rubber gasket should 1150 be insignificlnt, in view of the fact that most of the gasket is shielded from such radiation by the .. ta1lic lid and top of the HIe during transportation; after the HIC is buried, it will not. of course. be subject to ultra-violet rays. UV radiation effects on the vent material due to exposure during storage would be limited by covering the vent with UV opaque material (see the Operatjng Procedure, Section 5.S). The staff concludes that there is reasonable assurance that the effects of radiation have been Idequately considered in the design of the Fl-SO/EA-50 HIC. 3.7 Quality Assurance and Inspection High integrity container should be fabricated.

tested, inspected, prepared for use, filled, stored, handled, transported and disposed of in accordance with a quality assurance program (Ref. 2). Because the assurance of proper procedures for container fabrication, testing. transportation.

storage and use is critical in several areas, the NRC Staff issued (Ref. 8) several questions and comments concerning this subject. NuPac's responses (Ref. 10) can be separated into two general areas: (1) those lIatters having to do with fabrication, testing and inspection (i.e .* operations performed by the vendor or which are directly under the control of the vendor), and (2) items to be addressed by the user. With regard to the first category of operations, NuPac presented a substantial amount of information.

including documentation on required inspections.

referenced procedures, and specifications and procurement.

All the Fl-SO/EA-50 HICs will be fabricated and inspected in accordance with NuPac "QA level 1" 17

STAFF EVALUATION REPORT criteria.

According to HuPac, the level 1 inspection activity fully meets the requirements of (1) ANSI H 45.2, (2) 10 CFR SO, Appendix B, and (3) 10 eFR 71, Subpart H. This level designation is established after Quality Engineering review of the contract, regulatory, design and fabrication requirements.

Specifically required tests, inspections, lIaterial controls and data review requirements are then delineated in the inspection planning, drawings, referenced procedures and specifications and related procurement documents.

HuPac's program for inspection to assure compliance with lIaterial and construction specifications is delineated in a QA aanual. With to user QA requirements, the Operating Procedure (Appendix of this report) prescribes procedures to be adhered to by users of the FL-50/EA-50 HIe to assure compliance with handling and lIaterial restrictions.

HIC users will be required to certify that all required procedures and restrictions have been satisfied.

The staff concludes that there ;5 reasonable assurance that quality assurance requirements have been adequately addressed for the FL-50/EA-50 HIe. 3.8 Miscellaneous Requirements The preceding sections of this Staff Evaluation Report address the technical areas that received the most attention during the course of the review of the Fl-50/EA-50 HIe topical report. These items received the lIost attention because they were deemed to be the aost critical with regard to influencing the structural integrity of the HIe. The subjects discussed in the following paragraphs of this subsection, though not trivial, were sillpler in scope and in lIost cases easier to resolve than those addressed earlier. 3.8.1 Free liquid The Fl-SO/EA-50 HIe is designed for Cbntaining waste with less than U free liquid by volume. Because various types of waste are to be immobilized within these HICs, a variety of dewatering procedures could be used. HuPac has submitted a topical report, Ho. TP-02, "Dewatering System," dated August 6, 1984 that contains information on the dewatering for these containers.

With regard to the potential effects of dewatering internals on the HIC, NuPac has stated (Ref. 10) that all internal protrusions will be made of a plastic .ateria1.

All .etal1ic parts of a dewatering system would be restrained from contacting the sides of the HIe by either non-.etallic portions of the dewatering structure or by the waste form. Therefore.

the dewatering internals should not pose a problem with regard to (a) forming a corrosion couple with the Ferralium 255 HIe or (b) possibly penetrating the HIC during a drop event. 3.8.2 Design Mechanical tests for polymeric .aterial should be conservatively extrapolated from creep test data (Ref. 2). However, inasmuch as the FL-50/EA-SO HICs are to be fabricated from a high strength stainless steel (Ferra1ium alloy 255), creep of the stainless steel will be negligible under any conceivable condition that the HICs might have to endure. With regard to 18 WM-45 STAFF EVALUATION REPORT creep of the gasket, there is .etll-to-aetal contact between the lid and the body of the HIe when the HIe is closed; therefore, the effects of gasket creep on HIe integrity are expected to be insignificant.

The vent also is designed such that the creep load will be relatively low. and any effects of creep would not the service of the vent or integrity of the HIC. Hence, creep were not considered quantitatively in the review of the design of the Fl-SO/EA-SO HIe. 3.8.3 Biodegradation The biodegradation properties of the proposed HIC .aterials, wastes, and disposal lledil should be considered in the HIe design (Ref. 2). Certain standardized tests are called for in the HRC Staff TeChnical Position on Waste Fonn (Ref. 2). In the initial version (Ref. 6 and 7) of the Fl-50/[A-50 generic topical report, biodegradation is addressed (see Section 2.0, Qualification of Container Material).

As noted therein, biodegradation of a tital can be defined as the deterioration of the _tal by corrosion processes that occur directly or indirectly IS a result of the activity of living organisllls.

Subsequent discussion then addressed various aspects involving the presence of aerobic versus anaerobic bacteria.

For clarification, the HRC Staff requested (Ref. 8) additional fnforaation concerning (a) the effacts of potential sulfur-bearing cOllpounds in the waste, (b) the .agnitude of potential gas generation, and (c) the potential effects of .erobic bacteria in anoxic environments. "uPac's response (Ref. 10). which was quite comprehensive, basically can (along with the information in the original report) be summarized as follows: (1) (2) (3) Any gas generation that .ight occur within the container would be relieved by the special vent. or if the*vent were plugged by sOlIe unforeseen process, by the lid gasket (which under test was detected to leak It about 20 to 75 psig for the lead and s11 icone rubber gaskets, respect hely). Given the limited amount of oxygen and light within the interior of I HIe, the only possible sustained growth of .icro-organisms is through .icrobes that .etabolize fatty acids IS a carbon source. The IDOst common fatty acids are rarely uled at ceamercia1 power plants. and if they were they would. in .cst Clses, be in low concentrltions.

If sulfate, sulfite. or other sulfur-bearing compounds were present in the waste that is placed ;n the HIe. and/or should the growth of either aerobic or Inaerobic bateria occur. the end products would be low concentrations of sufuric acid and hydrogen sulfide. As described in the report. however. Ferral h.-255 has been shown to be very res 15tlnt to corrosive attack by such chemicals.

Therefore.

the effect of their potential presence on the perforaance of the Fl-50/EA-50 HIe is expected to be insignificant.

(4) An explanation of spedfic .1crobe .etabolhlll

.. thods. possible 19 WM-45 STAFF EVALUATION REPORT complicating effects of prolonged waste times. and a list of the .ost cOll'lllon fatty acids were submitted IS an attachment to the response (Ref 10) to Staff questions.

The Operating Procedure.

to be followed by HIC users. addresses the practical application of limiting organicf the length of and other appropriate related concerns.

While staff does not believe that NuPac's contention about the role of fatty acids in the biodegradation process is particularly persuasive.

because there is contrary evidence available from experience with operating reactor wastes, the fact h that (a> Ferralium 255 is very resistant to corrosion, (b) operating procedures (Appendix A) will preclude the loading of the most potentially troublesome waste lIaterials, and (c) the passive vent will alloW' for rel1ef of Iny internal pressure generated by biodegradation of containing deleterious chemicals such IS fatty acids. Considering these factors. the staff concludes that there is reasonable assurance that (a) biodegradation of the HIC .aterial (Ferralium 255) is so extremely unlikely that biodegradation testing of the alloy in accordance with ASTM or other standardized tests is unnecessary.

and (b) significant biodegradation of leading to a loss of structural integrity of the HIe (resulting from, for example, corrosion of the F255 alloy or extensive gas generation that would not be Il1eviated by the passive vent) is also unlikely.

3.8.4 Top Surflce Water Retention The HIC should be designed to avoid the collection or retention of water on its top surfaces to minimize the accumulation of trench liquids that could result in corrosive or degrading effects. NuPac has designed the HIC so that the retaining ring at the center of the upper head is slotted such that any entering the area can drain back out. All areas at the top head are designed to be self draining.

The staff concludes that there is reasonable assurance that there will not be a corrosion problem with the Fl-SO/EA-SO HIC due to collection or retention of water on the top surface. 3.8.5 Cold Weather Testing The test "criteria" for evaluating the container under normal conditions of transport includes determination of the effect of ambient cold temperatures as as -40°F on the HIC design. Concerns about cold weather testing expressed by the State of South Carolina (Ref. 20). and a multi-part question (No. 16c) regirding the impact resistance of Ferralium 255 at low temperatures was generated by the NRC staff (Ref. 8). In response, NuPac submitted (Refs. 10 and 16b) charpy impact data on welded Ferralium It temperatures as low as -100°F. While the impact strength of F255 weld lIetal decreases substantially with temperature.

the charpy impact values for weldments, at OOF for example, varied from greater than 10 ft. lbs. to approxillately 20 ft. lbs. Even at -40°F. weld metal charpy impact values equal to or greater than 8 ft. 1bs. (Ferralium 255 base metal exhibits much 20 WM-45 STArF EVALUATION REPORT higher toughness values than the .aterial at low temperatures).

Allowing for (a) the inherent difficulty in performing drop tests on fully-loaded Fl-SO/EA-SO HIes at temperatures as low as -40°F and (b) the fact that the charpy i.pact tests on weld aaterial demonstrate significant toughness at low temperatures, tht staff conclude that there is reasonable assurance that weather will not present an undue hazard with the FL-SO/EA-SO HIe and that further testing at low temperatures is not required.

3.9 Surveillance demonstration of the adequacy of any HIC design would involve three things: (1) laboratory testing, (2) analytical predictions, and (3) field experience.

Because field experience with F255 in soil is sparse, there is some uncertainty regarding the possibility for synergistic dfects or environmental degradation phenomena whose magnitude it .ay not be possible to predict or whose nature it may not even be possible to identify at this time. Final confirmation of the adequacy of a new HIe design such as NuPac's FL-SO/EA-SO can, however, be provided over through inspections of surveillance speci.ens buried at each licensed disposal site. NRC is considering a plan for establishment of surveillance protocols involving "archival trench" budals of HIe specimens (and ".in; -samples" of HIe materials) at llW budal sites. NuPac was requested (Ref. 8) to agree in principle to providing F2SS surveillance specilllens for use in a long-term surveillance program, with the understanding that the details of the program can be established on a schedule independent of and possibly subsequent to, the approval of the FL-SO/EA-SO HIe design. In response (Ref. 16b). NuPac expressed a positive interest in supporting a survei 1 lance program, centering around .. an "archival trench" concept in which surveillance specilllens (for example. corrosion coupons or an actual HIe) could be placed for subsequent periodic retrieval and inspection under an established protocol.

Until the specific details of such a program have been established, it is not practicable to mandate particular requirements or to expect vendors, bud a 1 site operators, state agenc i es. etc.. to lIake c i rcums tant i a 1 commitments.

However. it should be noted that verification of the adequacy of a HIC design and materials of fabrication can only be provided directly through actual surveillance, which would involve periodic inspections over several years. 4.0 REGULATORY POSITION NRC staff has completed its review of the topical report that is intended to serve as the referent i a 1 document that describes the des i gn ,of the NuPac FL-SO/EA-SO high integrity container (HIe) for low-level radioactive waste and provides the basis for determining the adequacy of the HIC design. In its evaluation staff primarily focussed on (1) applicable sections of 10 eFR 61, 10 eFR 71, and 49 eFR 173 and (2) additional requirements proposed by state agencies.

Based on its evaluation of the infonnation provided in (a) the topical report (original submittal plus revisions), (b) written responses by 21 WM-45 STAfF EVALUATION REPORT NuPac to NRC Stiff questions and cOlII!Ients, and (c) lH!etings and telephone discussions with NuPac representatives and consultants, the staff conclude that there 1s reasonable assurance that, considering the proposed use of the NuPac fl-SO/EA-SO HIe, the HIe .. ets the structural stability requirements of Part 61 and is consir",ent with the guida",! presented in the NRC staff Technical Position of Waste form. This approval of the FL-SO/EA-SO HIC and Topical Report is predicated on completion and issuance of the final Top;cal Report (proprietary and non proprietary versions) according to reviellt' agreements and the follo\liing conditions: (l) That the Fl-SO/EA-SO HIC shall be used in accordance with the Operating Procedure restrictions outlined in the Appendix to this Technical Evaluation and all additional restrictions and requirements specified by the burial site operators and governing state agencies.

(2) Users of the fL-SO/EA-SO HIe shall certify that all restrictions and required procedures have been adhered to and that the HICs do not contain proscribed chemicals or waste materials.

Based on responses (Ref. 16) to questions, staff understands that NuPac will provide appropriate .ateria' speciaens for a surveillance program where corros i on sup 1 es are to be buri ed 1 n an arch; va 1 trench at each llW buri a 1 site and retrieved and inspected at periodic intervals.

22 WM-45 STAFF EVALUATION REPORT S.O

REFERENCES:

1. 10 CFR 61, licensing Requirements for land Disposal of Radioactive Waste, U.S. Government Printing Office, January I, 1985. 2. Technical Position on Waste Form. Rev. D, U.S. Nuclear Regulatory Commission.

May 1983. 3. larry J. Hanson (NuPac), letter to Nancy Kirner (WA). File No. S8436.JCR, November 3, 1983. 4. T. R. Strong (Departlllent of Social and Health Services.

WA). letter to Donald A. (NRC), January S. 1984. 5. leo B. Higginbotham (NRC), Memorandum for Donald A. Nussbaumer, "Techn;cal Assistance to WasMngton State on the NuPac HIC," February 16, 1984. 6. John D. Simchuk (NuPac), letter to Michael Tokar (NRC),

Subject:

Affidavit to Withhold from Public Disclosure NuPac Proprietary Information on the Model FL50 High Integrity Container," File: FL50-G, February 13, 1984. 7. John D. Sillchuk (NuPac), letter to Michael Tokar (NRC),

Subject:

NuPac Model Fl-50 High Integrity Container dated 1/30/84, File: FL50-793, March 1, 1984. 8. Michael Tokar (NRC), letter to Richard T. Haelsig (NuPac), "Request for Additional InfoT'lllation on NuPac's Generic Fl50 HIC Report," October 25, 1984. 9. P.l. Piciulo, C.E. Shea, and Barletta, NAnalyses of Soils at Low-level Radioactive Waste Disposa' Sites," Draft Report. Brookhaven Nationa' laboratory, BNL-NUREG-31388.

May 1982. 10. Charles J. Temus (NuPac). letter to Michael Tokar (NRC),

Subject:

"NuPac FL-50/EA-50/EA-SO.

Response to NRC Questions.

dated 2/85," File:680-15, March 1, 1985. 11. FERRALIUM Alloy 255, Cabot Bulletin H-2005. 12. N. Sridhar, l. Flashe. J. Kolts. Corrosion/84, Paper '244. NACE Annual Conference, 1984. 13. N. Sridhar. l. Flashe, and J. Kolts. "Proceedings of ASH Conference on Advances in Stainless Steel Technology.

Detroit 1984. 14. W. F. Gerhold. E. Escalante.

and B. T. Sanderson. "The Corrosion Behavior of Selected Stainless Steels in Soil Environments." NBS Publication NBSIR 81-2228, February 1981. 23 WM-45 STAFF EVALUATION

15. M. Tokar (NRC), Telecommunications with C. J. Temus and S. O. Goetch (NuPac), April/May 1985. 16(1). 16(b). C. J. Temus (NuPac), Telecopy to Oan Huang (NRC), "NRC Structural Findings," April 24, 1985. C. J. Temus (NuPac), letter to M. Tokar (NRC), May 16, 1985, with Revised Responses to NRC Staff Comments.

17 Heyward G. Shealey (SC), letter to Stephen Goetsch (NuPac), September 27, 1984. 18. Robert Barbarin, "Selecting Elastomer Seals for Nuclear Service," Parker Hannifir Corporation/Seal Group. 19. R. EG. Jaeger, Compendium on Radiation Shielding, Springer-Verlag, 1975. 20. Virgil R. Autry (SC), letter to Donald A. Nussbaumer (NRC), September 17, 1984. 24 WM-45 STAFF EVALUATION REPORT 6.0 APPENDIX orUATIHG PlOCtD\.:u POl !NVt1W.t.Ol 1)ISPOs.u.

WIT'B SD!!S A MDCt) C"...oSlJU OM-32 1 ... 1 AUCiJS'T 29. 1985 .. nat. ,It I IS nlte -nata 9*G*AS Dat.

OH*32. Rev. 1 TABLE OF CONTENTS 1.0 General Scope 1.1 Purpose 1.2 Content 1.3 Applicability 2.0 References 3.0 Definitions 4.0 Lifting and Handling Procedure 4.1 Empty Container 4.2 Loaded Container 5.0 Storage Procedure 6.0 Closure Procedure

6. 1 1 Closure 6.2 Remote Closure 7.0 Waste Compatibility Procedure 7.1 Scope 7.2 Prereouisites 7.3 Chemical Compatibility Check Off Procedure 7.4 Chemical Corrosion to 8.0 Temperature limits for Waste 9.0 Oocumentation and Check Off i 8/85 2 2 2 2 3 3 3 4 4 4 5 8 10 11 11

-. . .... OM-32', RlJv. 1 8/8S G£NtaAL 'CQP! .1..1 Purpoll Thi. document delineate.

aeveral procedurea that are required for personnel and property .af_ty and adherence to the regulation.

for containment and burial of an tnvlralloy IUgh Integrity Contain.r (RIC). l..r.l Content !hi. procedure cSe.criba.

tbe method. and tachnique. quired to operate any container in the family of Bl;h Conta1nerl from tabrication through burLal. It 1. an all encompa ** in; generic procedure unl **** pecific al:e, cU'tomer, or application are indicated by the procedure cover page and Section 1.3, Applicability.

may be attached a. nece ** ary. Any addendum.

are noted in the Table of Contenta and Saction 1.3, Applicability.

l.l Th1. procedure appllea to the related activitie.

of all Nucl.ar 'ackaging, Inc. employee., their contract .onnel, utility cu.tomer.

and thair contract perlonnel.

Any applicable personnel that handle load, procure, Itore, and .hip the container .are bound by thil procedure.

1.1 United State. Cod, of 'ederal Regulation.

Title 10 'art 61 On1ted Itlte. Code of 'ederal Regulation.

Title 10 part 11 Rucl ** r Packaging C ** k handling procedure.

lhlcl ** : Packaging Qua11ty A.aurance Program, Approval No. 01'2 1 N1\C

. ... . . .... *.. j *

..... OM-l2, 1 fI/8S 1.1 luclta:

Inc. Enviralloy H1gh Int.grity talnerl 'epical .epert Ru'ac .rocedure e,-os, Cleanlng of Envlralloy tainerl RuPac 'recedure

10. L'-l', Ceneral Procedure

-Soap Bubble CLow 're ** ure) for Envlralloy Containtrs Nupac 'roc.dure RO. r5-01, See for rab/Mach of Sttel 'art. Criteria for High Integrity Container', Wathington State Radiation Control Program, Augult 2S, 1'S3. l..a.lJ. us NRC Pinal Wa.te Cla.llflclt10n aneS Wa.te Porro Technical Po.ition 'aper., May 11, 1"3 Q!PINITIONS SIC: 11gh Integrity Container Liquid Pree Wa.te, nry WI.t. luch a. dr1eeS filter., DAW, hardwart ttc. l.] DAN I nry Activated.

Waitt A.l Empty ContI in,; .. The tmpty container.

cln be lifted by any on. of the normal lifting connectionl (lifting .11ng., 11fting padeYI or lifting lye) or by lifting btneath the talntr with a forklift or othlr lultabl. device luch as I 11ftlng platform.

ear ** hould bt taken not to drop Of damagt the contaln.r.

The tare weight. of the ta1ner. arl noted in Tabl. 4-1. !.1 Loaded kpntaincr Lift the loaded container only by the lifting .ling **** mbly or the lpeelal lifting lug. designld for aote handling equipment or from bentath the container with a forklift or 11fting platform.

The maximum gro.s weight of each containlr 1a lilted in 4-:. 2

• OM-32, lev. 1 alB! 1a..O. KoeSe1 IA-2101 f!A-210B lA-l'Oa lA-l'OI lA-le2a £A-1421 £.\-1408 EA-UDI !A-7-l00a

£A-7-100B

£A-6-1008

!A-6-l00B

£A-50B £A-501 tabla 4-1 tarl wlight (lb ** ) '7'0 3450 1455 JO'O 2sIS 2545 2e30 218S 2640 2545 2110 2060 le35 1435 .-

ZBCC:tlaa, Gro ** We1gbt (lb *. ) 20000 20000 20000 20000 10000 10000 15000 15000 13000 13000 12000 12000 4200 4200 The containerl Ihall not be stored where they will come in contact wlth an env1ronmen:

that violate. the requirlment of 7.4 Store the clolurl ga.ket 1n a cool dry place out of direct lunlight.

Prot.ct the clolure ga.k.t. from abrallon, cutting, harlh chemicall and fum ** or exceilivi loaded ptellure during Itoraqe. Tatl precautionl to prevent the containlr from filling with rain water. Store containerl 1n an area vbere they vill not IUltain impactl, abra.1onl, 90uging, or other damage. Vent mUlt be covlr.d during Itorage witb

• ultraviolet (UV) opaque cover (l.e., black polyethylene, blaCk poly vinyl Chloride tape, etc.). 3 8/85 C10IU;, 1.1.1 f.1. 2 '.1. :) 6.1.4 6.l.S Cle.n ae.l .rt. both on contain.r .nd on the lid to remove .ny cUrt, 'r **** , 011s, or oth.r debr i ** In.pect , ** k.t tor .ny cuta or damag *. place it n.c **** ry. ,l.e. lid on gasket and .11;n h.ndlt ** 0 they .ce betw ** n clolur. w.dge hol.a on the leri ** A conta1n.r

• • Plac. vedge. !n hole ** nd driv, until ** eurl. Th. w,dge ** hould be driven until the lid 1a metal to on the atopi und.r the lid.

tht wedg ** do not normally reqYlre dr1v1n; to their full ramp len;th. Remov. v.nt UV cover. Bccotl C101Yt* 1.2.1 1.2.2 1.2.3 Plrtor. Itep' '.1.1 through 1.1.3 DrivI wI4ges in place ua1n, a r.aote clo.ur. tool.

vent OV cover. NOT!:

PROCEDOI!

S!C!ION APPLIES TO ALL PERSONNEL AS OUTLINED IN SECTION 1.3, APPLICABILITY.

SECTION MAY BE PARTICULARLY TO THE PLANT CHEMICAL MATER!ALS COORDINATOR, IADWAS'r!

SUPERVISOP., RADWASTf TlANSPORTATION SUPERVISOR

AND, TO TBOS! wao OS! THE CBEMCIALS SOC! AS THE APPROPRIATE OPERATIONS, CBEMIS!RY AND MAINTENANC!

ClOOPS. l...1 'sPR' '.1.1 'UCPOI' Th. w.at ** at.rial plaCId in the cont.in.r muat be compatible with the oplration of the conta1n.r 1n .ddition to the conta1ner'.

uterial corrolion properti...

V.ritication ot the compatibility ot the w ** t. and the proc ***** performed on it 1. requir.d to m ** t the applicable .af.ty, tranaportat1on and burial requirem.nt.

ot a 81gh Inttgrity t.1ne r (HIC) *

• OM-32", Rtv. 1 ,.1.1 8/85 Tht va.t. compatibility

i. designld to requlrt minimum ItlPI and no plant ell anlly.!..

Tha procedure rlquirt. le ** than! .tepl. Applie.bil tty Waite compatibility verification appl ie. to all valte plaCid in the contlinlr regardle ** of tht nature of the material or mixture. It includ ** , but i. not limited to: 7.1.3.1 Ion exchangl rl.inl 7.1.3.2 Cartridge fi1tlrl 7.1.3.3 Cloth uttrial 7.1.3.4 'aplr valte'I othtr Imall tainerl an. thlir contentl, 7.1.3.5 lardvarl and tht liquidl coating it 7.1.3.6 Stabilization .edia and the cal. incorporatld in the tion .edia. !.1 lr.r'guilit'l 7.2.1 7.2*3 gtilit i ,. Tee1* No utlliti.1 or tooll ar. required for thi. part of the procedure.

Ath.r Z,os,durel Cheeklilt.

No other procedure.

are required.

The lilt that 1. a duplicate of rigure 1 il required to complete thi. part of the cal compatibility .ection of the contain.r procedure.

Ifhe flow diagram, riqure 2, i. to be u.ed in conjunction with the chemical compatibility procedure found in Section 7.3. 5

• -.. -... . 8/85 rIGUl! 1 -

PROCEDURE OFF A) CONTAIN!R PRERtQOISITES PER THE PROCEDORE 1.0 U.a r ________ _ Data* ___ _ 2.0 Model Numbar _____ _ Serlal NWl\ber ______ _ 3.0 Wa.te De.cription (cation anion ra.in, filter., atc. ) 4.0 Containar.

handl.d par 4.0 of procedural S.O Container .torad per 5.0 of procadura.

6.0 Chemical Compatibility per Section 7.0. S) The va.ta 1. corro.i.e per .action '.3.1 temperature limit ** et" per .ection 1.0 USAGE V!RtrICA!ION

---

1.0 Container filled v1th dry va.ta or hal been dewlterad per an approved de w aterin9 procadure.

2.0 Clo.ure 2.1 Seal area clean prior to clo.in9. 2.2 Wedge ** ecured per f.l.4 of procedural Verification MO't'E. A COMPL!'1'ED COpy or THIS P'OM SHALL BE ntCLtrDED WI'!'! TS! SHIPMENT or EAC! APPLICAaLE LOADED CONTAINER.

THE ORIGINAL SSALL aE AETAINED BY THE USEl IN ACCORDANCE WITS THEIl RECORD KEEPING 'AOCEDUltE.

Signature

____________________

Titl' ________________

__

011-)2*, l.v. 1 8/BS , .. 1<----------L1qu1d

'r ** Waltt(DAW, Dry Filter., Etc.) I I t No 1 I Y.. \11 'I<-------------pa Cr.at.r !bln ' __________ I I I 11\ I No I ** utralizt I I \11 Yel Dilut., I Gr.attr Than 2 wt.' Cl-.lul F--->WAST!

IS CORROSIVE t I I t I I I 10 I \11 \11 No I------------->Wattr M'di.------>I I I Y.. I I \ll I Cautionary .hra.e on Oxidizer.

I , , .

\11 , I I I I------>WAST!

\11 IS 0.1. THE CONTAINER -Work the flov diagram the procedure found in Section 7.3. 7 01\-:12, 'Rtv, 1 8/8S l.l Ch.misal Ct,es P!oscdur.

The follow1ng check off for chlmical comp.ti* b1l1ty do.s net rlquirl tpecific chemical analYli. or

• plant chlmical invlntory.

Thl chick off proceoure Ilillinatl' .uch .n.1v.1. and invlntorit

• • the cheek of -

con.icer.

thl va.tl .ourci the operating bltere it. chemic.l compotition.

1.3.1 7.3.2 7.3.3 Ovlrall Chemical Compatibility a)

• II thl valte comPlltell frle ef liquids?

rl.in. an damp cloths are Vet) YI. -the va.te 1. not cerrelivl, note on thl check li.t and go to 7.3.2. No -continul.

b). DOl. the va.te liquid, or contact watlr, havi a pH gr.attr than 31 YI. -the va.tl 1. not corro.ivl, notl on the check ll.t and go to 7.3.2. 10 continul.

c!. DOl. thl valtl liquid, or centact watlr, havi grlatlr than 2' by wlight chloride plu. fluoridl 10nl? YI. -the waitt i. corro.lve, note on tht check lilt and go to 7.l .... Ne -thlre are no corre.ive., note on tht chick ll.t and continue.

Wattr Media a). I. the wa.tl lIedia ion exchangl resina? Ye. -continue.

Mo -90 to 7.3.4. Oxidizer Caution NOTE, OXIDIZERS 1)0 ROT POSE ANY PROBLEMS 'flO THE CONTAINER ITSELF. AN OPERATIONAL CAUTION IS n:CLOD!D IN THIS PROCEDURE APPLY INC TO TEE WASTE IANOLING AND PROCESSING THAT MAY !E PEUOR..MED IN CONJUNC"l'ION WI'I'B '1'B£ CONTAINER.

• OM-12, :R.v. 1 '.3 .* 1/8S CAUTION: ION EXCHANGE RESINS WHrN TO SOFFICIENT or OXIDIZING CHEMICALS (MITI:C ACID, ALIALINE 'ERMANGANATES, 'EROXID!S, HYPOCHLORITE!, ETC.) CAN PRODUC!

RANGING FROM INC1lEASED

'fIMP!AA'l't1!'t!S U' TO EXPLOSIONS.

SMALL AJIIOUN'1'1 or a.EANERS ANr USED IN NOlKAL WOOLD ROT IE EXP!CT!O TO B£ A PROIt.EM.

HOWEV!R, LAIG! 1!IAROWARE DECONtAMINATIONS OR LAIG! U!A CLEANINGS COULD POSE A PROILEM. AN EXAMPLE WOOLO IE THE 'l'REATM!NT or 'fHE IINS! WATER FlOM A RECIRC PIPE DECONTAMINATION

'ROCESS. TI! ION EXCHANGE RESIN VENDOR saOOt.D IE WHEN 'l'IElE IS ANY POTEJtIAL rOR LAOOING or OX!%)IZtlS ON ION EXCHANGE l!SINS. the va.te JII.dia il too corrolivt for container, the va.t. may bt dilyt.d, t:alil.d or rinl.d to m.tt the corrol10n cr J. tt ria. Conaul t vi t h Nu'ac ptr 10nn.1.

the .ntire proctdur.

vhtn the livt nature ot the va.te il corrected. Ch.m1cll Cgrrgl10n Chemicall on th11 lilt mUlt not be preltnt 1n the container in lufficJ.ent acidic concentrationa to rode the contain.r pa.t acceptable limitl for a 300 year I1f.. The UI., or tvolytion of hydrochloriC acid above a 2 V":., chlorid. conctntration and It ** than a pH of 3 il the lity.tion to avoJ.d.(pS<3 and Cl-+ >2'vt.) J OM-32, 1 8/85 TABLE 1.1 CORROSIVE CHEMICAL LIST Chemical .ame PO.lible Sourcel Ammonium Anion Ion Exchange carbon Cation Ion Exchangt le.int Chloroform

!)egrea.lr.

Frlont 11, 12, 13, 14, 20, 21, 22, 23, 30, 40. 41, 113, 114, 115, 142, 152, l'SO, 216, 500' * * (Muriatic Hy4rofluoric Methyllnt Murittic Ae:14) lefrigtrantl

-Ste Preont Sta Water Trichloroethylene Tr1chlotoeth.ne Trlfluoroacetic Ae:14 Chloride ** Acid. Treating ** awater with the IYltem L&b Wa.tt. Onu.ed or parti.lly u.14 hydrogen form re.1n Lab Waite. See 'reonl, Trichloroethylene, Trichloroethane lefrigerant .y.tem., lab w.ste., ultr ** onie:

Solvent., deg:e ** 1ng sump intru.1on+acid Solvents, 4eg:e ** 1ng Solvent., de;,ea.1ng 10

( " !!t"1nt '"',

• t , ., .

1It3. ". fer:" I" _:",-. _ard P':'I 1:'\0"'1 ..... 1'5 .,.. " I,or , ( .. " r. V

""!J't"-'tu. , . .,. .... l'\5 * " C. .... .....

_

ijJ]

.... ' .. GPU Nucl ** , Corporation Post Office Box 480 Route <<1 South State of Washington Attn: Ms. N. Kirner Department of Social and Health Services Radiation Control Section Mail Stop LF-13 Olympia. WA 98504

Dear Ms. Kirner:

clot. "".ly .... !!:,,,"_oc

• -I'" .t'5f;;:'PtP

-. ., 7/.1184 'j P"s JMiI . A..Lo.. tr. J\ 1-I<..ZN{ 711110r , I " L I 'I '!IO V \.1 , " ... --->' IS ........... .l.1. .1' J 'III

SUBJECT:

10 CFR Part 61 Variance Middletown, Pennsylvania 1705 'Z -0191 717944-7621 TELEX 84-2386 Writer's Direct Dial Number: (717) 948-8461 4410-84-0008 Document ID 0016A July 17, 1984 The purpose of this letter is to request variance from the provisiOns of 10 CFR Part 61 to permit the proposed processing vessel described herein to be classified and utilized as a High Integrity Container (HIC) to bury Class B and C wastes. GPO Nuclear's proposed use of a processing vessel constructed of an inert alloy (Ferralium 225) as an HIC was discussed at the May 18, 1984, meeting between you and your staff, Dr. K. J. Hofstetter

-GPO Nuclear, and representatives of Nuclear Packaging.

Inc. As discussed.

GPO Nuclear evaluated several alternatives relative to the disposal of radwaste exceeding the Class A requirements of 10 CFR 61 (i.e ** ion exchange resins used for processing water during TMI-2 cleanup).

Alternatives evaluated included:

solidification by cementation both in and ex-vessel; limiting water throughput to Class A limits; and use of HIC overpacks as well as HIC processing vessels. Based on this evaluation, the latter option was preferred.

It provides the least exposure to operators as the loaded liner requires handling only onee, accommodates illlllediate shipment thereby reducing on-site radwaste storage, requires DO off-site contractor involvement, and offers the greatest radwaste volume reduction thus requiring fewer shipments and less burial space. GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation

.{. Ms. Kirner July 17, 1984 4410-84-0008 It is GPU Nuclear's understanding that a variance bas been granted previously to bury radwaste in a similar HIC. As GPU Nuclear plans to construct the proposed HIC processing vessels in accordance with exterior design criteria of the already approved containers, the same general evaluation should apply. Attached is a summary evaluation of the alloy material.

a general description of the HIC. and the details of aodifications required to water processing at TMl-2. The data included in this summary are proprietary information and should be bandIed as such. During water processing at TMl-2, the curie deposition on the ion exchange media will be carefully monitored to ensure that loadings do not exceed the 10. CPR Part 61 Class C limits. After processing.

the HIC vessel will be dewatered, shipped in accordance with applicable regulations.

and buried at the US Ecology disposal site in compliance with the stability requirements specified in 10 CPR Part 61 and the NRC Branch Technical Position on waste forms. Based on the above integrated approach to the generation and disposal of radwaste requiring stabilization, GPU Nuclear requests authority to dispose of the proposed vessels at US Ecology's site. It is estimated that six HIC processing vessels will be generated by January 1. 1986. with.the first vessel ready for disposal in November, 1984. Disposal of the radioactive waste media generated at TMI-2 in this manner appears to be the most efficient method of disposal and meets all requirements of 10 CFR Part 61. If you have any questions concerning this information, please call Mr. J. J. Byrne of my staff. BKK/RBS/jep Attachment Sincerely, /s/ B. K. Kanga B. K. Kanga Director, TMl-2

• . ---July 2, 1984 ABSTRACT 'e' 3 L..) CONr!l')nmAt

-HOT TO sr mSClOSED GPU lG 11:, y WlTtiOUT THE EX-Pi1=S3 \'i::mEN OF PACIFIC Three Mile Island Nuclear Generating Facility-Unit 2, hereinafter referred to as TMI-2, has a need to dispose of filtration and demineralization media generated during the decontamination of highly radioactive liquids. most economical and safest thod of disposal is to use a container that can be used both as a process container as well as a disposal container.

Nuclear Packaging's 50 ft 3 (FLSO) Bigh Integrity Container with a few minor additions (insertion of an Independent Deminera1izer System) meets this requirement.

This container is already in the review process in two forms. Initially it is being reviewed as a container for the disposal of cartridge filters from Arkansas Power in a report submitted to the of Washington Nov. 4, 1983 entitled -Evaluation of Nuclear Packaging, Inc. PLSO tainer-. It is also being reviewed as a generic High Integrity Container for a variety of waste streams by the NRC in a report entitled -Topical Report Covering Nuclear Packaging, Inc. PLSO Container-Docket No. WM-4S. The major differences between the GPU/TMI container and those covered in the reports above are summarized below. More details are included in the sections that follow. Various waste forms are generically reviewed in the SOft 3 HIC (PLSO) topical report submitted under Docket No. WM-4S. TMI-2 waste from water processing consists of sand, zeolite, charcoal or ion exchange resin. Zeolites are a clay type material used for filtration and demineralization.

The choice of media is trolled by the composition of the waste stream to be nated. The major constituents of concern in the TMI-2 waste streams are: a) boric acid solution (up to 8 wt\), b) sodium hydroxide neutralized boric acid (up to 10 wt\), c) sodium xide neutralized phosphoric acid (up to 10 wt\) or d) sodium hydroxide neutralized su1famic acid (up to 10 wt\). None of the waste streams create a corrosive problem for the Envira110ySM container.

When the High Integrity Container is used as a process vessel (deminera1izer) it has a false bottom and internal piping which is independent of the container.

The internals are made of PVC 1

• . July 2, 1984 GPU and stainless steels. The demineralizer components do not cause a corrosive environment to tbe EnviralloySM Ferralium 255 (FR255), as in the case of the false bottom the material is sacrificed to the FR255. Since the false bottom material is anodic to the FR255 the material protects the base container.

In addition to the internals of the demineralizer tbe TMI-2 container has a remote lifting device attachment.

The interface on the container consist of a specially designed lug that can be fabricated out of Enviralloy.

The remote lifting device face permits the SOft 3 (FLSO) container to be handled remotely.

A remote lifting device will be made available to u.s. Ecology at the Banford site for offloading.

To make the SOft 3 (FL50) container a completely remotely operated container, a remote closing device bas been developed.

This device seats the retainer wedges using hydraulic cylinders to force the wedges into place. The above is a brief summary of the differences between the TMI-2 model and use of the NuPac SOft 3 EnviralloySM Integrity tainer. The enclosed attachment reviews the demineralizing HIC in more detail. Please refer to the previously submitted report for specific details on the baseline SOft 3 (FLSO) HIC. The verification of the dewatering process to meet the less than 1% free water critera is covered in a separate section. 2

\ July 2, 1984 EnviralloySM 50ft 3 Demineralizer 1.0 Introduction And Scope The Nuclear Packaging, Env'iralloySM FL50 High Integrity Container CHIC) bas been designed to meet and exceed the criteria defined in 10 CFR Part 61 and liR.C TechnicAl Position Paper c..n IiUfi NuPac's EnviralloySM 50 HIC DeminerAlizer design porates bigh-technology concepts to meet the present day ments. As part of the EnviralloySM 50 design, a superior alloy, lium R Alloy 255 has been chosen to provide a posi tive barr ier for the required 300 years. Ferralium-255 is A ferritic-austentic stainless steel which combines bigh mechanical strength, ductility, and hardness with resistance to corrosion And erosion. The use of this duplex stainless steel alloy together with design innovations And computer augmented stress analysis has culminated in the development of a container with high strength, optimized weight, extreme durAbility and superior corrosion resistance.

Table One is a list comparing the mechanical properties of proven austentic ,alloys to Ferralium-225 CFR 255). The corrosion of FR-255 to waste streams and burial trench ronments is superior to that of fully austentic type 304 And 316 stainless steels. FR-255 has excellent resistance to sulfuric, phosphoric, nitric and many other acids and salts as well as acetic, formic and organic acids and components.

It is larly suitable for concentrations and temperatures where pitting and localized corrosion is a common cause of failure for most conventional stainless steels. The duplex stainless steel base metal and weld material is superior to austentic stainless steels in adverse environments.

The EnviralloySM 50ft 3 Demineralizer utilizes the proven 50ft 3 (FL50) HIC as the container.

The demineralizing process is totally encased by the 50ft 3 HIC. The demineralizing HIC porates a processing cover that is recessed into the 24 inch HIC lid opening. This allows the standard HIC closure configuration to be used wben the resin is saturated.

The demineralizer cess cover design bas incorporated the required number of process fittings which use quick release mechanisms for ease of tion and low radiological body burden. The demineralizer system incorporates a flow through concept. Tbe waste stream is 3

( \ July 2, 1984 f'\-Co L;:> GPO injected into a header which distributes the waste stream to lateral defusers over the demineralizing agent. At the bottom of the HIC, NuPac has incorporated a positive suction system, with a conical shaped false bottom to channel the waste to the outlet distributors.

Again the demineralizer system NuPac has designed is independent to the structure of the liner. A remote handling system has been included in the design. HIC lifting lugs are designed to interface and automatically engage and lock into a lifting harness. As previously stated, FR255 is very resistant to chemical attack. The media that GPO proposes to use are zeolite, sand, charcoal or organic ion exchange resin which all have corrosion tics that are no worse than a weak acid. As listed in Table three of the four waste solutions buffered. sodium hydroxide is typically used to buffer compounds to high pH levels to prevent the stainless steel of the Nuclear Steam Supply System (NSSS) from breaking down. Since the duplex stainless steel (FR225) has better corrosion resistance than austenitic or tic stainless steels the buffered waste will not cause tion of the EnviralloySM HIC. The non-buffered boric acid is reported to have a pH of greater than 4.5. Studies have shown that high PH compounds do not aggressively attack FR255. This fact combined with the low concentration and that boric acid is a weak acid, leads to the conclusion that the boric acid waste stream will have no adverse affect on the Ferralium Alloy 255. Therefore, there is no chemical attack anticipated from any of the resin media. The NuPac EnviralloySM BIC's have exhibited their ability to vive through all of the of the design evaluation and testing. The EnviralloyS 50ft 3 BIC has already gone through the full design and testing program, with excellent results. The EnviralloySM

'SOft 3 BIC has already been analyzed for structural integrity, including loads that would be expected from burial, lifting, handling, and transportation.

In addition, the Enviral-10ySM SOft 3 HIC has been fully drop tested, from 4 ft. as well as from 25 ft. onto compacted sand. EnviralloySM SOft 3 BIC sustained virtually no impact damage in any of the drop rations.

" July 2, 1984 The combination of state-of-the-art materials and methods has culminated in a highly integrated container of extremely high integrity, and a design that is anticipated to withstand the test of time. Oualification EnyirallQY SB Blt Cgntainer Material Structual requirements for reacting physical loading associated with impact and burial pressures can be easily met by many materials.

The most critical concern associated with long term isolation is corrosion resistance.

An extensive search was ducted in order to isolate an optimum metallic material.

The field was reduced to a series of stainless steels. Fully nitic stainless steels such as 316 provided acceptable mental resistance for long term storage in all areas except that of crevis corrosion and pitting. Neither condition tends to reduce the effective thickness of the material for load carrying purposes.

It was imperative that the material chosen exhibited great resistance to pitting and corrosion cracking.

Just such properties were found in a space age material manufactured by the High Technology Division of Cabot Corporation.

The material is Ferralium Alloy 255. Ferralium Alloy 255 is a patented ferritic-austenitic stainless steel which combines high mechanical strength, ductility and hardness with excellent resistance to corrosion and erosion. Ferralium has a minimum ultimate tensile strength of 110,000 psi and a minimum yield strength of 80,000 psi. The corrosion tance of Ferralium alloy 255 under the proposed environment is superior to that of all fully austenitic stainless steels. It is also highly resistant to stress corrosion cracking, crevice rosion and pitting in the burial environment, even in the sence of chlorides and flour ides. Ferralium Alloy 255 is a duplex alloy with approximately equal portions of austenite and ferrite matrix phases. The EnviralloySM High Integrity Container (HIC) will not suffer any Significant corrosion from either its internal or external environment.

The EnviralloySM HIC has been designed to resist external sion due to the disposal site environment as presented in the -Evaluation of Nuclear Packaging, Inc. FL50 Container w* There is 5

,', July 2, 1984 PAlit:. 0 GPU no appreciable seasonal fluctuation in temperature.

A typical temperature is 20 0 C and a design value of 20 0 C + 100C has been used for this submittal.

A soil wetting cycle is not experienced at typical eastern or western sites; a near-steady state condition prevails in each case. For western sites the steady state condition is dry soil since precipitation only percolates a few meters before its downward movement is reversed by evaporation at the surface. For the purposes of this submittal it has been assumed that the water content of the soil can range up to 100%. The chloride content of the soil, is assumed to range from 0 to 6 parts per million. Bigh chromium nickel steels such as 304 and 316 are known to be susceptible to pitting where oXYgen is cluded locally and in the presence of chlorides when the nickel content is less than 9%. is the case with 304 and 316 stainless steels. Ferralium Alloy 255, while it contains only 5.5% nickel, is highly resistant to stress c;:orrosion crackin9, crevice corrosion and pitting. In this duplex stainless steel, ni trogen as an alloying element is effective in mitigating the degree of chromium separation between the austenitic and ferritic phases. Thus, it serves to protect the austenitic phase against chloride attack. Climax Molybdenum Company of Michigan conducted testing on Ferralium Alloy 255 and type 316 stainless steel in a 600 ppm chloride solution.

Ferralium Alloy 255 did not corrode while 316 stainless steel showed a corrosion rate of as much as 7.7 mg km-2/day. Similarily Climax tested these metals for pitting potential.

Ferralium Alloy 255 did not pit while 316 stainless steel showed a pitting potential of 0.060 volts. In fact, of all the duplex stainless steels tested Ferralium Alloy 255 was the only one which did not pit. The chloride level in the soil will not pose a corrosion problem for this container.

For more see the previously submitted report titled -Evaluation of NUCLEAR PACKAGING, INC. FL50 Container-.

The pH of the soil could range between 4.0 and 9.0. The tainer design criteria has conservatively taken into ration a pH range of 4.0 to 11.0 with a soil water content of up to 100%. Ferralium Alloy 255 is highly resistant to organic acids and compounds and is particularily suitable for the higher concentrations and temperatures where pitting and preferential corrosion are common causes of failure with most austenitic stainless steels in the presence of chlorides and other impuri-6 July 2, 1984 GPU ties. A slightly acidic burial environment (pH of 4) is not expected to have any effect on the life of this container. ralium Alloy 255 is also well suited to a caustic environment.

for-general corrosion and stress corrosion cracking tance were performed on this material to determine its ability in severe caustic environments.

tests are very severe and serve only to illustrate Ferralium's resistance to stress corrosion cracking in severe caustic environments.

A mildly alkaline burial environment is not expect to reduce the life of this container. internal environment of this container will be essentially dry (0 to 1 , free standing water) with a broad pH range possible.

Dewatered bead resin represents the worst case, yielding a pH range of 0 to 14. The corrosive effects of bead resin has been analyzed and an estimated corrosion rate has been established.

For details see the generic -Evaluation of Nuclear Packaging, Inc. FL50 Container w* Potential exists for galvanic corrosion between the carbon steel lifting hardware and the Ferralium lifting lug, Ferralium being noble and carbon steel being active. potential generated by a galvanic cell consisting of dissimilar metals caused a flow of current and corrosion to occur at the anodic or active electrode.

Therefore, the steel lifting hardware will probably undergo gal-.vanic corrosion over the 300 year design life. However, the container material, being noble, will not sustain any galvanic corrosion and the life of the container will not be reduced. Biodegradation or biological corrosion is not a type of sion; it is the deterioration of a metal by corrosion processes which occur directly or indirectly as a result of the activity of living organisms.

These organisms include micro forms such as bacteria and macro forms such as mold or fungus. Microorganisms are classified according to their ability to grow in the presence of oxygen. Aerobic organisms grow only in rient medium containing dissolved oxygen. Anaerobic organisms grow by reducing sulfate to sulfide according to the following equation:

SO -2 4 +. 2 --> 7

July 2, 1984 The source of hydrogen for this could be cellulose, sugars or other organic products.

Should the growth of either organism occur, the end products of this growth (i.e. low concentrations of sulfuric acid, ferric hydroxide, thiosulfate, sulfate, sulfur or sulfide) would not corrode Ferraluim.

But the generation of hydrogen sulfide gas within the container is not possible because the organisms would require sulfur compounds to produce it, and there will not be measurable quantities present in the container.

Sowever, if some gases were generated in the container it would be a benefit in that it would tend to equalize the external pressure on the container due to burial. Should the growth of macroorganisms occur within the container, it could be sustained for only a very short per iod of time due to the limited supply of oxygen. The small volume of gases loped would not be a concern because they once again would only tend to equalize the external pressure due to burial. The by products of the growth of macroorganisms (i.e. low concentrations of organic acids) in close proximity to the exterior/of the container will not corrode the Ferralium as it is highly tant to corrosion by organic acids and compounds even in very high concentrations.

All of the information available indicates that the EnviralloySM SIC manufactured from Ferralium 255 will suffer no detrimental corrosion over its 300 year life. Structural Analysis Container Material, Design Ana Lifting Arrangement The design of the Nuclear Packaging EnviralloySM 50ft 3 SIC will provide for stable, durable and, safe container for radioactive wastes. The container material, as outlined in Section 2, is highly stable under all environments to which the container may be exposed. 8 l t. '. July 2, 1984 waste freparation The SOft 3 HIC does not see loads of any significance during waste preparation.

The load associated with the weight of the tainer contents the weight of the container filling.head are insignificant when compared to the loads which the container will see during burial and under conditions of normal transport as defined in 49 CFR 173 .. 398 (b) and 10 CFR 71. Transportation The FLSO is capable of meeting the requirements for a Type A package as specified in 49 CFR 173.412. For detailed structural analysis, see . -Topical Report Covering Nuclear Packaging, Inc. FLSO Container-previously submitted. DeSign SpeCification 4.1 The High Integrity Container The NuPac EnviralloySM SO cubic foot container is a simple right angle cylinder with a flat top and bottom manufactured entirely of Ferralium 255. It is approximately 47 inches in diameter by 51 inches tall. The top and bottom are fabricated from .375 inch material which is over five times thicker than current carbon steel packages.

Ferralium's 80,000 psi yield strength is more than twice that of the carbon steel. Side walls are fabricated from .250 plate. The top head has a 24 inch diameter gasketed opening for loading. Closure of this opening is accomplished with a 3/8 inch Ferralium Alloy 255 plate held in place after loading the container..

This provides the proper gasket sion as well as positive closure of the vessel. The NuPac EnviralloySM 50 Cubic Foot Container (FL50) is similar in size to current packages and will be handled in an identical manner. Lifting of the container is accomplished using a cable sling or a special remote lifting ring. The sling consists of a Single 3/8 inch steel cable which is attacbed to two lifting eyes on the container with anchor sbackles.

Tbe lifting ring is designed to engage and disengage the lifting lugs remotely.

A prototype container has been destructively tested verifing tbe analysis previously presented

.. 9 l

• July 2, 1984 ;he following specification applies to the EnviralloySM 50 Cubic Foot container:

A. Dimensions and tolerances will be as shown on the tion drawing (See Drawing 1). B. A corrosion allowance of 0.125 inches bas been incorporated in the container design. C. container lifting devices have been designed to three times the maximum gross container weight. D. The closure has been designed to maintain a positive seal under all anticipated conditions of usage, including during" impact after a free drop of four feet. E. The container will be fabricated from Ferralium Alloy 255 as manufactured by the Cabot Corporation.

F. design had no identifiable parameters that would reduce the design life below 300 years. 10

\.

• July 2,. 1984 Demineralizer One the features of the EnviralloySM BIC design is that the demineralizer system is independent of the BIC's structure.

The demineralizer system incorporates a recessed process cover/manway (see Drawing 2) that is totally enclosed by the BIC, when the lid is in place. The processing cover is recessed into the BIC by rolling a less steel angle and seal welding it onto the inside of the B IC 24 inch opening. The processing cover is then bolted to the rolled angle, with a gasket (Dura 40) between the two surfaces to insure a positive seal. All of the processing connections are made to the processing cover outer ring. Quick connection tings are used in the connections to provide for a) ease of connection and disconnection process, b) and w ill minimize the radiological burden. The process connection will be specifically designed to interface with the plant systems. The demineralizer inlet and outlet are fabricated form PVC piping and plastic mesh. This design provides two basic benefits; a) the PVC pipe will not damage the FR255 body or lid if the SIC is mishandled and, b) the PVC will not chemically react with any of the anticipated waste streams. The PVC demineralizing system distributes the liquified waste stream evenly over the filter and demineralizer media. This is accomplished with a distribution header feeding meshed laterals.

At the bottom of the deminerializer bed another header with meshed laterals draws the filtered and demineralized liquid out of the HIC through a PVC vertical riser to the outlet connection.

To provide a positive draw point a conical false bottom is vided. The false bottom is made of 0.048 inch 304 or 316 less steel. To insure that there is no entrapped water between the false bottom and the bottom of the BIC, a seal weld is made around each interface.

Cabot Corporation, the manufacturer of FR255, reports that FR255 and Austenitic stainless steels are compatable for welding purposes.

Reviewing the galvanic cell that will result in the Duplex to Austenitic stainless steel interface, it was found that the stainless steel 304 and 316 will be sacrificial to the FR2SS. The 304 or 316 stainless steel will be the active component since the FR2SS is Ilore nobel. As is shown on the galvanic series chart (see Table 3) FR2S5 falls between -Lead-Tin Soldier-and -12'Ni, 18' Cr, 3' Mo steel* on the active side and between -12'Ni, IS, Cr, 3' Mo* steel and 11 l July 2, 1984 r .. __ f '"f * *si1ver-on the passive side. Therefore, the 304 or 316 less steel false bottom is anticipated to under go a certain degree of corrosion.

Since it is totally enclosed inside the FR255 SIC no containment leakages is possible. seal welds will not cause any detrimental affect on the SIC since they are not a structural weld. Liner Bandling Specific SOft) liner lifting lug analysis bas been presented in the -Topical Report Covering Nuclear Packaging, Inc. FL50 tainer 8* The Sigh Integrity Container used for the deminera1izer system incorporates a handling mechanism that allows remote bandling, see Drawings) and 4. Drawing 3, shows the remote closure vice, which allows the container locking dogs to be installed remotely.

The remote closure device is hydraulically accuated by the operator after the processing connections and the portable lid shield (see Drawing 5) are removed. For detailed closure and sealing machanism analysis see the -Topical Report Covering lear Packaging, Inc. FL50 Container*.

After the closure lugs are installed the remote e10sure devise is lifted off. crane is then fitted with the remote lifting ring. lifting ring has been designed to positively lock into the 50ft 3 SIC lifting lugs. BIC is then lifted into the shipping cask for transportation. iaste Demineralizer Media The deminera1izer system at TMI-2 are designed to handle a number of demineralizing media. The media used must filter the waste stream effectively and provide for a relatively non-aggressive environment with respect to the liner. The media that are cipated to be used are listed on Table 2; a) Zeolite b) Sand c) Charcoal d) Organic ion Exchange Resin The zeolite, sand and activated charcoal are all a balanced media. zeolite consists of dry bydrous tectosi1icate ral, which captures large cations and loosely bo1ds water mole-\.. ' cules. sand is clean filter support media. No substantial 12 l July 2, 1984 organics will be present in the sand. charcoal activation process insures that any free sulfates are driven off. If the charcoal contains any sulfate compounds after drying, they would be in the -crystal structure, therefore they would not be chemical leachable. pH levels of the above media and waste stream are well above the critical pH limit (pH of 3) of Ferralium 255 has been proven to be inactive to substances with a pH of greater than three. Dewatered bead resin (organic ion exchange resin) represents the worst case, yielding a pH range of 0 to 14. corrosive fects of bead resin have been analyzed and an estimated corrosion rate has been established, in the -Evaluation of Nuclear Packaging, Inc. FLSO Container-report.

does not appear to be any significant corrosion due to the resin, if the resin is depleted and has a pH of greater than 0.5 for the contacting water. If the pH of the resin is greater than 0.5, the hydrogen affinity has been satisfied. resin used and disposed of in the demineralizer will be well within these limits. Haste Streams reported waste streams can be grouped into four basic cal formulations:

a. Boric acid solutions

<<8 wt ') b. Sodium hydroxide neutralized boric acid <<8 wt ') c. Sodium hydroxide neutralized phosphoric acid <<10 wt ') d. Sodium hydroxide neutralized sulfamic acid <<10 wt ') It has been reported that each of the major waste streams pH range from 4.5 to 9, see

2.

of the four waste streams have been buffered (b,c,& d). While the other a) Bor ic acid with less than eight weight percent is a very weak acid. As described in the -Evaluation of Nuclear Packaging, Inc. FLSO Container-report, the worst possible acid attack comes from hydrochloric acid and sulfuric acid which has been shown to cause little or no corrosive activity.

waste steam will not aggravate the corrosion resistance of the EnviralloySM lium HIC*s. 13

\ July 2, 1984 PI\GE /"7-Of GPU Ie> . 2:5 Dewatering Dewater ing of deminerlizer media has been successfully lished at 'I'M 1-2 since the accident.

More than 50 low level 4 X 4 and 6 X 6 liners containing ion-exchange resin have been buried at US Ecology; all contained less than one gallon free standing water as shown by dewatering and road tests. In order to demonstrate the ability to meet the less than 1% free water criteria required by 10CFR6l, a series of tests will be performed at '1'MI-2 to verify the dewatering capability of the present system with the new liner design. A vessel will be fabricated with internals identical to the Enviralloy 50 ft 3 Bigh Integrity Container processing vessel but with a drain plug_ It will be loaded with a typical mixture of media including li tes, a medium not included in the geher ic topical report mitted by Nuclear Packaging, Inc. Non-radioactive solutions will be circulated through the vessel to simulate operation.

The vessel will then be dewatered using the existing system The amount of free water will be determined immediately after the dewatering, after an extended storage period at TMI-2 and after transport to the Richland, Washington area. All tests will be performed using approved procedures and witnessed and documented in accordance with the requirements of the TMI-2 Ouality ance Plan. With the present proven dewatering sustem and the TMI-2 ence in handling zeolites, we feel confident that the lOCFR61 dewatering criteria can easily be met. 14 l July 2, 1984 HBLE 1 DATA AT ROOM TEMPERATURE Alloy FERRALIUM Type 304L Stainless Steel Type 3l6L Stainless Steel Type 3l7L Stainless Steel Ultimate Stength, hi (MPa) 126 (869) 81 (558) 81 (558) 86 (593) Ferralium Alloy 255 t "' Yield Strength at 0.2\ offset, Iai (MPa) 98 (676) 39 (269) 42 (290) 38 (262) Cabot High Materials-Division

• 15 . Elongation in 2 lin. (50.8mmm), percent 30 55 50 55 l l July 2, 1984 TABLE .2. Anticipated Waste Stream P!\GE GPU z:;-A. Possible media to be loaded in EnviralloySM 50 BIC pated curie loading) a. Zeolites <<1000 Ci 137 Cs + 134 Cs +90 Sr) b. Sand (low Ci absorbance)

c. Charcoal (low Ci absorbance)

d. Organic ion exchange resin <<10 Ci/ft 3) B. Types of Waste Streams a. Miscellaneous waste holdup tanks (MWBT) b. Reactor building decontamination solutions (RBS) c. Reactor coolant solutions (RCS) C. Major chemical components of waste solutions

a. Boric acid solutions

<<8 wt ') b. Sodium hydroxide neutralized boric acid <<8 wt \) c. Sodium hydroxide neutralized phosphoric acid <<10 wt \) d. Sodium hydroxide neutralized sulfamic acid <<10 wt \) D. Expected .pH ranges a. MWBT -(5-9) b. RBS -(>4.5) c. RCS -(7.5-7.8) " 16 l July 2, 1984

• E. Expected activity ranges of solutions processed

a. MWBT -1 uCi/m1 137 Cs and 1 uCi/m1 90 Sr b. RBS -2 uCi/m1 137 Cs and 2 uCi/m1 90 Sr c. RCS -0.1 uCi/m1 137 Cs and 0.1 uCi/m1 90 Sr and 1 uCi/ml 125 Sb 17 July 2, 1984 l l ' TABLE 1 GAI.VANIC SE1UES or eo .... ON Au.oya. Anodic Mar;MSium llacMliulD aJlo)'l Zinc Aluminum.

2S Cadmium Alummum .11011780 T C:arbon Itct'l CutiroD 4 &.0 G% Cr a&eel 12 &.0 Cr IItftI } 16 &.0 ISCPO Cr lteel Active: . 23 &.0 30% Cr IItftI Ni..f'llllilt 1'1> N, "'I> '" ""I } N"a, 18% Cr 14% Ni. 23% Cr *,"1 2l% Ni. 2S'o Cr mel 12% Ni. 18% Cr. 1% )10 ICeeI LNd*tiA JOkier Lead Tin ,,,,,,,I } 10% Ni. 15% Cr Acti\*c Ineone! 10% Ni. 20% Cr Bruscs Copper BroDae Nic:keHih-cr CopOt'f-aickcl It oncl metal N;,'01 } Ni,15%Cr P&!ISh'c Inconc! 800/0 Ni. 20% Cr 12 &.0 14% Cr Acel 'l6 &.0 18% Cr 1Itft1. i% Ni. Cr liftl 8% Ni. 18% Cr lteel 14% Ni. 23% Cr lift! 23 &.0 30% Cr *,,1 20% Ni. 2S% Cr Itrt'l 12% Ni, 18'70 Cr, 3% Mo Itl't'l Sil\"\'r Cathodic Graphite Acth'c P .. h*c GPO P \GE 20 z 5 * . .

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.. ?ilNuclear GPU Nuclear Corporltion Post Office Box 480 Route 441 South Middletown.

Pennsylvania 17057-0191 717944-7621 State of Washington October 26, 1983 4410-83-L-0259 Department of Social and Health Services Attn: Mr. Lee Gronemyer Radiation Control Section Mail Stop LF-13 Olympia, WA 98504

Dear Sir:

TELEX 84-2386 Writer's Direct Dial Number: m-2 Df.atrf.ll

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I ...... Three Mile Island Nuclear Station, Unit 2 (TMI-2) Operating License No. DPR-73 Docket No. 50-320 10 CFR 61 Exemption 1 ( ed on recent conversations between you and members of my staff, GrU Nuclear has been informed of the State of Washington's intentio to change the license of the Hanford Disposal Site to implement appropriate requirements contained in 10 CFR Part 61. It was also learned that this license change is intended to become effective by the end of this year. Although GPU Nuclear has not, as of yet, had the opportunity to study this change, we understand that the new license will require shipments to the disposal site to be classified in accordance with the requirements in 10 CFR 61.55 and un. a-Pat alP. UOD-A.d. I *. .,.

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• ll..,D. J.-Pa .. I.I>. :DIlS-Ad.

I ** iIlCC-'.u1J>

• ;oc-Ge1tbeubll1'l clo I ..... 1. tole saC-Ad. la* .ro... _ I . * ----anlllS--

• • • • OH*E meet the waste characteristics requirements of 10 CFR 61. The purpose of this letter is to request, from the State of Washington, a variance to this change 80 that EPICOR II resin liner could be classified as Class "A" waste and, therefore, be buried in a dewatered condition as is the current case. Under the current license these EPICOR II resin liners comply with condition 27j in that the specific activity of materials with half-lift greater than five (5) years is less than one (1) uCi/cc. Under 10 CFR Part 61.55, however, these liners would be classified as Class "B" waste and require stability in accordance with 10 CFR Part 61.56. l Table 2, Column 1 of 10 CFR Part 61.55 lists the maximum concentration for Class "A" waste. Isotopes of interest to the TMI-2 Recovery P -am are Sr 90 and Cs 137. The limits for these isotopes are C uCi/cc and 1 uCi/cc respectively.

For the rest of the nuclear r'utry, these values are a relaxation of the current license c __ .dition 27j. However, due to TMI 18 unusually high Sr 90 and Cs 137 ratio, these values are more restrictive.

Implementation of the more restrictive Sr 90 criteria for unstabilized waste (Class "A") at TMI GPU Nuclear Corporation is a subsidiary of the General PubliC Utilities Corporation

( Mr. Lee Gronemyer P\GE: L-'; lQ,rl q; 4410-83-L-0259 would result in the generation of approximately ten (10) times more waste than would be generated under the current limit. Compliance with the proposed license Class "B" conditions would also result in an increase of burial volume and ALARA concerns.

EPICOR II -liners are used for miscellaneous processing and for polishing the effluent of our Submerged Demineralizer System (SDS) and they are sodium limited rather than curie limited. As a result, the present curie limits cannot be increased above their current I uCi/cc level because these resins chemically deplete at this level. Stabilization via solidification of resins at this level would result in a 30 to 40 percent increase in volume due to solidification efficiency.

Because the EPICOR II resin liners have no insitu solidification capability, the resins would have to be sluiced from the EPICOR liner into another container.

The sluicing activity and volume increase from solidification would cause additional handling and, therefore, personnel exposure at both TMI and the burial site leading to ALARA concerns along with the possibility of a radioactive release. The NRC staff performed an evaluation in October of 1981, at the request of GPU Nuclear, to determine the Sr 90 concentration limit for an unstabilized EPICOR liner that would be acceptable for burial at the Hanford site. The results of the NRC's evaluation show that a concentration limit of 24 uCi/cc of Sr 90 would be acceptable for waste to be considered Class "A" waste under the criteria used to develop the limits in 10 CFR Part 61. A copy of the NRC's evaluation is enclosed for your information.

The limits expressed in 10 CFR 61 are for the burial of Class "A" waste at a humid site and at normal burial depths, less than three (3) meters. Provisions for exemptions from specific limits are provided for within 10 CFR Part 61 if the performance objectives can be met by consideration of options such as burial at an arid site and at a depth greater than five (5) meters. Based on the NRC's analysis, GPU Nuclear is requesting a variance to allow a 1 uCilcc limit on Sr 90 as the upper Class "A" limit for TMI EPICOR II waste. All other Table 2, Column 1 limits would remain the same. In addition, the liners would be requested to be buried at the bottom of the disposal trench. It is our belief that this variance would be granted without any adverse effect on the health and safety of the public. GPU Nuclear believes that this variance would be in compliance with the full intent of 10 CFRPart 61. If you have any questions, please contact Mr. J. J. Byrne of my staff. BKK/JJB/jep Enclosure CC: Mr. L. H. Dr. B. J. Sincerely, Is/ J. J. Barton for B. K. Kanga Director, TMI-2 Barrett, Deputy Program Director -TMI Program Office Snyder, Program Director -TMI Program Office

'. .. ** (, UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON.

D. C.1055S Mr. J. Barton Acting D1 rector of TMI-2 Metropolitan Edison eompany P. O. Box 480 Middletown, PA '7057

Dear Mr. Barton:

October 22. 1981 This is in response to Mr. Hovey'S letter LL2-81-0214 of September 11.1981. concerning the use of EPICOR-Ilfor SOS effluent polishing, which included Metropolitan Edison's plans fOr EPICOR-II liner radioisotope loading and disposal.

In that letter. Met-Ed proposed to load the EPICOR-II liners to a IDIximum concentration of 1 ut./cc of isotopes with hal f"lives greater than five years and dispose of the liners (with resins in a dewatered, but fied fonn) at the bottom of a disposal trench (approximately 10 meters deep). Even though not specifically stated, we understand that Met-Ed is proposing to dispose of the EPlCOR-II liners at an arid disposal facility.

Prior to final promul gation of Part 61. your proposal would be allowable under current NRC regul ations. Subsequent to final promulgation of 10 CFR 61, the remaining waste covered by your proposal would require an exception to the Sr 90 concentration limit (0.04 uc/cc) in Table 1 for Class A waste if the regulation is approved as proposed by the staff * . The NRC staff has performed an eval uation of the waste and di sposa' condi tions proposed by Met-Ed. The evaluation indicates that the proposed conditions would be acceptable fOr the waste to be considered a Class A unstabilized waste under 10 CFR 61, provided all other requirements of the proposed 10 eFR 61 for Class A wastes were met (e.g., the waste is segregated from Class Band C stabilized wastes and disposed of in a separate trench). Since the existing commercial disposal Sites are regulated by the individual States, acceptability of the waste form and disposal conditions would rest with them. However. it is our position that we would recommend acceptance of your proposal.

( ( P r,,:;: 4-":

V J

.'1 J ()

• Mr. John J. Barton . . It is requested that 10U continue ¥Our careful anal¥t1cal program to determine the content of tbese. isotopes in the various waste containers to ensure conformance with the dis,osal criteria discussed above. . cc: See Service Distribution List Sincerely.

Bernard J.

TMI Program Office . Office of Nuclear Reactor ReQulat10n

. .. ( ( 1 "Acceptability for Disposal of UDatabilized TMl-2 Devatered leain 90 .. " *Waates Iavins Sr Concentrationa Creater thaD 0.04 uc/cc Purpoae: 1'be JlUrpoae of tb1.a evaluation 18 to 4etel'll1lle the acceptabi90tr of d1apos1Da of uaatabllized tKI-2 4evatered realn .. stes Sr concentrationa areater than 0.04 uc/cc.-the upper limit for Sr concentratlona for Claaa A .. atea apecified in the proposed 10 cn 61. leferences:

1. Proposed rule, 10 en. 61, Licensing Requirements for Land Disposal of Radioactive Waate. Federal legister, Vol. 46, No. 142. July 24, 1981, pp. 38081 -38105. 2. Draft InvirOlllDental Impact Statement on 10 CFR. 61 "Licenaing _nts for Land Disposal of Radioactive Wute," N'OREG-0782, Appendix C. 3. INVERSI code run, .June 12. 1981. Results: Disposal of tKI-2 4ewatered resin wastes having Sr 90 concentrations less than 24 uc/cc WDuld be acceptable for d1sposal in an unatablllzed condltion at depths greater than 5 at an arid disposal site. If other isotopes 11sted in Table 1 of the proposed 10 CFR 61 are also preaent, these isotopes would also need to be accounted for using the concentration ratio factor identified in Table 1. Evaluation:

The proposed rule for low-level vaste unagement, 10 cn 61, I includes a .. ate classification syatem (Reference 1). The upper co\Sentration limit for the dispoaal of unstabUized waates (Class A) fo: I Sr is liven .. 0.04 uc/cc. This limit 8. cletermined by evalusting th' effecta of intrucler pathways at a reference disposal facUi ty. The intrucler pathways included construction and agricultural caaes. The *

. . .. ( ( . . . t I . '-2 draft envirODMntal iapact atat_nt for 10 en 61 (bference

2) provides a detailed deacription of th .. e pathv.ya.

!be allowable CODcentrationa for the intruder pathway .va1uationa in the ... te ca1aaification

.,at .. are baaed on a performance objective that the intrqcler riceives an annual doae to the whole t.ocJy of le.s than 500 _em. !be ... te c1asaification

&yat.. 111 Raference 1 require. that .. stes buried at DOnaal depths (111c1ucla.

diW.al at less than 3 _ters) at either humid or arid ait .. havtDa Sr concentrations Iraater than 0.04 uc/cc be atabilb.d.

Iovever, 10 en 61 does provide for exemptions if the .pecific d18poaal condit1Dn.

provide asaurance that the performance objectives are _to In .va1uatina certain options Which could provide the aasurance that the performance objectiv .. are at, aeveral 90 alternatives could be considered for unatabil1zed

.. stes with Sr concentrations Ireater than 0.04 uc/cc. !bese alternatives include: burial at depths araater than 5 _tera (that 1a, with an intruder barrier), burial at an arid aite, or a combination of these. Because the propoaed vate vou1d be unatabilized, the .stes vou1d be disposed of in a trench Cla.s A .stes. Cla.s A .. stes would be aeareaated from the atabilized Cl .. s I and C ... tes. The basic assumptio118 111 the Class A .. ste acenarios for DOraa1 depths and deeper depths (Iraater than S _ter.) are .. fo11ovs: 1. The reference eli.poaal aite is located in a humid Southeastern aite. 2. Inadvertent intrusion 18 .. de after institutional control is lost fo1lowins an active control period of 100 years. 3. At the time of intruaion the vastes have dearaded to the extent that they are unrecognizable as vaste and undistinauishable from aoU. 4 *. '!'he vaste delradation takes place at a rate independent of aite location.

That 1a,* the dearadation 18 the aame for an arid and a humid aite. S. Aaricultural activities occur only in .. stes located less than 3 _tera below &raele. 1'h18 18 baaed on the construction of a residence v:1th a basement excavated to 3 _terse The aoils

,. ( ( *. .,.----. "... .. 3 removed for the buUdaa are araded about the ru1clence and .fooc:la are &rOVD in the excavated aoila. 6. Cou truc tion .enta aorully take place at depths l.s. t1wl 3 _tera. 7. When deep disposal 18 .. suaed, it 18 judaed less U.ltely that a1pificant coutruction vU1 tab place at these depths (h1Jh rise buUdi11J c0118truction, for example).

For .stes thus disposed, it 18 .. auaed that o111y 10 percent of the vaates are contacted and become available for dispersion into the air and eubsequent tahalati011 b7 humans. Purther. potential direct aamma exposures from vork1na on homoaeneouely contaminated around are .. sused to be reduced by a factor equal to one aeter of eoU ahieldiDJ (1/1200). . With these basic .. sumptions the allowable Sr 90 concentrations for the stated options were computed ueina the INVER.SI code which was also used to determine the l1m1tinJ radionuclide concentrati0118 for the 10 CFR 61 waste classification system (Reference 3). The results are provided in Table 1. Table 1 Allowable Sr 90 Concentrati0118 for U118tablized Wastes Option UUstabilized waste, regular disposal (normal depths) UUstabilized Maste, burial at depths areater than 5 _ters Allowable Concentration, Construction ScenariO, uc/cc 2.0 24 Allowable Concentration, Aaricultural Scenario, uc/cc 0.04

• NA

• Agricultural activities are not .. sumed to take place for wastes disposed at depths areater than 5 _terse

( '.... . -.. , .... , A . .:, Since the di.po.al effect. for aD aric! ancl a llumie! .ite are .. sumee! to be the _e, the allowble COIlcentratlou 1IDule! .. the,eame..

Bowever, tbe-a'bove.*lIYaluafi01l baa coDaidereel 01&1, .the isotope II' aDd baa aot 14 IIYtg.U&ted effect. of other I1111t1ll&

loq-liveel isoto,.. web .. C , 1'c

• or I which 1I1&ht 'be pre ** t in a ** te of thia .. ture. these l8oto,.. have hlah pot.ntials at buaid aite. but are aenerally . IlOt .pecifically

.... ur.d at power plut. clue. to low c01lc.ntratioua and analytic coapl.zity.

Allovina dilpolal of hlaher activity UJ18tabUized wstes at humid elilpo.al lit.1 could raeult ill iIlcr .... d Iroundwater aigration of such 11.m1tiq lOllI-lived .cbUe uotopes as vell &s increased post operational

_iIlteuance co.ts. Sillce it ia possible that tMl-2 wastes aiaht also contain .ome of these louaer-lived ilotopes in concentrationa Ileal' their Cl ... A 11a1t., it 11 judaed to 'be prudent to dispo.e of such h1&her activity UDatabUized

.. It ** at an arid aite where it can be ueumed that II1&ration 11 DOt a .1an1ficant pathway. This evaluation, therefore, c01lc;Udu that clilposal of uutabilized

'l'MI-2 dewatered resin wastes havina Sr c011centratioll8 up to 24 uc/cc would be acceptable provided the waat .... re buried at depths area tel' than S _tel'S at &11 arid dispo.al .ite. Other 1Iotopes luted ill Table 1 of leference 1, of .cour ** , would a.ed to .. accouuteel for ua1na the concentration ratio factor ic!entified in Table 1. Evaluation perfomecl by Date ,..JI!-I

__

Approved by: i '\ .. *

• Date __ _ /

REN RAHM )ecretary 1111 t::!/\ M.ETI rllll;X f&1 \

ffI.rJN. * --.:D'IIl!

SeA III: ** 1 -STATE Of WASHINGTON DEPARTMENT OF SOCIAL AND HEALTH SERVICES GPU Nuclear Corporation P.O. Box 480 Route.441 South OI)lmpia.

Washington 985().4 July 17, 1985 -\cgO Middletown, Pennsylvania 17057-0191 Attention:

B.K. Kanga

Dear Mr. Kanga:

• • '" r r

• r t .. .

• E. ,;:r:u -.., This letter is in response to the variance request contained in your letter of October 26, 1983. The US Ecology Radioactive Materials License WN-IOI9-2.

Condition (27)(j) requires ion exchange resins containing radioactive material having a total specific activity of 1 uCi/cc or greater of materials with half-lives greater than 5 years must be stabilized by solidification or be placed in. a high integrity container and shall contain no detectable free-standing liquids. Your request for a variance to the requirements is hereby granted provided that the following conditions are met: A. Sr-90 are not to exceed 1 uCi/cc. B. Wastes will comply with Class A waste requirements specified in 10 CFR 61.56. c. Wastes are disposed of at the bottom of the trench. D. Wastes are segregated from stable Class Band C wastes. E. Wastes do not contain other radionuclides listed in Tables 1 and 2 of 10 CFR Part 61.55 which exceed the Class A limits by themselves or giving consideration to the partial fractions rule. C * 'v' ! r.

, ';' GPU Nuclear Corporation July 17. 1985 Page 2 F. A copy of this variance approval must accompany the required shipping papers to the US Ecology Handford Disposal Site. If we can be of further assistance, please feel free to call. JS:MJE:pm Sincerely, a..,JU. Stohr, Manager Radioactive Waste Program

/-c A-.. Mikel J. Elsen Radiation Health Physicist Radiation Control Section Olympia. Washington 98504 cc: US Ecology -Louisville, KY US Ecology -Richalnd, WA Bob Bidstrup -DSHS Kathy Schneider

-USNRC

I ' ( ( UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON.

D. C. 20555 Gentlemen:

6771 for the contents:

Up to 1,500 curies of radioactive solids including 60 grams fissile material and 40 curies of plutonium.

The material must be packaged in accordance with GPU Nuclear Corporation application dated August 27, 1984. One package per shipment as Fissile Class III. All other conditions of Certificate of Compliance No. 6771 shall remain the same. This authorization shall expire December 31, 1984.

Enclosure:

Approval Record cc w/enc1: Mr. Richard R. Rawl Department of Transportation FOR THE U.S. NUCLEAR REGULATORY COMMISSION ,e,. "Transportation Certification Branch Division of Fuel Cycle and Material Safety, NMSS

. l UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON.

D. C. 20555 P,\GE L; 1 jOiJ c:, Transportation Certification Branch Approval Record Model No. SN-l Package Docket No. 71-6771 By application dated August 27, 1984, GPU Nuclear Corporation requested additional contents in the Model No. SN-l package. The contents consist of up to 1,500 curies of solid radioactive waste containing up to 60 grams fissile material and 40 curies of plutonium.

The package is currently approved for up to 5,000 curies of (dewatered) solid radioactive waste. The fissile material is well subcritical and the certificate has been conditioned to require the shipment to be made as Fissile Class III. The waste material is divided between 20 canisters and positioned within the package in a lead shielded basket. The canisters are fabricated from 3.75" 00 aluminum tubing with a wall thickness of 1/4" and approximately 36" long. The canisters have a 5/16" thick bottom welded to the tube and a 3" NPT threaded pipe plug for a top closure. The lead shielded basket has about 4" of lead on the bottom and sides and 5-inch thick plugs installed above each canister.

The basket has 20 compartments to accept each canister.

The basket plus l-inch thick lead shielding on the sides is placed within an additional sealed stainless steel container prior to placement into the Model No. SN-l package (cask). The cask is sealed a silicone rubber O-ring. Based on the distribution of the waste material (averaging less than 2 curies per canister, multiple packaging, and the form of the material (mainly metallic), the plutonium bearing solid has been exempted from the requirements of 10 CFR §7l.63. The NRC staff is in agreement with that applicant that this one-time shipment of waste material would have no adverse effect on the health and safety of the public. Da te :_O_C_T_2_9_19...;8;..:..4_

Transportation Certification Branch Division of Fuel Cycle and Material Safety, NMSS E..:.J Nuclear Office of Nuclear Material Safety and Safeguards Attn: Mr. C. E. MacDonald Chief Transportation Certification Branch US Nuclear Regulatory Commission Washington, DC 20555

Dear Mr. MacDonald:

GPU Nucl ** r Corporation Post Office Box 480 Route 441 South Middletown, Pennsylvania 17057-0191 717 944-7621 TELEX 84-2386 _ Writer's Direct Dial Number: (717) 948-8461 44lo-84-L-0142 Document 10 0062A August 27, 1984 lit bitt: Three Mile Island Nuclear Station, Unit 2 (TMI-2) ..... r: r .. r. j,i. ""_. . d-. Operating License No. DPR-73 Docket No. 50-320 10 CFR 71.7 Requirements c-.... :

Ac.o,-..... 000-.... ln, rx.. .a ... ,..,.:-r.-

.. , -,

."'l t "'Ie: .a .. *

""r. .... :. -, a"'l:>f!'r t: CIa::.:>'" . ,,-. !" 'r .. :, 6e ..

or* r.'". ':50",". 0"':..-" -v. :. *. &. PI: ** *

-.. "'. c!-. 1i!o. ..... ..... In accordance with the requirements of 10 CFR 71.7, it is requested that the Nuclear Regulatory Commission authorize a one-time use for the Model SN-l radioactive material shipping cask to transport a small quantity of fissile material.

The NRC Certificate of Compliance, No. 6771, for this cask limits the use of the container to greater than Type A quantities but less than 5,000 curies per package of radioactive material of non-fissile classification in leak-tight secondary containers.

This one-time use is being requested to use the container to ship greater than Type A quantities of radioactive materials, but less than 5,000 curies, containing small quantities of fissile materials.

The total curie content of the radioactive waste shipment is 1,460.64 curies and is packaged in a DOT Type 7A-container with the Type B shipping cask. The weight of the loaded shipping cask is 58,784 lbs. which is less than the 60,000 Ibs. maximum allowable weight for the Type B shipping cask. The quantity of fissile material in the package to be shipped is 54.14 grams of UranillTl and PlutonillTl.

The UranillTl and PlutonillTl materials, even if arranged in the worst-case configuration, would still not present a criticality concern. ( -GPU Nuclear Corporation is a subsidiary of the General Public Utilities CorporatIon

( ( ,. Mr. C. August 27, 1984 441D-84-L-0142 This one-time shipment is being made by General Electric Company for the United States Department of Energy (US DOE), Knolls Atomic Power Laboratory in Schenectady, New York, to the US DOE Oak Ridge National Laboratory in Tennessee.

This shipment will consist of a total shipping activity of 1,460.64 curies, which will include curies of Plutonium.

This shipment will include 54.14 grams of fissile material.

This shipment will be classified as a Highway Route Controlled Quantity in accordance with Department of Transportation regulations.

The radioactive materials are metallic and non-metallic material composed of irradiated fuel, structural material, and miscellaneous contaminated items. This shipment will contain no hazardous materials other than radioactive.

The radioisotopic distribution and the estimated activity of each is as follows: Curies 8a 137 Curies Sr 90 Curies y 90 321 Curies Pm 147 17 Curies Cs 134 59 Curies Eu 154 26 Curies Co 60 47 Curies Cm 244 1.8 Curies Pu 238 26.3 Curies (l.5 grams) Pu 241 8.3 Curies (0.08 grams) Am241 0.18 Curies Other MFP 0.06 Curies u 235 0.0001 Curies (52.56 grams) TOTAL: 1,460.64 Curies (54.14 grams Fissile) The material is contained in twenty (20) sealed canisters fabricated from 3.75 inch DO 6061-T6 aluminum alloy tubing with a wall thickness of 1/4 inch, approximately thirty-six (36) inches long. The canisters have a 5/16 inch thick bottom welded to the tube and a 3.00 NPT threaded pipe plug which is provided with a lifting attachment on the top end. A more detailed description of the canister is contained on the attached KAPL Drawing 131C6820.

The waste filled canisters are packed in a lead shielded storage container which has twenty (20) individual square Document 10 0062A

(, l *Mr. C. r>,\GE ; l1Qfj August 27, 1984 4410-84-L-0142 compartments to accept each canister.

The container has approximately four (4) inches of lead on the bottom and sides and five (5) inch thick plugs which are installed above each canister.

The approximate overall dimensions of the shielded container are forty-three (43) inches long by 38.8 inches wide by sixty-two (62) inches high. The canisters and upper shield plugs are held down by a cover plate secured in place by sixteen (16) 1/2 inch bolts. A more detailed description of the shielded container is shown on the attached KAPL Drawing 296E182. An additional one (1) inch of lead was added to the sides of the container to minimize exposure to personnel during handling operations.

The lead is attached to the container in a manner such that it will not become displaced during shipping and handling operations and is shown in detail on Sheet 2 of the attached KAPL Drawing 296E182 and on Applied Engineering Drawing 0-300-101-02.

This shielded container and the additional shielding will be packed into a DOT Spec 7A, Type A container made of stainless steel shown on attached Applied Engineering Drawings 0-300-101-01, 0-300-101-02, and 0-300-101-03.

A certification, as authorized by 49 CFR, by Applied Engineering Company, showing compliance with Type A packaging requirements is also included.

This Type 7A container will be placed into the Model SN-l Radioactive Material Shipping Cask. KAPL Drawing 901E523 shows the placement of the Type 7A container in the shipping cask. The distribution of the fissile material contained within the shielded container is shown on Sketch 1 dated November 4, 1983. The sketch shows the location of each canister within the shielded container and the quantity of fissile materials in grams contained in each canister.

As you will note, fifteen (15) of the twenty (20) canisters, which are 6.51 liters in volume, meet the provision in 10 CFR 71.53(e) and contain less than five (5) grams fissile per ten (10) liter volume. The Plutonium quantity meets the provision as stated in 10 CFR 71.53(f) with the Pu 239 and Pu 241 concentration of five (5) percent of the Plutonium mass. The package, as discussed above, contains subcritical quantities of fissile materials.

The "Nuclear Safety Guide", TI0-7016, Revision 2, June 1974, specifies 760 grams of u2 35 and 510 grams Qf Pu as being the subcritical limits. The quantity of 52.56 grams of UL35 annd 1.5 grams of Pu 238 contained in this shipment, even if arranged in the worst-case configuration, would not present a criticality concern. Therefore, the quantity of fissile material does not pose a criticality concern. None of the fissile mass quantity limits in 10 CFR 71.22 are exceeded.

It is the belief of the licensee that this one-time use would cause no adverse effect on the health and safety of the public and should, therefore, be granted. Document 10 0062A

( . C. J 1 JurJ \.0 August 27, 1984 4410-84-L-0142 Pursuant to the requirements of 10 CFR 170, enclosed is a check for $150.00 for NRC review and approval of this request. If you have any questions concerning this information, please call Mr. J. J. Byrne of my staff. FRS/RBS/jep Attachments Sincerely, /s/ F. R. Standerfer F. R. Standerfer Director, TM1-2 cc: Program Director -TMI Program Office, Dr. B. J. Snyder Acting Deputy Program Director -TMI Program Office, Dr. W. D. Travers Document 10 0062A

"""'ITATII MueUAR REGULATORY COMMISSION WAIMINGTON.

D. Co _ December 19. 1984 Docket No. 50-320 --Mr. F. R. Standerfer, Director Three Mile Island Unit 2 GPU Nuclear Corporation P.O. Box 480 Middletown.

PA 17057

Subject:

Three Mile Island Nuclear Station. Unit 2 Operating License No. DPR-73 Docket No. 50-320 Technical Specification Change Requests 39.41. 43 Recovery Operations Plan Change Requests 19. 20. 22 Exemption Request from 10 eFR 50.55a (Code Safety Valves) Exemption Request from 10 eFR 100. Append1.x....A and 10 eFR 50.36(3) (Seismic Instrumentation) . The Nuclear Regulatory Commission has issued the enclosed Amendment of Order; Recovery Operations Plan Change Approval of Exemption from the requirements of 10 eFR 50.55a for e Safety Valves; and Approval of Exemption from the seismic instrumentation requirements of 10 CFR 100, Appendix A. and 10 eFR 50.36(3).

The Amendment of Order which modifies many sections of the Proposed Technical Specifications (P15) was requested by General Public Utilities Nuclear Corporation (GPUNC) in letters dated January 12, 1983, September 12, 1983 and September

30. 1983. Other documents related to this request include: Recovery Operations Plan (ROP) Changes which were requested in separate letters also dated January 12. September

12. and September 30, 1983; a request for Ixemption from the llents of 10 CFR 50.55a with respect to Code Safety Relief Valves in a litter dated April 18, 1984; and a request for an exemption from. the seismic monitoring requirements of 10 eFR 50.36(3) and 10 CFR 100. Appendix A. Paragraph VI(aH3) in a letter dated April 18, 1984. . As previously explained 1n a litter issued by the staff on JUly 17. 1984. your _ PTS and ROP change requests were divided into separate issuancls.

The first issuance .s .. de on July 17. 1984 and .s fllll1ediately effective.

The staff has reviewed your* safety evaluations for the above docll1lents and concludes that your requests addressed by this issuance are acceptable with changes as discussed With your staff. PTS changes that are the subject of this litter w111 became effective on January 7, 1985. The Exemptions to 10 CFR 50.5Sa, 10 CFR 50.36(3) and 10 eFR 100. Appendix A, Paragraph are Iffective upon issuance.

..

( . . \ Mr. F. Since the February 11, 1980 Order i.posing the Proposed Technical cations is currently pending before the Atomic Safety and Licensing Board, the staff will be advising the Licensing Board of this Amendment of Order through a Notice of Issuance of Amendment of Order and a Motion to Confonn Proposed Technical Specifications in Accordance Therewith.

-Federal Register Notices for the discussed i$suances are enclosed.

Copies of the related Safety Evaluation and revised pa,es for the Proposed Technical Specifications and the Recovery Operations P an are also enclosed.

Enclosures*

1. Amendment of Order 2. Safety Evaluation

3. Proposed Technical Specification Page Changes Sincerely, --t!r 4'A J D.Jr... Bernard J. Snfl,e'r, Program Director Three Mile Island Program Office Office of Nuclear Reactor Regulation

4. Recovery Operations Plan Change Pages 5. Exemption from 10 CFR 50.55a 6. Exemption from 10 CFR 100. Appendix A, Paragraph VI(a)(3) and 10 CFR 50.36(a) 7. Notice of Environmental Assessment and Finding of No Significant Impact S. Federa' Register Notices cc: J. Barton R. Rogan S. Levin R.. Freemennan J. Byrne . Service Distribution List (see attached) l \ UNITED STATES NUCLEAR REGULATORY COMMISSION In the Mltter of GENERAL PUBLIC UTILITIES WUCLEAR CORPORATION

-(Tffiree Mile Island Nuclear Station, Unit 2) EXEMPTION I. Docket Wo. 50-320 Enclosure 6 6PU Nuclear Corporation, Metropolitan Edison CCllPlny, Jersey Central Powr Ind Light Company and Pennsylvlnia Electric Call1pany (collectively, the licensee) are the holders of Facility Operating License Wo*. DPR-73, "'ich had luthorized operation of the Three Mile Island Nuclear Station, Unit 2 (TMI-2) at powr levels up to 2772 atgatetts thet'llal.

The flcility, "'1ch is located in Londonderry Township, Dauphin County, Pennsyl"an1a, is a pressurized

.. ter reactor previously used fOr the commercial generltion of electricity

• . By Order fOr Modification of License, dated July 20, 1979, the 11censee's authority to operate the facilfty .. 5 suspended Ind the lfcenseels luthority was limited to maintenance of the facilfty fn the present sllltdown cooling mode (44 Fed. Reg. 45271). By further Order of the Director, Office of Nucl ear Relctor Regulation, dated February 11. 1980. I new set of fOmal license requirtnents

.5 i_posed to reflect the post-accident condition of the facility and to Issure the continued

.. intenance of the current slfe. stable, long-tenn cooling condition of the facility (45 Fed. Reg; 11292). This license prOVides, among other things, that it is subject to 111 rules. regulations Ind Orders of the Caamission now or hereafter in effect.

l \ \ -II. In I litter dated April 18, 1984, the liclnsee requested In IX-.pUon fran -the of 10 CFR 50 relative to Seismic MOnitorfng Instrumentation.

10 CFR 50 50.36(c)(3) rlquires surveillince regufrements

• • .** to Issure that the necessary qual fty of systems Ind cCllllponents is Nintained, that flcf1ity operation will be .rtthin the safety lfmits, Ind thlt the conditions of operation .rtll be met.* 10 CFR 100, Appendfx A, Section Vl(I)(3) stites that: *Suitable instrumentation sha" be provided so that the seismic response . of power plant features important to safety Cln be detemined promptly to pemit C(lTlparison of such response wi th that used IS the design basis. Such a C(lTlparison is needed to decide Whether the plant can be operated safely Ind. to pemit such timely Iction IS .. y be appropriate.

These criteria do not Iddress the need for instrumentation_thlt would matically shut down a nuclear power plant an earthquake occurs Which Ixceeds a predetemined intensity

• Presently.

Section 4.3.3.3.1 of the TMl-2 PTS requires thlt Transaxia' Time -History Accelographs be operable for the Reactor Building Ring Girder Ind the Reactor Building Mati that Triaxial Peak Accelographs be operable for the Reactor Service Structure.

aB-Core Flood Tlnk Piping Ind 2-1£ gear. that Triaxial Seismic Switches be operable for the Reactor Building Base and that Response -Spectnn Recorders be oper"ble for -the Reactor Building Mat.

( \ 111. The 1MI-2 core is cooled via loss of heat to reactor bullding .. envJ.roment

• . This 15 a passfve mode does not require any mechanical equipment to be operating to .intain an abllity to cool the core. As stated in 10 eFR 100, Appendb A, Section Y1(a)(3), one of the reasons for seismic instrumentation is to decide Whether or not the plant can be operated safely. In the July 20, 1979 Order for Modification of License, the authority to operate the facility suspended and the licensee's authority was limited to the maintenance of the facility in the present shutdown ing Therefore this basis for Section YI(a)(3) does not apply to 1MI-2. In reference to the seismic instrumentation providing infonnation for timely actions by plant personnel and the NRC, it is the staff's that if a seismic event were to occur at 1MI, the status of the core 'l)uld not be affected because of the passive cooling lede and therefore no immediate actions would have to be taken to maintain the health and safety of public. It is also the staff's opinion that when considering the above discussion, maintenance and surveillance for seismic mentation is also not justified and is an unnecessary burden on the licensee.

Because of the suspension of the licensee's authority to operate the facility in other than the present recovery lede as defined in the proposed technical specifications, of the regulations, which are intended tQ apply to normal operating plants, are simply inappropriate and, .are significantly, are unnecessary to protect the public health and safety. Given the unique ...

. l . . -. -status of the plant in t.nas of F1l11ry s1St .. t_peratur.

and pr.ssure.

avail-abl. ftssion product 1nv.ntory.

the abl1ity to cool the reactor wtthout forced -circulation (loss-to-ambi.nt).

and the low decay heat rate ... intenance of the fac11 ity "'th the .xemptions granted her.by "':11 provide an adequate l.vel of safety. IV

• Accordingly.-

the eClTl'llission has detennined that. pursuant to 10 eFR 50.12. an .xemption is authorized by law and will not endanger 11fe or property or the common defense and security and is in the public interest.

Based on the discussions abov ** the Commission her.by grants an exemption to the requirements of 10 eFR 50.36(c)(3) and 10 eFR 100. Appendix Ai Section VI(a)(3) relative to seismic instrumentation.

It is fUrther determined that the exemption does not authorize a change" in eff1uent types or total amounts nor an increase in power l.vel and wi' 1 not result in any significant environmental i.pact. In light of this mination and as reflected in the Environmental Assessment and Notice of Finding of No .Significant Environmental Impact prepared pursuant to 10 CFR 51.21 and *51.30 through 51.32. issued concurrently herewfth.*it "5.

( l . , concluded that the instant action is insignificant fran the stindpo1nt of environmental impact and an environmenta1 i.pact statement need not be -, of "4!e AT" . prepared.

Effective Date: December 19, 1984 Dated It Bethesda, Maryland Issuance Oate: December 19, 1984 FOR THE NUCLEAR REGULATORY COMMISSION

/,paL Hlr01 d R. Denton, Director Office of Nuclear Reactor Regulation

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c. LAte* ( . .a:iEl Nuclear . llnd*Trlr.H 11/ IV I GPU Nue .. ., Corporation Post Office Box 480 Route 44' South Middletown, Pennsylvania 17057*0191 717 944*7621 l 1MI Program Office Attn: Dr. B. J. Snyder Program Director US Nuclear Regulatory Commission Washington, DC 20555

Dear Dr. Snyder:

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• • • • c,.,,...., , : . TELEX 84*2386 Writer's Direct Dial Number: (717) 948-8461 4410-84-1..-0050 Document ID 0481U April 18, 1984 Three Mile Island Nuclear Station, Unit 2 (TMI-2) Operating License No. OPR-73 Docket No. 50-320 Seismic Monitoring Exemption Request Your letter of January 13, 1984, which provided comments on various Technical Specification and Recovery Operations Plan Change Requests required that GPUNC submit a specific relief request from the seismic monitoring requirements of 10 CFR Part 50. Based on the attached justification GPUNC requests an exemption from tne seismic monitoring requirements of 10 OFR Part 50. As this request is submitted in conjunction with Technical Specification Change Request No. 43, no additional fee is required.

Please call Mr. J. J. Byrne of my staff if you have any questions on this information.

EEK/J.:B/jep Attachment Sincerely, /sl E. E. Kintner E. E. Kintner Executive Vice President cc: Deputy Program Director -1MI Program Office, Mr. L. H. Barrett GPU Nuclear Corporation IS a subSidiary of the General Public Utilities Corporation

• c ( .lJSTlf"ICATION f"rA IELETING SEISMIC MONlTORlf<<.i f'lr.\tIU.&.N.r

...... _ <"'" c--):;" C1< \ 1\ 'O)r '-1 .,:'1 I l Paragraph (c) of Section 50.36, "Technical Specification", of 10 CFR Part 50 provides that the Technical Specifications will include surveillance requirements to assure that the necessary quality of systems and components is maintained, that facility operations

<will be within safety limits, and that the Umiting Conditions for (4leration will be met. Appendix A, "Seismic and Geologic Siting Criteria for ftJclear Power Plants", to 10 a:-R Part 100, "Reactor Site Criteria", requires in Paragraph VI{a)(3), a suitable program for this requirement with regard to seismic instrunentation needed to determine promptly the seismic response of nuclear power plant features Important to Safety to permit comparison of such response with that used as the desig1 basis. That is, the seismic monitoring instrumentation is used only to record and define actions after a seismic event. These actions consist primarily of engineering evaluations to determine damage caused by a seismic event and the repairs required prior to restart. As performing the surveillances on the seismic instrumentation would require 3 to 5 man-rem per year (Reference GAUNC Letter 4410-83-L-0151 dated July 20, 1983), and the data provided would, in general, not be needed unless a decision is made to restart TMI-2 continued performance of these surveillances is not a prudent man-rem expenditure.

Additionally, US NRC Regulatory Guide 1.12 by reference to ANSI Standard Nl8.5 provides guidance on seismic instrumentation required for multi-unit sites. Section 4.4 of ANSI Nl8.5 states that, "Instrumentation in addition to that installed for a single unit will not be required if essentially the same seismic response is expected at the other units based on the seismic analYSis used in the seismic design of the plant." Given that both units are located in close proximity and are both founded on bedrock, it is expected that the Unit 2 seismic response will closely approximate the Unit 1 response.

Therefore, given the above guidance and the Recovery Mode status of TMI-2, Technical Specification 3.3.3.3 which requires surveillance of seismic instrumentation in Unit 2 can be deleted without imposing a sig1ificant risk to the health and safety of the public and would save an exposure fo 3 to 5 man-rem per year to TMI-2 workers. OocLJnent 10 0481U

r, PAGE UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 May 25, 1984 NRC/TMI 84-035 Mr. B. K. Kanga. Director Three Mile Unit 2 MAY 1984 GPU Nuclear Corporation P. O. Box 480 4ilVllJWAlCOD

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Subject:

Reactor Coolant Purification System Ion Exchanger Wastes! I am responding to GPUNC letter 4410-84-L-0066, dated April 18, 1984, from E. Kintner to H. Denton requesting relief from the commitment to prepare the Reactor Coolant Purification System Ion Exchange resin for off-site shipment in late 1984. This commmitment is contained in GPUNC letter 4410-82-L-0026 dated October 8, 1982. The bases for your request are that deferral of resin removal from the demineralizer vessels will not present a hazard to the public health and safety and that the deferral would allow additional resources to be devoted to the reactor defueling effort. Our review of your request and discussion of the technical issues with members of your staff in a meeting on May 15, 1984, concluded the following:

1. Efforts already completed to characterize the contents of the vessels indicate that there are no chemically corrosive species present and that the proposed elution process will be able to remove 90 percent of the activity from the resins. 2. There are no hazards associated with combustible gas generation in the demineralizer vessels because of the low gas generation rate and means exist for venting of the vessels to the waste gas system. 3. Deferral of resin removal and preparation for shipment will contribute significant resources to the defueling effort. 4. Delay of shipment of the purification resins will not adversely impact future funding or planning for shipment of other abnormal wastes. 5. Delay of resin removal will not impact on any plans for future modifications to the spent resin tanks. We therefore approve your request for relief from your commitment contingent upon our approval of your Safety Evaluation Report (SER) for cesium elution. The SER wilT need to address the following technical issues to storage of the eluted resin in the demineralizer vessels until later in the plant cleanup schedule.

Hr. B. K. Kanga . -2 .. p Z--G May 25, 1984 You will proceed with *plans to elute the cesium from the resin in the demineralizer vessels and process the eluent stream. The elution process will include acceptable means to quantitatively determine the effectiveness of the process. If the 90 percent activity removal goal is not met,.you are requested to re-evaluate your purification demineralizer plans and report your conclusions to this for our review. Upon completion of the elution. the resin will remain in the demineralizer vessels in water of such a quality to assure material compatibility and a non-corrosive environment commensurate with the planned length of the storage period. You will implement approved procedures for the monitoring and maintaining of the system integrity after elution.

IJ 1,61 . v f 4 provide. operate and maintain present fire detection and protection equipment in the areas near the purification demineralizer r:..",,fe f i c 1 es. If these issues are adequately addressed.

deferral of the resin removal will not jeopardize the health and safety of the public, and the potential environmental effects from this action fall within the scope-of conditions considered in the PElS. . cc: J. Barton R. Rogan E. Kintner R. Freemerman A. Hi ller J.

..,--. J. ayrnev _. -See Servlce Distributior List / .I ! J i lake H. Barrett Deputy Program Director TMI Program Office

• t ( ,. . . . -...... Nuclear GPU N'-' eorpo.atlon Post Office Box .. 80 Route .... ' South Middletown.

Pennsylvania' 7057-0191 717 e.u*7621 TElEX 84*2386 Write'" Direct Dial Number: (717) 948-8461 11410-84-1...-0066 Document 10 0159U April 18, 1984 Office of Nuclear Reactor Regulation Attn: Mr. H. R. Denton Director US Nuclear Regulatory Commission W8shin;Jton, DC 20555

Dear Mr. Denton:

Three Mile Island NLclear Station, ltlit 2 (TMI-2) C>>eratirg license No. OPR-7:3 Docket No. 50-320 Reactor Coolant Purification System Ion Exchange Wastes GRJN:: Letter 441D-S2-L-0026 dated (ktctler B, 1982, which was provided in response to your letter of September B t 1982, provided, in part, an anticipated t1meframe for readying the Reactor Coolant Purification System Ion Excnange wastes for off-site shipment in late 1984. This letter also noted, however, that removal of these resins was largely a research and development effort as the assessnent of resin removal activities still needed to be canpleted.

Since the subnittal of 441D-82-L-0026, nJch information has been obtained corcerning these resins and has 1qlacted our previous decisions regarding their dispOSition.

Specifically:

1. Characterization of the contents of Ion B is COOlllete. (See Attact1nent for details.)

Dlaracterization of Ion Exchanger A is but the results should not be sigdf1cantly different fran B and there is very little fuel present in the Ion Exchangers.

2. ttl hazards are associated with corrDustible gas generation.

I GRJt£ Letter 441G-83-L-0227 dated Novent>er 30, 19B3, stated that hydrogen gas generation was at a rate of no greater than 0.25 liters per day. This can easily be vented to the Off Gas System. GPU Nuclear Corporation is 8 subsidiary of the General Public Utilities Corporation

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3. No chemically corrosive species have been identified in the obtained from the Ion Based on these results, the remaining source of concern with respect to these resins is the radionu=lide loading of the resins. In order to alleviate this concern, GPUNC plans to elute the cesium activity during the mid 1984 timeframe.

Laboratory scale tests on Ion Exchanger B resin samples confirm that a 90 percent elution goal can be ad1ieved and that the material eluted can be decontaminated by the SUJmerged Demineralizer System using zeolite resins similar to those for Reactor Building SlJr1:l water processing.

lFUl'£ is in the process of verifying this information on Ion Exchanger A resins but the results should be essentially similar. Additionally, access to the Reactor Coolant Purification System Ion Exchanger Roan is not required for maintenance of the unit in a safe shutdown condition; thus t no increase in worker exposure will result from continued storage of the resins after cesiLlTl elution within the ion exchangers.

Therefore, as the hazard associated with these resins is being removed via elution of the radioacti vi ty, which was not anticipated in 44lD-S2-L-D026, the need to expeditiously remove the resins is eliminated.

In addition, in support of GPUNC's program plan to utilize resources to achieve early defueling of TMI-2, without the health and safety of the publiC, and as noted above, this hazard will be eliminated by elution of the Reactor Coolant Purification System Ion Exchange Resins. GPUNC requests relief from the cormlitment contained in 441G-B2-L-0026 and, as a result, will remove these resins fran the site in conjunction with other wastes of similar characteristics at a future date as part of the normal cleanup process. Please call Mr. J. J. Byrne of my staff if you have any qLEstions on this information.

EEKI J:J3/ jep Attad'lnent Sincerely, /sl E. E. Kintner E. E. Kintner Executive Vice President cc: Program Director -TMI Program Office, Dr. B. J. Snyder OetlJty Program Director -TMI Program Office, Mr. L. H. Barrett

-"-" ." ;. Units *E1. ppm B C ppm Ha ppm Mg ppm Al ppm Si ppm P ppm S04 ppm C1 ppm K ppm Ca ppm V ppm Cr ppm Mn ppm La ppm Ba ppm Cs* ppm Te* ppm Sn ppm In ppm Cd ppm Ag ppm Rh ppm Nb ppm Zr ppm Sr* ppm Rb* ppm As ppm Zn ppm Cu ppm Ni ppm Co ppm Fe ppm U ppb Pu JolC; Ig 134Cs JolCi/g 137Cs JolCi/g 90Sr pH *Flss10n Product Attachment 1 4410-8.4-L-0066 TABLE III P 'GE 5 11 !OJfl (, Analytical Results for Coolant Purification System Ion Exchanger B

1983 1983 Solution Sol1d Uquid Sol1d 3000 3000 )200 1000 ppm 900 ppm )101. 7000 -10.000 )1000 <1 <1 2 10 10 70 <3 <3 <5 O. 1 0.1 <1 9600 6000 15,000 5 20 30 3 0.8 4 20 10 30 1 0.3 0.6 5 0.2 0.1 5 3 <1 30 30 100 <1 <1 30 -2 0.2 0.3 30 <1 <1 60 0.4 2 30 <3 <.1 <.1 <1 1 1 6 1 <1 1 6 4 15 <.5 <.5 <1 <.2 0.2 <1 1 0.3 <1 0.5 0.6 40 <.1 <.1 <1 10 10 200 0.064 1620 0.109 283 0.72 3550 0.64 787 0.181E3 0.778E3 0.101 E3 1. 13E3 2.64E3 l'.2E3 1.48E3 16.9E3 0.014E3 0.49E3 9.46 0.88E3 5.7 5.3

.. I . ," . . , TABLE III (Continued)

ARrt 1 UnHs E1. B-Solution At 1 234U At 'I. 23SU At 1 236U At 1 238U At 1 238Pu At 1 239Pu At 1 240Pu At 'I. 241Pu At 1 242Pu Original Sample R at contact 0.022 2.23 0.128 97.62 <0.07 87.85 1-0.29 1. 79 <.05 Insoluble Residue -20 mg/ml **Sma 11 ali quot. .... 1983 B-Solid 0.023 2.46 0.072 97.45 <.05 91.0 7.6 1.4 <.05 I 4410-84-L-0066 pl\GE (p Or. Mai: 1983 B-2 Solid B-2 Ug. 0.023** .026 2.5** 2.17 -.1 ** 0.10 97.5** 97.70 <.1 <. 1 84.4 82.88 13.8 13.25 1.82 3.87 <.1 <.1 (110 L1q. and Solid) -500 mg/ml I ."

. UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. c._ Docket Mo. 50-320 f: Mr. F. R. Sunderfer Vice President/Director Three Mile Island Unit 2 6PU Nuclear Corporation P.O. Box 480 Middletown, PA 17057

Dear Mr. Sunderfer:

May 16, 1985

Subject:

Three Mile Island Nuclear Station, Unit 2 Operating license No. DPR-73 Docket No. 50-320 -t.=._. ___ _ -.:11:1', ? .. !tmt.y 6r.£MrT'i'J?" ,0 d"1l.. S"'O *S4 (.) _ ..

__ _ -... 1.....------

Partial Exemption from the Requirements of 10 eFR 50.54(a) We have reviewed lour request for a partial Exemption from the requirements of 10 eFR 50.54(a} dated April II, 1983. We conclude that your request is acceptable as stated in our attached Exemption issued by the Director of Nuclear Reactor Regulation.

A Federal Register notice for this issuance is also enclosed.

Enclosures:

1. Exemption

2. Environmental Assessment and Motice of Finding of No Significant Environmental

3. Federal Register Motice cc: T. F. Demmitt R. E. Rogan S. Levin W. H. Linton .1. d. Byrne A. W. M11ler Service Distribution list (see attached)

Sincerely. --If) .. -rna :l';' s',.Ae'r, Program Di rector Three Mile Island Program Office Office of Nuclear Reactor Regulation I ,. Iaapact

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Enclosure 1 UNITED STATES NUCLEAR REGULATORY COMMISSION In the. Matter of UTILITIES NUCLEAR (Three Mile Island Nuclear Station, Unit 2) EXEMPTION

1. Docket No. 50-320 GPU Nuclear Corporation, Metropolitan Edison Company, Jersey Central Power and Light Company and Pennsylvania Electric Company (collectively, the licensee) are the holders of Facility Operating License No. DPR-73. The facility, which is located in Londonderry Township, Dauphin County, Pennsylvania, is a pressurized water reactor previously used for the commercial generation of electricity.

By Order for Modification of License, dated July 20, 1979, the licensee's authority to operate the facility was suspended and the licensee's authority was limited to maintenance of the facility in the present shutdown cooling mode (44 Fed. Reg. 45271). 8y further Order of the Director, Office of Nuclear Reactor R,gulation, dated February 11, 1980, a new set of formal license requirements was imposed to reflect the post-accident condition of the facility and to assure the continued maintenance of the current safe, stable, long-term cooling condition of the facility (45 Fed. Reg. 11292). This license provides, among other things, that it is subject to all rules, regulations and Orders of the Commission now or hereafter in effect. . t

• . 11. On 1983, General Public Utilities Nuclear Corporation (GPUNC) submitted Revision 2 to their Recovery Quality Assurance Plan (RQAP) for Three Mile Island, Unit 2. In the letter accompanying the revised plan, they also requested a partial exemption from the update requirements of 50.54(a).

The staff to the 1983 letter on October 17, 1983 but because a separate exemption had to be issued, the NRC did not address the partial exemption request in that correspondence.

On April 17, 1984, GPUNC submitted Revision 3 to the RQAP which was approved by the staff on June IS, 1984. The partial exemption was still under staff review and as a result was also not addressed in the latter correspondence.

Therefore, the staff is now issuing a partial exemption as discussed herein. 111. 10 CFR 50.54(a) requires that an update to the Quality Assurance Program, as described in the Safety Analysis Report (SAR), be provided to the priate NRC regional Office for inclusion in the SAR. The licensee's submittal of April II, 1983 and April '17, 1984, satisfied the regulatory requirement for submittals; however, the plan was not incorporated into the TMI-2 FSAR. In a letter dated February 4, 1982, the staff exempted GPUNt from the requirements of 10 CFR 50.71(e) relative to FSAR updates. In lieu of this regulation the licensee was required to update certain System Descriptions and Technical Evaluation Reports on an annual basis. Therefore the 1141-2 << l FSAR is ir longer a current document.

Since the February 1982 exemption relieved the licensee from Iny FSAR updating requirements, it is able to also exempt the licensee from 10 CFR 50.54(a) FSAR updating requirements relative to QA program revisions.

Therefore, the stiff is exempting GPUNC from the requirement to submit revised FSAR pages whenever the QA program is modified.

However, whenever the licensee's QA program description commitments are reduced, the modified program with modified pages must still be submitted to the NRC who will still approve the changes prior to implementation.

The exemption from submitting FSAR pages does not affect the level of Quality Assurance at TMI-2 since all other regulatory requirements of 10 CFR 50.54 remain in effect. IV. Accordingly, the Commission has determined that, pursuant to 10 CFR 50.12, an exemption is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest.

The Commission hereby grants an exemption to the requirements of 10 CFR i 50.54(a) with respect to incorporating updated QA plans into the TMI-2 FSAR. It is further determined that the exemption does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. In light of this mination and as reflected in the Environmental Assessment and Notice of Finding of No Significant Environmental Impact prepared pursuant to 10 CFR 51.21 and 51.30 through 51.32, issued concurrently herewith, it was

• , concluded*

that the instant action is insignificant from the standpoint of environmJhtal impact and an environmental impact statement need not be prepared.

Effective Date: June 24, 1985 Dated at Bethesda, Maryland Issuance Date: May 16. 1985 FOR THE NUCLEAR REGULATORY COtt1ISSION Harold R. Denton, Director Office of Nuclear Reactor Regulation r' Enclosure 2 UNITED STATES NUCLEAR REGULATORY COMMISSION GENERAL PUBLIC UTILITIES NUCLEAR CORPORATION DOCKET NO. 50-320 ENVIRONMENTAL ASSESSMENT AND NOTICE OF FINDING OF NO SIGNIFICANT ENVIRONMENTAL IMPACT The U.S. Nuclear Regulatory Commission (the Commission) is planning to issue a partial Exemption relative to Facility Operating License No. DPR-73. issued to General Public Utilities Nuclear Corporation (the . licensee).

for operation of the Three Mile Island Nuclear Station. Unit 2 (TMI-2), located in Londonderry Township, Dauphin County. Pennsylvania.

ENVIRONMENTAL ASSESSMENT Identification of Proposed Action: The action being considered by the Commission is an exemption from the 10 CFR 50.54(a) requirement to update the facility's FSAR whenever the QA plan is revised. This partial exemption was requested in the licensee's letter dated April 11. 19B3. The Need for the Action: The exemption is warranted because GPUNC ha, already been given an exemption from the FSAR updating requirements of 10 CFR 50.71(e).

The subject exemption was issued on February 4, 1982. Since the FSAR is not being meintained current, as permitted by the going exemption, it is therefore consistent and justified that an exemption from the FSAR QA plan update requirements of 10 CFR 50.54(a) be granted. Pursuant to the February 1982 exemption, however. the licensee is still required to submit changes to its QA plan to the NRC. Environmental Impacts of the Proposed Actions: The staff has evaluated the subject exemption and concluded that it will not result in significant

• . , increases in airborne or liquid contamination radioactivity inside the reactor (uilding or in corresponding releases to the environment.

There are also no non-radiological impacts to the environment as a result of this action. Alternative to this Action: Since we have concluded that there is no nificant environmental impact associated with the subject Exemption, any alternatives to this change will have either no significant environmental impact or greater environmental impact. The principal alternative would be to deny the requested action. This would not reduce significant wental impacts of plant operations and would result in the application of overly restrictive regulatory requirements when considering the unique conditions at 1MI-2. Agencies and Persons Consulted:

The NRC staff reviewed the licensee's request and did not consult other agencies or persons. / Alternate Use of Resources:

This action does not involve the use of resources not previously considered in connection with the Final matic Impact Statement for TM!-2 dated March 1981. Finding of No Significant Impact: The Commission has detennined not to prepare an environmental impact statement for the subject Exemption.

Based upon the foregoing environmental assessment, we conclude that this action will not have a significant effect on the quality of the human environment.

i 1 * , For further details with respect to this action see letter to B. J. Snyder, USNRC. from R. C. Arnold, SPUNC, 1141-2 Quality Assurance Plan. Revision 2, dated April 11, 1983. The above document is available for inspection at the Commission's Public Document Room. 1717 H Street. N.W., Washington, DC, and at the Commission's Local Public Document Room at the State Library of Pennsylvania, Government Publications Section, Education Building, Commonwealth and Walnut Streets, Harrisburg.

Pennsylvania 17126. FOR THE NUCLEAR REGULATORY COMMISSION Director Three Mile Island Program Office Office of Nuclear Reactor Regulation

• /

-. ." .. . j ... I Docket No. 50-320 UNrTED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON.

D.c. 20555 May 16, 1985 Dockling and Service Section Office of the Secretary of the Commission

SUBJECT:

Three Mile Island Nuclear Station, Unit 2 Operating License No *. .DPR-73 Partial Exemption from the Requirements of to CFR 50.54(a) Two signed Originals of the Federal Reg!!!!!.

Notice identified below are enclosed for your transmittal to the Office of the Federal Register for publication.

Additional conformed copies ( ) of the Notice are enclosed for your use. . o Notice of Receipt of Application for Permit(s) and Operating Ucense(s).

o Notice of Receipt of Partial Application for Construction Permlt(s) and Facility License(s):

Time for Submission of Views on Antitrust Matters. o Notice of Availability of Applicant's Environmental Report. o Notice of Proposed Issuance of Amendment to Facility Operating License . . o Notice of Receipt of Application for Facility License(s);

Notice of Availability of Applicant's Environmental Report; and Notice of Consideration of Issuance of Facility License(s) and Notice of Opportunity for Hearing. o Notice of Availability of NRC Draft/Final Environmental Statemenl o Notice of Umited Work Authorization.

o Notice of Availability of Safety Evaluation Report. o Notice of Issuance of Construction PermIt(s).

i i -o Notice of Issuance of Facility Operating L1cense(s) or Amendment(s).

o Other: Partial Exemption

Enclosure:

Office of Nuclear Reactor Regulation As Stated 0

Bernard J. Snyde , Three Mile Isla d Program Ofc. ¥ ...

00l'E: '!he QA Plan will be distributed internally by the Quality Assurance

.* D-r>artmen

t. Nuclear GPU Nucl ** r Corpor.tlon Post Office Box 480 Route 441 South Middletown, Pennsylvania 17057 , 717 944-7621 TELEX 84-2386 Wnter"s Direct Dial Number: April 11, 1983 4410-83-L-0032

'lMI Program Office Attn: Dr. B. J. Snyder, Progran Director US Nuclear Regulatory Ccmnission Washington, OC 20555

Dear Sir:

ntree Mile Island Nuclear Station, Unit 2 ('00-2) Operating IJ.cense No. DPR-73 DoCket No. 50-320 'lMI-2 Recovery Quality Assurance Plan, Revision 2 Attached for your acceptance is the 'lMI -2 ReccM!ry Quality Assurance Plan, Revision 2. The Plan has been revised to include the changes to the safety review and apprCMl1 process as described in the lMI-2 TecluUca1 Specification C:'1ange Request No. 40 (TSCR 40) and Organization Plan, RevisiCll 6, subnittals.

Fu111np1ementation of the requ:!rements of the Recovery QA Plan relating to the safety review process will be performed in conjl.mCtiCll with the 1np1ementation of TSCR 40, pend:f.:ng NRC approval.

RevisiCll 2 1ncorparates clarifications, fomat, and general st:ructure changes to be CCI'lSistent, 1ilere possible, with the GPWC Operational QA Plan, Revision 0, which has been su1:mitted to the NRC. These changes do not alter the degree of to regulatory requirements.

LDite, ___ _ m&:ld-Trlr.

118 17.5 IC2deJ-M.

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A GPUN ccmnitments with respect to Regulatory Glides 1.58 and 1.146 as reflected LP. ______ _ in GPU letter IU-81-0184 dated Septerrber 20, 1981, have been 1ncorparated

.. :In this revision.

A general of GPt.NC positions CIl Regulato:ty QD.des bas been made to reflect positions specific to the '00-2 ReCC?Ye1Y.

'lbese positions have been previously presented to the NRC in the '00.-2 TeChnical Specifications, the 00-2 Project Desigrl-Criteria, the Operaticma1 Plan, Revis1Dn 0, and the Bechtel ToPica1 QA Plan, with the exception of Begulatary Qddes 1.54 and 1.94 Wich are presented for the first time :In this su1:mittal \dditionally, Regulatory QD.de 1.63 was deleted fran Recovery QA Plan as it 4XJntains 110 QA. requirements.

GPU Nuclear Corporation is a subsidiary of the General Public UtilitieS Corporation

'e,.' '" "" h 3.3.. g X.A .. S.. . -_.

=

• _ \-,.. Dr. B. J. Snyder, Progran Director 4410-83-L-0032 1:-; addition, GPUNC is requesting a partial ex£!11)tion to the update -:nents of 10 CFR Part 50.54(a) for 'lMI-2. 10 CFR Part SO.54(a) requires that an update to the Quality Assurance Progran, as described in the Safety Analysis RepOrt, be provided to the appropriate NRC Regional Office for inclusion in the Safety Analysis Report. Based on an NRC letter dated July 20, 1981, B. J. Snyder to G. K. Hovey, 'lMI-2 (Ucense No. DPR.-73) was issued an exsrption to the Final Safety Analysis Report (FSAR) update requirements of 10 CFR Part SO.7l(e) with the exception that certain SystE!ll Descriptions and Teclmica1 Evaluation Reports be updated on a six DDnth basis. An NRC letter dated February 4, 1982, B. J. Snyder to J. J. Barton. extended the update period fran every six DDnths to annually.

In accordance with the above referenced docLIDentaticn, GPWC requests that the NRC consider the attached '00-2 Recovery QA Plan, Revision 2, as the Quality Assurance Program update required by June 10, 1983, instead of requiring GPWC to update the program description in the '00-2 F'SAR.. Changes between this sulmittal and the previously approved '00-2 Recovery QA Plan, Revision 1, subnitted via GPt.N: letter 4410-82-L-0012 dated Septelrber

20. 1982, are as described above. If you have any questions, please feel free to contact Mr. J. J. Byrne of my staff. BKK/JJB/jep Sincerely, lsI R. C. Arnold R. C. Arnold President cc: Mr. L. H. Barrett, Deputy Progran Director -'IMI Program Office '.

".. . UNITED STATES NUCLEAR REGULATORY COMMISSION Docket No. 50-320 Mr. F. R. Standerfer Vice President/Director Three Mile Island Unit 2 GPU Nuclear Corporation P.O. Box 480 Middletown, PA 17057

Dear Mr. Standerfer:

WASHINGTON.

D. C. _ August 8, 1985

Subject:

Three Mile Island Nuclear Station Unit 2 Operating License No. DPR-73 Docket No. 50-320 Technical Specification Change Request 46 Exapt10n Request from 10 CFR 50, Appendix A General Design Criteria 34 and 37 . The Nuclear Regulatory Commission has issued the enclosed Amendment of Order and Exemptions from 10 CFR 50, Appendix A, General Design Criteria 34 and 37 effective September 23, 1985. The Amendment of Order which modifies sections of the Proposed Technical Specifications (PTS) was requested by General Public Utilities Nuclear Corporation in a letter dated November 6, 1984. Other correspondence related to this request includes a request for exemptions from the requirements of 10 CFR 50 General Design Criteria 34 and 37 in a letter dated March 26, 1985 and additional information which was supplied in a letter dated Harch 27, 1985 to support the changes requested in the PTS. Since the February 11, 1980 Order imposing the Proposed Technical cations is currently pending before the Atomic Safety and Licensing Board, the staff will be advising the Licensing Board of this Amendment of Order through a Notice of Issuance of Amendment of Order and a Motion to Conform Proposed Technical Specifications in Accordance Herewith.

Mr. F. Federal Register Notices for the discussed issuances are enclosed.

Copies of the related Safety Evaluation and revised pages for the Proposed Technical Specifications are also enclosed.

Enclosures:

1. Amendment of Order 2. Safety Evaluation

3. Proposed Technical Specification Page Changes Sincerely.

--6 .. ,/ Bernard J. Sn er. Program Director Three Mile Island Program Office Office of Nuclear Reactor Regulation

4. Exemption from 10 CFR 50. Appendix A. GDC 34 and 37 5. Notice of Environmental Assessment and Finding of No Significant Impact 6. Federal Register Notices cc: T. F. Demmitt R. E. Rogan S. Levin W. H. Linton JJ. J. Byrne A. W. M11 ler Service Distribution List (see attached)

Enclosure 1 . UNITED STATES NUCLEAR REGULATORY COMMISSlON In the Matter of GENERAL PUBLIC UTILITIES NUCLEAR CORPORATION Docket No. 50-320 (Three Mile Island Nuclear Station Unit 2) AMENDMENT OF ORDER I. 6PU Nuclear Corporation, Metropolitan Edison Company, Jersey Central Power and Light Company and Pennsylvania Electric Company (collectively, the licensee) are the holders of Facility Operating License No. DPR-73, which has authorized operation of the Three Mile Island Nuclear Station, Unit 2 (TMI-2) at power levels up to 2772 megawatts thermal. The facility.

which is located in Londonderry Township, Dauph'in County, Pennsylvania, is a pressurized water reactor previously used for the commercial generation of electricity.

II. By Order for Modification of License, dated July 20, the licensee's authority to operate the facility was suspended and the licensee's authority was limited to .. intenance of the facility in the present shutdown cooling mode (44 Fed. Reg. 45271). By further Order of the Director, Office of Nuclear Reactor Regulation.

dated February 11. 1980. a new set of formal license requirements was imposed to reflect the accident condition of the facility and to assure the continued

.. intenance of the current safe, stable, long-term cooling condition of the facility (45 Fed. Reg. 11292). Although these requirements were imposed on the licensee by an Order of the Director of Nuclear Reactor Regulation, dated February 11, 1980, the TMI-2 license has not been formally amended. The requirements are reflected in the proposed Recovery Mode Technic" SpeCifications (PTS) presently pending before the Atomic Safety and Licensing Board. The revisions that are the subject of this order do not give the licensee authorizations that may be needed to undertake specific cleanup activities.

These activities will require separate consideration by the staff per Section 6.8.2 of the PTS t individual staff safety evaluations and/or licensing actions as appropriate.

Hereafter in this Amendment of Order, the requirements in question are identified by the applicable Proposed Technical Specification.

III. By a letter dated November 6, 1984, General Public Utilities Nuclear Corporation (GPUNC) proposed changes to the Proposed Technical cations (PTS) for Three Mile Island Unit 2 (TMI-2) to reflect current plant conditions.

The staff has reviewed the licensee's proposed changes which can be grouped into the following categories:

(1) Modifications to the existing Limiting Conditions for Operation (LCD) that were proposed to more correctly state what systems or equipment are necessary based on the present status of THI-2. The proposed changes would delete the LCO that the Standby Reactor Coolant System Pressure Control System, Mini-Decay Heat Removal System, the Decay Heat Removal System pumps and its recirculation pathways and the Nuclear Service Closed Cooling System be operable.

The proposed changes would also modify the LCO for the required minimum amount of borated water in the Borated Water Storage Tank from 100,000 gallons to 390,000 gallons and the number of operable flow paths from the BWST from one to two. (2) New Limiting Conditions for Operation were also proposed to more correctly reflect what systems or equipment are necessary based on the present status of THI-2. The proposed LCO would ,require that dedicated on-site equipment for a Reactor Building Sump Recirculation System be operable.

The proposed LCO would also require that two flow paths downstream from the BWST be operable.

(3) Revisions to the Bases were proposed that reflect corresponding changes in the Limiting Conditions for Operation.

Exemptions from 10 CFR 50, Appendix A, Design Criterion 34 and Criterion 37 were also requested because of some of the subject deletions and alterations to the PTS. Other changes proposed by the licensee were applicable to the Recovery Operations Plan (ROP) and are addressed in separate correspondence.

The staff concludes that these changes are appropriate to more accurately reflect the current conditions and requirements at THI-2. The staff's safety assessment of the foregoing, which concludes that the proposed changes are acceptable from the standpoint of public health and safety. is set forth in the concurrently issued Safety Evaluation.

Since the February 11. 1980 Order imposing the Proposed Technical Specifications is currently pending before the Atomic Safety and Licensing Board. the staff will be advising the Licensing Board of this Amendment of Order through a Notice of Issuance of Amendment of Order and a Motion to Conform Proposed Technical Specifications in Accordance Herewith.

It is further determined that the modification does not authorize a change in effluent types or total amounts nor an increase power level and will not result in any significant environmental impact. In light of this determination and as reflected in the Environmental Assessment and Notice of Finding of No Significant Impact prepared pursuant to 10 CFR 51.2. and 51.30 through 51.32 issued concurrently herewith.

it was concluded that the action is insignificant from the standpoint of environmental impact and that an environmental impact statement need not be prepared.

IV. Accordingly.

pursuant to the Atomic Energy Act of 1954. as amended. the Director's Order of February 11. 1980, is hereby revised to incorporate the deletions.

additions.

and modifications set forth in Enclosure 3 bereto. This Amendment of Order shall be effective on September

23. 1985. For further deta 11 s wi th respect to thi s act i on, see (1) Letter to B. J. Snyder, USNRC, from F. R. Standerfer, GPUNC, Technical Specification Change Request No. 46 dated November 6, 1984, (2) Letter to F. R. Standerfer, GPUNC, from B. J. Snyder, USNRC, NRC Questions on Technical Specification Change Request No. 46, dated February 6, 1985, (3) Letter to B. J. Snyder, USNRC, from F. R. Standerfer, GPUNC, Technical Specification Change Request No. 46 (responses to NRC questions) dated March 27, 1985, (4) Letter to B. J. Snyder, USNRC, from F. R. Standerfer, GPUNC, General Design Criteria 34 and 37, dated March 26, 1985, and (5) The Director's Order of February II, 1980. All the above documents are available for inspection at the Commission's Public Document Room, 1717 H Street, N.W., Washington, DC 20555, and at the Conmission's Local Public Document Room at the State Library of . Pennsylvania, Government Publications Section, Education Building, wealth and Walnut Streets, Harrisburg, Pennsylvania 17126. FOR THE NUCLEAR REGULATORY COMMISSION Harold R. Denton, Director Office of Nuclear Reactor Regulation UNITED STATES NUCLEAR REGULATORY In the Matter of GENERAL PUBLIC UTILITIES NUCLEAR CORPORATION (Three Hile Island Nuclear Station Unit 2) l ) l EXEMPTIONS I. Docket No. 50-320 GPU Nuclear Corporation, Metropolitan Edison Company, Jersey Central Power and Light Company and Pennsylvania Electric Company (collectiyely, the licensee) are the holders of Facility Operating License No. DPR-73, which has authorized operation of the Three Mile Island Nuclear Station, Unit 2 (TMI-2) at power levels up to 2772 megawatts thermal. The facility, which is located in Londonderry Township, Dauphin County, 'PennsylYania, is a pressurized water reactor previously used for the commercial generation of electricity.

By Order for Modification of License, dated July 20, 1979, the licensee's authority to operate the facility was suspended and the licensee's authority was limited to maintenance-of the facility in the present shutdown cooling mode (44 Fed. Reg. 45271). By further Order of the Director, Office of Nuclear Reactor Regulation, dated February 11, 1980, a . new set of formal license requirements was imposed to reflect the accident condition of the facility and to assure the continued maintenance of the current safe, stable, long-tenl cooling condition of the facility (45 Fed. Reg. 11292). This license provides, among other things, that it is subject to all rules, regulations and Orders of the Commission now or hereafter in effect. 11. On November 6. 1984. Seneral Public Utilities Nuclear Corporation (GPUNC) submitted Technical Specification Change Request No. 46. This pondence contained a request to delete the Decay Heat Removal System from the TMI-2 Proposed Technical Specifications.

The staff responded to this and other change requests with a list of questions forwarded on February 6. 1985. The licensee was asked to consider whether exemptions from 10 CFR 50. Appendix A. General Design Criteria (SOC) 34. 35. 36 and 37 were appropriate.

SPUNC responded in correspondence dated March 27. 1985 which stated that exemptions from GDC 35 and 36 were not required.

However. an exemption request from &DC 34 and 37 was requested by SPUMC in a letter dated March 26. 1985. The staff is issuing the requested exemptions as discussed herein. Ill. 10 CFR 50. Appendix A, &DC 34 requires that a system to remove residual heat shall be provided.

The purpose shall be to transfer;fission product decay heat and other residual heat from the core at such a rate that acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded.

Since January 1981, the n41-2 core has been cooled passively via the to-ambient IIOde. At present the decay heat level is less than 12 Kw thermal with an associated maximum core temperature of less than 100°F. The .xi ... temperature that is credible while in this lIOde (no forced circulation) is less than 170°F assUling water level is lowered to the bottom of the hot leg nozzles. At this temperature sufficient buffer is still "maintained between the maximum anticipated core temperature and the temperature at which the water in the vessel boil (212°F). fore, the staff concludes that sinfe the current loss-to-ambient mode is effective for all anticipated core temperatures, the requirement to have a residual heat removal system (GDC 34) is no longer necessary at TMI-2. On the other hand, portions of the residual heat removal system at TMI-2 still contain radioactive contamination resulting from the accident.

Operation of the system could result in the spread of radioactive contamination.

In addition, the requirement to maintain an operable residual heat removal system would result in an unnecessary burden for ..

surveillance and testing and could result in unnecessary radiation exposures to the workers. Accordingly, an exemption for GDC 34 is warranted.

The licensee has proposed in Technical SpeCification Change Request No. 46 that a Reactor Building Sump Recirculation System (RBSRS) be used for emergency core cooling at TMI-2. The system would only be installed in the . event of an unisolable leak in the RCS. Licensee calculations, which are supported by the staff in an Amendment of Order concurrently issued with this exemption, conclude that at least 10 days are available between the detection of the worst-case credible leak and when the RBSRS would be required.

This gives ample time for the system to be put in service. As stated in the referenced Amendment of Order, the staff has accepted the RBSRS and its proposed method of use. This acceptance included Recovery Operations Plan requirements for testing the operability of .. jor system components on a regular basis (see the staff's Safety Evaluation Report dpproving the modifications to the Proposed Technical Specifications related to Borated Cooling Water Injection).-

GDC 37 requires the testing of the emergency core cooling system including the operab;11ty of the system as a whole and the performance of the full operational sequence.

Since the staff has accepted the installation of the RBSRS in the reactor building only in the event of an unisolable leak in the RCS, the testing of tHe system according to GOC 37 is not necessary.

In addition, since the reactor building basement still contains accident generated contaminated water, testing of a basement sump recirculation system in a full operational sequence could result in the spread of and radiation exposures to the workers. Accordingly, an exemption from GDC 37 is warranted.

IV. Accordingly, the Commission has determined that, pursuant to 10 CFR 50.12, an exemption is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest.

The Commission hereby grants exemptions from the requirements of 10 CFR 50, Appendix A General Design Criteria 34 and 37 in accordance with the licensee's request dated March 26, 1985. It is further determined that the exemptions do not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. In light of this .ination and as reflected in the Environmental Assess.ent and Notice of

-5: .. Finding of No Significant Environmental Impact prepared pursuant to 10 CFR 51.2 and 51.30 through 51.32, issued concurrently herewith, it was cluded that the instant action is insignificant from the standpoint of environmental impact Ind In environmental impact statement need not be prepared.

.. Effective Date: September 23, 1985 Dated at Bethesda, Issuance Date: August 8, 1985 FOR THE NUCLEAR REGULATORY COMMISSION

/.p ,aL Hlrold R. Denton, Director Office of Nuclear Reactor Regulation A. .I'---____ -.zr:f,,..DC .l, t 17 l.rtn,eT I:cCtlcff

-... '-------apu N ..... CorporatIon Post Office Box 480 Route 441 South \ b lITI Nuclear Middletown, Pennsytvania 17057-0191 717844-7621 1Nl Program Office Attn: Or. B. J. Snyder Program Director US Nuclear Regulatory Commission Washington, DC 20555

Dear Or. Snyder:

TELEX 84-2386 Writer's Direct Dial Number: (717) 948-8461 .. 10-85-1.-0055 Docunent 10 0198A March 26, 1985 Three Mile Island Nuclear Station, ltlit 2 (00-2) ,

License No. OPR-" Docket No * .50-320 General Design Criteria 34 and 37 In accordance with your request, elated February 6, 1985, GPU has evaluated General Design Criteria (GOC) 34 and 37 and has determined that exemption from these criteria is justified; rationale to support approval of these exemption requests are provided below. Q)C 34 Q)C 34 requires that a system be provided to remove residual heat. As discussed in the 00-2 FSAR, Steam Generators ere used to cool the plant to 2SOOF after which the decay heat removal system was used. As discussed in Section 4.3 of Technical Specification Change Request No. 46 (Tsat 46), the requirements for a system to remove residual heat no longer exists at 00-2. This is demonstrated by the fact that the 00-2 reactor has been maintained in a safe shutdown state by loss-to-ambient cooling since January 1981. This natural phenomenon occurs independent of plant CCIIq)onents, features, interconnections or a need for electrical power. Therefore, an exemption fran General DeSign Criteria 34 is appropriate. " GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation r ,'. Dr. B. J. Snyder March 26, 1985 441D-85-L-Q055 "CJ)C >> requires testing of the emergency core cooling system to assure (1) the structural and leak tight integrity of ITS components, (2) the operability and performance of the active components of the system, and (3) the operability of the-system as a whole under conditions as close to design as practical.

Since the Reactor Building Sump Recirculation System -(RRS) is intended only to be installed in the event of an L.Ilisolable leak from the reactor vessel, an eXeqJtion from OOC 'J7 is appropriate.

As discussed in TSCR No. 46, an L.Ilisolable leak is an extremely U1likely event. Should it occur, approximately 13 days are available to install the RRS and make it operational.

Based on this rationale, as discussed in tSCR No. 46, GPU ttJclear has concll.l:led that it is not ALARA to install this system and be required to perform the periodic maintenance in the Reactor Building.

Therefore, the RRS would not comply with the requirements in OOC 'J7 to assure: 1) the structural and leak tight integrity of ITS components, and 2) the operability of the system as a whole U1der conditions as close to design as practical.

However, as discussed in GPU Nuclear letter 441D-85-l-D054, 1q)ortant components of the system will be tested in accordance with the 1MI-2 Recovery Operations Plan in order to provide assurance that the system will operate, if required.

Therefore, even though an eXeqJtion to OOC 'J7 is required, the health and safety of the public w111 not be jeopardized.

Based on the above justification, GPU ttJclear requests eXeqJtion from the provisions of OOC 34 and 'J7 for Three Mile Island Unit 2. These exemptions do not involve a significant hazards consideration as discussed in Section 7.0 of TSCR #46. Sincerely, /sl F. R. Standerfer F. R. Standerfer Vice PreSident/Director, TMI-2 FRS/SSlvjf cc: Deputy Program Director -1MI Program Office, Dr. W. D. Travers !

I '1Ii-' Ill.

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON.

D. C. 20555 I Y:1 Docket No. 50-320 Mr. F. R. Standerfer Vice Pres identlDi rect"or Three Mile Island Unit 2 GPU Nuclear Corporation P.O. Box 480 Middletown, PA 17057

Dear Mr. Standerfer:

j I I October 17, 1985 I \ :Els.f:"" P 'f \ IISSI8C 10. e"'U) 1M: DIllE' r '" an .,.. tt ,u-. ...,. II or . elfor aunt' . tntr . .. " . . nt", tH .... r .t -:cc au iIRIi * ..,. ,IiII", r.

Subject:

Approval of Exemption from 10 eFR 30.51, 40.61, 70.51(d).

and 70.53 We have reviewed your request, dated April 18, 1985, for exemptions from the requirements of 10 eFR 30.51, 40.61, 70.5l(d), 70.53 and 70.54 ing record keeping, inventorying, and reporting of core special nuclear, source and byproduct materials.

As discussed in the attached Exemption, we have determined that you will have sufficient information to comply with the requirements of 10 eFR 70.54 and that an exemption from this regulation is unnecessary.

However, we conclude that your request for exemptions from the other regulations are appropriate and acceptable, as stated in the attached Exemption issued by the Director of the Office of Nuclear Reactor Regulation.

An environmental assessment of the action considered and a Federal Register notice for this issuance are also enclosed.

Enclosures:

1. Exemption

2. Environmental Assessment and Sincerely, -t3,u c t Bernard J.

Program Director Three Mile Island Program Office Office of Nuclear Reactor Regulation Notice of Finding of No Significant Environmental Impact 3. Federal Register Notices cc: T. F. Demmitt R. E. Rogan S. Levin w. H: Linton J. J. Byrne A. W. Miller Service Distribution List (see attached)

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Enclosure 1 UNITED STATES NUCLEAR REGULATORY COMMISSION In the Matter of GENERAL PUBLIC UTILITIES NUCLEAR CORPORATION (Three Mile Island Nuclear Station Unit 2) ) ) ) EXEMPTION I. Docket No. 50-320 GPU Nuclear Corporation, Metropolitan Edison Company, Jersey Central Power and Light Company and Pennsylvania Electric Company (collectively, the licensee) are the holders of Facility Operating License No. DPR-73, which has authorized operation of the Three Mile Island Nuclear Station, Unit 2 at power levels up to 2772 megawatts thermal. The facility, which is located in Londonderry Township, Dauphin County, Pennsylvania.

is a pressurized water reactor previously used for the commercial generation of electricity.

By Order for Modification of License, dated July 20, 1979, the licensee's authority to operate the facility was suspended and the licensee's authority was limited to maintenance of the facility in the present down cooling mode (44 Fed. Reg. 45271). By further Order of the Director, , Office of Nuclear Reactor Regulation, dated February 11, 1980, a new set of license requirements was imposed to reflect the post-accident condition of the facility and to assure the continued maintenance of the current safe, stable, long-term cooling condition of the facility (45 Fed. Reg. 11292). The license provides, amoni other things, that it is subject to all rules, regulations and Orders of the Commission now or hereafter in effect. II. By letter dated April 18, 1985, the licensee requested exemptions from 10 CFR 30.51. 40.61. 70.51(d), 70.53. and 70.54 regarding the requirements for record keeping, inventorying, and reporting of core special nuclear, source and byproduct materials.

Specifically, 10 CFR 30.51 40.61 specify the requirements for keeping records which show the receipt, transfer and disposal of source and byproduct material.

10 CFR 70.51(d) specifies the requirements for the periodic conduct of a physical inventory of all special nuclear material in possession.

10 CFR 70.53 specifies the requirements for the periodic submittal of a Material Balance Report and Physical Inventory Listing of special nuclear material possessed by the licensee.

10 CFR 70.54 specifies the requirements for submitting Nuclear Material Transaction Reports for the transfer or receipt of special nuclear material.

In meetings with the licensee held subsequent to the April 18, 1985 exemption request, staff representatives of the NRC and Department of Energy (DOE) have determined that the licensee will have sufficient information to comply with the requirements of 10 CFR 70.54 and that an exemption from this regulation is not necessary.

III. The accident at Three Mile Island Unit 2 severely damaged the reactor core. Video inspections and topography measurements indicate a cavity in the upper core region which rep:esents approximately 26% of the total original core volume. No more than 2 of the original 177 core fuel assemblies remain intact and only 42 assemblies have any full length fuel rods. The core damage extends radially all the way out to the core former walls. As a result of the accident induced embrittlement of virtually all fuel rods, no fuel assemblies are expected to be withdrawn intact. There is a significant amount of core debris in ex-core region locations.(e.g., an estimated 10 to 20 tons in the lower reactor vessel head) and much of the core byproduct material has been released from the fuel. For example, analyses of core debris bed samples indicate that, on the average, only about 13% of the original Cs-137 inventory remains in the fuel although the percentage retained can vary considerably from sample to sample. During the defueling of the damaged core, the fuel debris will be collected in canisters by vacuuming or *pick and place" techniques.

However, as a result of the damaged condition of the core, the licensee will have no means of accurately characterizing (e.g., U-235 enrichment and total uranium content, fission product radionuclide content and distribution, plutonium content) the fuel debris during the defueling sequence.

The capability for characterizing the fuel debris in each canister would require sophisticated hot cell and laboratory facilities with the means to homogenize, sample, weigh, and analyze the contents of each canister.

Such facilities do not exist at Three Mile Island. Given the damaged condition of the core and lack of sophisticated hot cell and laborato!y facilities, ther.e is no practical means for the licensee to perform the measurements or precise calculations necessary to comply with the Commission's regulations related to accountability of special nuclear, source and byproduct materials.

The staff therefore concludes that exemptions from the requirements of 10 eFR 30.51, 40.61, 70.51(d), and 70.53 are appropriate.

As previously stated in Section 11 of this evaluation, staff representatives of the NRC and DOE have determined that the licensee will have sufficient information to comply with trans!er requirements of 10 eFR 70.54 and that exemption from this regulation is not necessary.

The granting of these exemptions does not mean that the licensee will not provide any record keeping or reporting of the canister core debris which is intended to be transferred to the custody of the DOE for research and/or storage at DOE facilities in Idaho. In lieu of the reporting requirements of 10 eFR 70.53, the licensee will provide to the DOE all available information describing the physical contents of each canister including:

the canister identification number, canister type (i.e., knockout, fuel, or filter), date of shipment, the shipment number, the empty weight of the canister, the loaded weight of the canister, the dewatered weight of the canister, maximum total curies, the canister pressure, general physical description of the canister contents including videotape data (if available), and any additional information based on mutual agreement between the licensee and the DOE. Further, following the completion of defueling and the offsite shipment of the packaged fuel debris, the licensee will be in a position to comply with the requirements of 10 eFR 70.53 and the licensee will be required to submit a Material Balance Report and Physical Inventory Listing at that time. In lieu of the requirement in 10 CFR 70.51(d) for the periodic conduct of a physical inventory of all special nuclear material, the licensee will conduct such an inventory upon the completion and analysis of a defue11ng survey. In lieu of the record keeping requirements of 10 CFR 30.51 and 40.61, the licensee will maintain records of each fuel shipment in accoraance with the requirements of 10 CFR 71.91. Such records will include an identification of the shipment packaging, the maximum total curies, the total quantity of each shipment, and the date of shipment.

IV. Accordingly, the Commission has determined that, pursuant to 10 CFR 30.11, 40.14, and 70.14, these exemptions are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest.

The Commission hereby grants exemptions from the requirements of 10 CFR 30.51, 40.61, 70.51(d), and 70.53. The exemption from 10 CFR 70.53 shall expire following the completion of the defueling effort, including an assessment of any fuel fines and debris which remain within the plant, and the subsequent offsite shipment of all packaged fuel debris. It is further determined that the exemptions do not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. In light of this determination and as reflected in the Environmental Assessment and Notice of Finding of No Significant Environmental Impact prepared pursuant to 10 CFR 51.21 and 51.30 through 51.32, issued September 20, 1985, it was concluded that the instant action is insignificant from the standpoint of environmental impact and an environmental impact statement need not be prepared.

' Effective Date: October 17, 1985 Dated at Bethesda, Maryland Issuance Date: October 17, 1985 FOR THE NUCLEAR REGULATORY Harold R. Denton, Director Office of Nuclear Reactor Regulation Enclosure 2 UNITED STATES NUCLEAR REGULATORY COMMISSION GENERAL PUBLIC UTILITIES NUCLEAR CORPORATION DOCKET NO. 50-320 ENVIRONMENTAL ASSESSMENT AND NOTICE OF FINDING OF NO SIGNIFICANT ENVIRONMENTAL IMPACT The U.S. Nuclear Regulatory Commission (the Commission) is pTanning to issue an Exemption from certain regulations relative to the Facility Operating License No. DPR-73, issued to General Public Utilities Nuclear Corporation (the licensee), for operation of the Three Mile Island Nuclear Station, Unit 2 (TMI-2), located in Londonderry Township, Dauphin County, Pennsylvania.

ENVIRONMENTAL ASSESSMENT Identification of Proposed Action: The action being considered by the Commission is the granting of exemptions from the inventory, record keeping and reporting requirements of 10 CFR 30.51, 40.61, 70.51(d), and 70.53 for core special nuclear, source and byproduct materials.

Specifically, 10 CFR 30.51 and 40.61 require the maintenance of records showing the receipt, transfer and disposal of source or byproduct material.

10 CFR 70.51(d) specifies the requirements for the periodic conduct of a physical inventory of all special nuclear material in a licensee's possession.

10 CFR 70.53 specifies the requirements for the periodic submittal of a Material Balance Report and a Physical Inventory Listing for special nuclear material (SNM). The Need for the Action: Given the severely damaged condition of the TMI-2 core fuel, the dislocation of fuel material from its original location in the reactor pressure vessel, and the nonhomogeneity of the dislocated material which has settled into piles of rubble at the bottom of the containment vessel. the licensee is unable to determine the bulk quantity of material in the vessel or to obtain representative samples for mination of source. byproduct.

and SNM content in compliance with the core accountability requirements of 10 CFR 30.51. 40.61, 70.51{d) and 70.53. Accordingly, some relief from the Commission's regulatory requirements related to core accountability is warranted.

Environmental Impacts of the Proposed Actions: The staff has evaluated the exemptions and concluded that, as the exemptions are related to record keeping and reporting requirements.

there are no significant radiological or nonradiological impacts to the environment as a result of this action. Alternate to this Action: Since we have concluded that there is no nificant environmental impact associated with the exemptions.

any alternatives will have either no significant environmental impact or greater environmental impact. Alternatives to the exemptions would not reduce present environmental impacts of plant operations and would result in the application of overly restrictive regulatory requirements when considering the unique conditions of TMI-2. Agencies and Persons Consulted:

The NRC staff reviewed the licensee's i request and did not consult other agencies or persons. Alternate Use of Resources:

This action does not involve the use of resources not previously considered in connection with the Final matic Impact Statement for TMI-2 dated March 1981. Finding of No Significant Impact: The Commission has not to prepare an environmental impact statement for the subject Exemption.

Based upon the foregoing environmental assessment, we conclude that this action will not have a significant effect on the quality of the human environment.

For further details with respect to this action see, (1) Letter from F. R. Standerfer, GPUNC, to B. J. Snyder, USNRC, Core Accountability Exemption Requests, dated April 18, 1985. The above documents are available for inspection at the Commission's Local Public Document Room, 1717 H Street, N.W., Washington, DC, and at the Commission's Local Public Document Room at the State Library of Pennsylvania.

Government Publications Section, Education Building, wealth and Walnut Streets, Harrisburg, Pennsylvania 17126. FOR THE NUCLEAR REGULATORY COMMISSION Bernard J.

Program Director Three Mile Island Program Office Office of Nuclear Reactor Regulation ENCLOSURE 3

'6' u .*.** "a .... It.T N' ....... ""'ca. 1175-5.1-712 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 October 17, 1985 Docket No. 50-320 --Docketing and Service Section Office of the Secretary of the Commission

SUBJECT:

Three Mile'-Is1and Unit 2 Approval of Exemption from 10 eFR 30.51, 40.6l,.70.51(d).

and 70.53 Two signed originals of the Federal Register Notice identified below are enclosed for your transmittal to the Office of the Federal Register for publication.

Additional conformed copies ) of the Notice are enclosed for your use. O.Notice of Receipt of Application for Construction Permit(s) and Operating Ueense(s).

o Notice of Receipt of Partial Application for Construction Permit(s) and Facility Ucense(s):

Time for Submission of Views on Antitrust Matters. D Notice of Availability of Applicant's Environmental Report. o Notice of Proposed Issuance of Amendment to Facility Operating License. D Notice of Receipt of Application for Facility License(s);

Notice of Availability of Applicant's Environmental Report; and Notice of Consideration of Issuance of Facility Ucense(s) and Notice of Opportunity for Hearing. o Notice of Availability of NRC Draft/Final Environmental Statement.

D Notice of Limited Work Authorization.

o Notice of Availability of Safety Evaluation Report. o Notice of Issuance of Construction Permit(s).

o Notice of Issuance of Facility Operating License(s) or Amendment(s).

[J Other:_,;..Ex

.... pl:Am"ip.L.It"-.Jiu.owDL-.-

_______________________

_

Enclosure:

Office of Nuclear Rt£r Regulation As Stated

.. UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 September 16, 1985 Docket No. 50-320 Docketing and Service Section Office of the Secretary of the Commission

SUBJECT:

Three Mile Island Unit 2 Environmental and Notice of Finding of No Significant Environmental Impact Two signed originals of the Federal Register Notice identified below are enclosed for your transmittal to the Office of the Federal Register for publication.

Additional conformed copies ( ) of the Notice are enclosed for your use. o Notice of Receipt of Application for Construction Permit(s) and Operating License(s)

.. o Notice of Receipt of Partial Application for Construction Permit(s) and Facility License(s):

Time for Submission of Views on Antitrust Matters. o Notice of Availability of Applicanfs Environmental Report. o Notice of Proposed Issuance of Amendment to Facility Operating License. o Notice of Receipt of Application for Facility License(s);

Notice of Availability of Applicant's Environmental Report; and Notice of Consideration of Issuance of Facility License(s) and NC?tice of Opportunity for Hearing. . o Notice of Availability of NRC Draft/Final Environmental Statement.

o Notice of Limited Work Authorization.

o Notice of Availability of Safety Evaluation Report. o Notice of Issuance of Construction Permit(s).

o Notice of Issuance of Facility Operating License(s) or Amendment(s). Other: Environmental Assessment and Notice of Finding of No Significant Environmental Impact

Enclosure:

As Stated NRC FORM 102 {1*76)

Office JJluclear Reactor Regulation

• ..

I ,J -um Nuclear GPU Nucl ** r Corporation Post Office Box 480 Aoute 441 South TMI Program Office Attn: Or. B. J. Snyder Program Director US Nuclear Regulatory Commission

.. :;;;;-.... . . Washington, DC 20555

Dear Or. Snyder:

717944*7621 TELEX 84*2386 Writer's Direct Dial Number: (717) 948-8461 441D-S5-L-0174 Document 10 0310A August 27, 1985 .. Three Mile Island Nuclear Station, Unit 2 (TMl-2) Operating License No. OPR-73 Docket No. 50-320 10 CFR 50.61 Exemption Request New rulemaking, promulgated as 10 CFR50.6l, -Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events", and issued on July 23, 1985, imposes certain requirements on licensees with respect to pressurized thermal shock (PTS) events. Specifically, Paragraph (b)(l) of this rule requires licensees to submit projected values for Reference Temperature, pressurized thermal shock (RTPTS), by January 23, 1986. Additionally, Paragraph (b)(3) of this rule requires the licensee to submit, by April 23, 1986, an analysis and schedule for implementation of a flux reduction program if the projected values are expected to exceed PTS screening criterion at any time during the II fe of the plant. GPU Nuclear Corporation Is a subsidiary of the General Public Utilities Corporation

_ ...

, > Dr. B. J. Snyder August 27, 1985 4410-85-L-D174 Because of the lI'1ique shutdown condition of TMI-2, (i.e., the reactor vessel head has been removed and the reactor coolant system, which is open to >the reactor building atmosphere, is no longer capable of retaining pressure), the potential for a PTS is not a concern during tne recovery period. Based on the above rationale, GPU Nuclear requests an exemption from 10 CFR 5O.61(b)(1) and (3). In the event a decision is made to restore TMI-2 to an operable condition, GPU Nuclear will comply with the PTS requirements specified in the amendment to Paragraph (b) of 10 CFR 50.34 Which applies to operating license applications.

Per the requirements of 10 CFR 170, an application fee of $150.00 is enclosed for review of this document.

FRS/ROW/eml Sincerely, lsI T. F. Demmitt for F. R. Standerfer Vice President/Director, TMl-2 Enclosed:

GPU Nuclear Check No. 00017373 cc: Deputy Program Director -TMI Program Office, Dr. W. D. Travers

, UNITED STATES NUCLEAR REGULATORV COMMISSION WASHINGTON.

D. C. 20tH Mr. F. R. Standerfer Vice President/Director Three Mile Island Unit 2 GPU Nuclear Corporation P. O. Box 480 Middletown, PA 17057

Dear Mr. Standerfer:

, , '.

Subject:

Three Mile Island Nuclear Station Unit 2 Operating License No. DPR-73 Docket No. 50-320 Approval of Exemption from 10 CFR 50.61 Mm, tAL Atm.oy.tn, OC'

--1,..::u"A:J..-

___ _ We have reviewed your request dated August 27, 1985, for exemption from the requirements of 10 CFR 50.61 regarding steps necessary to protect the Reactor Coolant System (RCS) against pressurized thermal shock events. As discussed in the enclosed Exemption, the lack of pressure in the RCS and essentially ambient core and RCS temperatures, a pressurized thermal shock is not a credible event. Therefore, measures taken to protect against pressurized thermal shock are not warranted.

We conclude that your request for exemption from 10 CFR 50.61 is appropriate and acceptable, as stated in the enclosed Exemption issued by the Director of the Office of Nuclear Reactor Regulation.

An environmental assessment and a Federal Register notice for this issuance are also enclosed.

Enclosures:

1. Exemption Sincerely, William D. Travers, Director TMI-2 Cleanup Project Directorate Office of Nuclear Reactor Regulation

2. Environmental Assessment and Notice of Finding of No Significant Environmental Impact 3. Federal Register Notices cc: See next page.

Mr. F. R. Standerfer cc: T. F. Demm1tt R. E. Rogan S. Linton W. H. Linton J. J. Byrne A. W. Miller Service Distribution List (see attached) . ,

TMI-2 SERVICE LIST Dr. ,.....S "'rl., .. g'ol\ll ..,.'n,strltor.

bV'on t u.s. luellir Regulltory co.-fss'on 131 hrlt Avenue lUng of "","'1. PA 11'06 .Jolm F. Molf ** Esq ** CIII,,...n.

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D.C. 20555 IIr'n ". Carur Assistant Attorntf5tntr.'

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PA 17120 Or. Judith H. Johnsrud £nvironMent.'

CoIl1t'on on luele.r Power

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1800 M. St., IIW. Wlshington, D.C. 20036 Atomic Safety .nd Licensing IoIrd ",.., U.S. lIucle.r Regul.tory washington, D.C. 2D555 AtomiC Saf.ty Ind licensing Appell Pa"" U.S. lucl,.r R,gul.tory C ... i,sion washington, D.C. 20555 Secr,tary U.S. lucl,.r Regul.tory C",'ssion ATTN: Chief, Oock'tlng , SeryiCe .... nch washington.

D.C. 20555 ftr. Larry Hochtndontr Dauphin County eo..issfoner

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PA 17108-1295 John E. Minnich, Clllirperson, D.uphfn County Io.rd of C"'issiontrl Dauphin County Courthouse Front .nd Market Strltts Harrisburg, PA 17101 Dauph'n County Offfc, of m.trgency

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17602 Enclosure 1 UNITED STATES NUCLEAR REGULATORY COMMISSION In the Matter of GENERAL PUBLIC UTILITIES NUCLEAR CORPORATION (Three Mile Island Nuclear Station Unit 2) EXEMPTION I. Docket No. 50-320 GPU Nuclear Corporation, Metropolitan Edison Company, Jersey Central Power and Light Company and Pennsylvania Electric Company (collectively, the licensee) are the holders of Facility Operating License No. DPR-73. which has authorized operation of the Three Hile Island Nuclear Station, Unit 2 (nlI-2) at power levels up to 2772 megawatts thennal. The facil1.ty.

which is located in Londonderry Township, Dauphin County. Pennsylvania.

is a pressurized water reactor previously used for the commercial generation of electricity.

By Order for Modification of License, dated July 20, 1979. the licensee's authority to operate the facility was suspended and licensee's authority was limited to maintenance of the facility in the present shutdown cooling mode (44 Fed. Reg. 45271). By further Order of the Director.

Office of Nuclear Reactor Regulation, dated February 11. 1980. a new set of fonnal license requirements was imposed to reflect the post-accident condition of the facility and to assure the continued .aintenance*of the current safe. stable, long-term cooling condition of the facility.

(45 Fed. Reg. 11292). The license provides, among other things. that it is subject to all rules, regulations and Orders of the Commission now or hereafter in effect. II. By letter dated August 27, 1985. the licensee requested exemptions from 10 CFR 50.61 requiring the submission to the U.S. Nuclear Regulatory Commission of projections, analyses, schedules and other steps necessary to protect against pressurized thermal shock events. Specifically, Paragraph (b)(1) of 10CFR 50.61 requires licensees tb submit projected values for Reference Temperature for each weld and plate or forging in the reactor vessel belt11ne and Paragraph (b)(3) requires an analysis and schedule for implementation of a flux reduction program if the projected values of Reference Temperature are expected to exceed the pressurized thermal shock screening criteria set forth in Paragraph (b)(2) of 10 CFR 50.61. Additionally, the rule requires certain steps be taken if the flux reduction program does not result in reducing the value of the Reference Temperature below that of the pressurized thermal shock screening criteria.

III. Nuclear plant pressure vessels are fabricated from ferritic steels. A pressure vessel must be designed to maintain fracture toughness of the vessel material for the life of the plant. The pressure vessel of a nuclear plant can be subjected to a pressurized thermal shock (PTS) when an extended cooling transient to the vessel wall is accompanied by primary system pressurization.

Under these conditions repeated thermal and . pressurization stresses on the internal surfaces of the vessel can cause the fonmation of cracks. An adequate level of fracture toughness provides assurance that small cracks will not propagate in a -brittle" manner as a result of stresses during an abnormal transient such as a PTS event. Failure in a brittle manner could fracture the vessel wall and lead to severe failure of the primary pressure boundary in the core area. Due to irradiation damage, older pressure vessels generally have a greater probability of shifting the fracture toughness curve to higher temperatures.

thereby increasing the probability of nonductile or brittle vessel failure. ,; , For a pressurized shock to result in a significant nonductile failure the following conditions lUst be present: o o o The nuclear plant pressure vessel must exhibit significant loss of fracture toughness through neutron irradiation.

An overcooling transient must occur that is of sufficient duration to cause a steep thermal gradient across the vessel wall and cooling to the low-toughness temperature range. A flow must be present of sufficient size and be located at a critical vessel beltline location where reduced fracture toughness and high thermal stress exist. A s1mu,taneous high reactor coolant pressure must be present. IV. The staff has reviewed the past and present condition of the damaged TMI-2 reactor and has determined that: ° ° ° The plant went critical on March 28, 1978 and went into commercial operation on December'30, 1978. The accident at TMI-2 occurred on March 28. 1979. Neutron irradiation damage to the vessel is minimal. Since the middle of July 1982, the Reactor Coolant System (ReS) has been essentially vented to the reactor building.

Since July of 1984. the reactor pressure vessel head has been removed. With the reactor vessel head removed the ReS cannot be pressurized.

The licensee has no plans at this time to repressurize the ReS. As of the middle of September 1985, the incore thermocouple readings range from 70°F to 91°F with an average of 79°F. ,The average cold leg temperature is 54°F. The incore temperature continues to drop over time. ReS cooling is by natural heat loss to the reactor building ambient atmosphere.

No future increase in temperature is expected but rather continued slow cool down. With the licensee readying for the commencement of fuel removal, the lack of pressure in the RCS and essentially ambient core and RCS temperatures.

a pressurized thermal shock is not a credible event. Therefore.

the determination of projected values for Reference Temperature for each weld and plate or forging in the reactor vessel beltline and the development of mitigative should the Reference Temperature exceed the criteria are not warranted.

Undertaking the analyses and other actions required by 10 CFR 50.61 would impose an unnecessary burden and expense on the licensee with no concomitant benefit. v. Accordingly, the Commission has determined that. pursuant to 10 CFR 50.12. an exemption is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest.

The Commission hereby grants an exemption from the requirements of 10 CFR 50.61. It is further determined that the exemption does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any Significant environmental impact. In light of this determination and as reflected in the Environmental Assessment and Notice of Finding of No Significant Environmental Impact prepared pursuant to 10 CFR and 51.30 through 51.32. fssued on December 19. 1985. ft was concluded that the instant action is insignificant from the standpoint of environmental impact and an environmental fmpact statement need not be prepared.

Effective Date: December 3D, 1985 Dated at Bethesda, Maryland Issuance Date: December 3D, 1985 FOR THE NUCLEAR REGULATORY COMMISSION Bernero. Acting Director Office of Nuclear Reactor Regulation

," Enclosure 2 UNITED STATES NUCLEAR REGULATORY COMMISSION GENERAL PUBLIC UTILITIES NUCLEAR CORPORATION DOCKET NO. 50-320 ENVIRONMENTAL ASSESSMENT AND NOTICE OF FINDING OF NO SIGNIFICANT ENVIRONMENTAL IMPACT The U.S. Nuclear Regulatory Commission (the Commission) is planning to issue an Exemption relative to the'Facility'Operating License No. DPR-73 issued to General Public Utilities Nuclear Corporation (the licensee), for operation of the Three Mile Island Nuclear Station, Unit 2 (THI-2), located in Londonderry Township, Dauphin County, Pennsylvania.

ENV I ASSESSMENT Identification of Proposed Action: The action being considered by the Commission is an exemption from assessments, analyses and other requirements of 10 CFR 50.61 for protection against pressurized thermal shock events. Specifically 10 CFR 50.61 requires the licensee to submit to the U.S. Nuclear Regulatory Commission projected values for reference temperature for each weld and plate or forging in the reactor vessel beltline and an analysis and schedule for implementation of a flux reduction program if the projected values of reference temperature are expected to. exceed the pressurized thermal shock criterion set forth in Paragraph (b)(2) of 10 CFR 50.61. The Need for the Action: Given the lack of pressurization of the Reactor Coolant System (RCS) and the low core and RCS temperatures.

pressurized thermal shock is not a credible event. Accordingly.

analyses to determine the potential for and actions to protect against pressurized thermal shock for each weld and plate or forging in the reactor vessel beltline are not warranted.

Undertaking the'analyses and other actions required by 10 CFR 50.61 would impose an unnecessary burden and expense on the licensee with no concomitant benefit. Environmental Impacts of the Proposed Actions: The staff has evaluated the subject exemption and concludes that there are no significant radiological or nonradiological impacts to the environment as a re$ult of this action. The exemption removes the Commission's requirement to conduct analyses and make assessments of pressurized thermal shock events. Alternate to this Action: Since we have concluded that there is no . significant environmental impact associated with the subject Exemption, any alternatives to this change will have either no significant environmental impact or greater environmental impact. This would not reduce significant environmental impacts of plant operations and would result in the cation of unnecessary regulatory requirements.

Agencies and Persons Consulted:

The NRC staff reviewed the licensee's

-request and did not consult other agencies or persons.

.' Alternate Use of Resources:

This action does not involve the use of resources not previously considered in connection with the Final Programmatic Impact Statement for TMI-2 dated March 1981. Finding of No Significant Impact: The Commission has determined not to prepare an environmental impact statement for the subject Exemption.

Based upon the foregoing environmental assessment, we conclude that this action will not have a significant effect on the quality of the human environment.

For further details with respect to this action see; (1) Letter from F. R. Standerfer, GPUNC, to B. J. Snyder, USNRC, 10 CFR 50.61 Exemption Request, dated August 27, 1985. The above documents are available for inspection at the Commission's Local Public Document Room, 1717 H. Street, N.W., Washington, DC, and at the Commission's Local Public Document Room at the State Library of Pennsylvania, Government Publications Section, Education Building, Commonwealth and Walnut Streets, Harrisburg, Pennsylvania 17126. FOR THE NUCLEAR REGULATORY COMMISSION

'\ William D. Travers, Director TMI-2 Cleanup Project Directorate Office of Nuclear Reactor Regulation

• ENCLOSURE 3

• , Docket No. 5V-320 UNITED STATES NUCLEAR REGUlATORY COMMISSION WASHINGTON, D.c. 2D555 lie ..... 1'. 1t1S . Docketing and Service Section Office of the Secretary of the Commission

SUBJECT:

Enyiror..:er.u 1 As,n,..-.l lad IiOt tea 01 f tnet IIi. of Ho S tgrl1f ie.nt £nYlrGl.,.,ta 1 l..-ct Two signed originals of the Federal Reg!!!![ Notice identified below are enclosed for your transmittal to the Office of the Federal Register for publication.

Additional confonned copies ( ) of the Notice are enclosed for your use. " o Notice of Receipt of Application for Construction Pennit(s) and Operating Ucense(s) . . o Notice of Receipt of Partial Application for Cons1ruction Pennit(s) and Facility License(s):

Time for Submission of Views on Antitrust Matters. o Notice of Availability of Applicant's Environmental Report. o Notice of Proposed Issuance of Amendment to Facility Operating Weense. o Notice of Receipt of Application for Facility License(s):

Notice of Availability of Applicant's Environmental Report: and Notice of Consideration of Issuance of Facility License(s) and Notice of Opportunity for Hearing. o Notice of Availability of NRC OraftlFinai Environmental Statement.

o Notice of Umited Work Authorization.

o Notice of Availability of Safety Evaluation Report. o Notice of Issuance of Construction Permit(s).

o Notice of Issuance of Facility Operating License(s) or Amendment(s).

III Other:

Assessrreot EncIosLI'e:

As Stated \ : lilt 11 WI &,. Trlver,_ ilirectur 1.11-2 Cldliup 'roJett ltirectorite

• Office of Nuclear Reactor Regulation

• i t I I * ! 1 ,-... . l!I i .. J *

• J Ii 1 . 1 i ;. -Ij 1 ' . ---------_

.. -----------.--------.--------UNrTED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 December 30. 1985 Docket No. 50-320 Docketing and Service Section Office of 1he Secretary of 1he Commission

SUBJECT:

Approval of Exemption from 10 CFR 50.61 -.\ 'It Two signed originals of the Federal Reg!ster Notice identified below are enclosed for your transmittal to the Office of the Federal Register for publication.

Additional conformed copies ( ) of the Notice are enclosed for your use. -o Notice of Receipt of Application for Construction Pennit(s) and Operating License (s). o Notice of Receipt of Partial Application for Construction Permit(s) and Facility License(s):

Time for Submission of Views on Antitrust Matters . o Notice of Availability of Applicant's Environmental Report. o Notice of Proposed Issuance of Amendment to Facility Operating LIcense. o Notice of Receipt of Application for Facility License(s);

Notice of Availability of Applicant's Environmental Report; and Notice of Consideration of Issuance of Facility License(s) and Notice of Opportunity for Hearing. o Notice of Availability of NRC Draft/Final Environmental Statement.

o Notice of Umited Work Authorization.

o Notice of Availability of Safety Evaluation Report . o Notice of Issuance of Construction PermIt(s).

o Notice of Issuance of Facility Operating LIcense(s) or Amendment(s).

IJ Other: Exenption EncIos\.re:

AI Stated William D.

Director . TMI*2 Cleanup Project Directorate Office of Nuclear Reactor Regulation

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Dear Mr. Arnold:

Your letter to me of February 26, 1982 requested that the NRC vacate " ' its May 22, 1979 Order relating to the preservation-of records pertaining to the accident at Three Mile Island, Unit 2. We hive considered your request and, because of the potential importance of the rec*)rds to the continuing technical review of the accident and to the resulting from the accident, we believe that many of them need to bf. retained.

However, the Comnission did issue the enclosed Order on August 6, 1982 the disposal, in accordance with applicable NRC directives and regulations.

of catalogued physical samples taken after the ,accident, where the radioactivity of the sample has been and the resulting data recorded.

This Order also authorizes me to subsequent destruction of other particular records or categories of records which we detennine no longer need be retained.

Should you desire additional future relief from the original Order, you may request such relief by specifically identifying to this Office such records, together with the and options for their retention.

Enclosure:

Commission Order of August 6. 1982 R. Denton, Director Office of Nuclear Reactor Regulation

" ,

UNITED ST,o.1ES OF AMERICA NUCLEAR RESJLATORY COMMISSION Nunzio J. Palladino, Victor Gilinsky John f. Ahearne Thomas M. Roberts James K. Asselstine . ' .. , , ...... , ... SERVW AUG n 1932 Docket Ho e' 50-289 In the Matter of METROPOLITAN EDISON COMPANY (Three Mile Island Huclear . , U/Aii:; 110--Station, Unit No. 1)-DOCKa NUM3Ea _, " PROD ... UTIL FAC. ....

G ORDER SERVED AUG 111982 In order to assure the effectiveness of 1nvestiga\ions into

• aspects of the Three Mile Island Unit 2 acc1den,t.

the COIIIJIission on May 22, 1979, ordered the preservation of records relating to the accident * . The order required retention of all data. including documentary material and physical samples unless otherwise by 'the Director of the Commission's Special Investigation.

All persons relevant sources of data were ordered to preserve such records intact.* See . .-44 Fed. Reg. 30788 (May 29, 1979). ; On February 26. 1982. GPU Huelear requested that the Conmission vacate the May 22, 1979 record retention order. GPU seeks to reestablish its nOnlal course of business and retain business documents in accordance with existing regulat4r,y retention criteria.

rather than those imposed in the May ,22. 1979 order. , The Commission has considered the GPU request and finds that retention of some records covered by May 22. 1979 order is no longer * ' .

necessary.

Where the radioactivity of physical samples taken after the. TMI-2 accident has been detenmined and the resulting data recorded, ,. there is no need to retain such samples. Accordingly, the COIIITIission's MAY 22, 1979 order is vacated with respect to the retention of catalogued physical samples. They u.v be of in accor.dance with applicable HRC directives and regulations.

\

• Records other than catalogued physical samples remain valuable to the continuing technical review of the TMI-2 in connection with the anticipated examination of the reactor core. In addition.

these records lIay be important in litigation resulting from the acc1d'!nt. . . Records other than catalogued physical samples shall be retained as provided by the Commission's

29. 1979 order. unless the Director of the Office of Nuclear Reacto ulation finds that particular records or categories of records no 10nge be retained.!!

• The Director is .* hereby designated the authority allow destruction of .records covered by that order. Commissioner Ahearne dissents from this order. His dissenting views are attached.

It so *ORDEREO.Y Dated at Washington.

D.C. this 6 day of 1982 *1

• For the Commission This order does not affect the requirements for retention of.records contained in Appendix B of TMI-2 License No. (lPR-73 or imposed by the Director's Order of February 11. 1980 (.45 Fe-j. Reg. 11282, February 20, 1980). or the requirements of '10 CfR

• 50.n.: . Commissioner Gilinsky was not present when this Order was affinned.

but had previously his disapproval

• . Had Commissioner Gilinsky been present he would have affinaed his prior vote. ,

, ' , ,. . ' DISSENTING VIEWS OF COMMISSIONER AHEARNE I Am not prepared to join the Commission's Order. BAsec. on the information provided by the .taff, I was unable to identify even in general terms (1) cAtegories of the licensee believes must be retained under the Commission's . ' order of May 22, 1979, (2) categories of records the licensee . finds most burdensome,-and (3) categories of records the Oepartment of Justice and others (such as DOE) are interested in retaining for some specified If there categories of material that GPU'believes 1t is required to

• keep, that GPU finds burdensome to keep, and that NRC, DOE & DOJ cannot justify keeping, we should allow GPU to get rid of them. The NRC staff should have taken steps to identify such areas. '-

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( . , .", .. . " UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON.

O. c. 20IU J'UDe 18, 1981 , , --") Docket No. 50-320 . ' ---; . ., ... .:. i.,. * .., h.t .. ".-" Mr. Gal e Hovey Vice President and Director of TMI-Z Metropolitan Edison Company P.O. Box 480 Middletown.

PA 17057

Dear Mr. Hovey:

-JUl. 1 1981 The Director of the Office of Nuclear Reactor Regulation has issued the enclosed Order for the Three Mile Island Nuclear Station, Unit 2. This Order is effective immediately and requires that you shall promptly commence and complete processing of the intermediate level contaminated in the auxiliary building tanks and the highly contaminated water in the reactor building sump and in the reactor coolant system using the Submerged Deminera11zer System with effluent polishing by the tPICOR-II system. if

• '1'H1-2 Distribution A copy of the related Safety Evaluation Report (NUREG-0796) is als 0 AI'. -

Subject:

.ra" 1Z0Ila.1 6!&t!&G enclosed.

Sincerely, 1MI Program Office Office of Nuclear Reactor Regula

Enclosures:

1. Order 2. Safety Evaluation Report Related to the Operation of the Submerged Demineralizer System at Three Mile Island Nuclear Station, Unit 2 NUREG-0795, June 1981 t t Aadpecl to: 7 t-Due Z 7 Dist. C .. ol4-AII.Ie.

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• UNITED STATES OF AMERICA NUCLEAR REGULATORY In the Ma tter of METROPOLITAN EDISON COMPANY, ET Al. (Three Mile Island Nuclear Station, Unit No.2) ) ORDER I. 6/18/81 Docket No. 50-320 Metropolitan Edison Company, Jersey Central Power and Light Company and Pennsylvania Electric Company (collectively, the licensee) are the holders of Facility Operating License No. DPR-73. which had authorized operation of the Three Mile Island Nuclear Station, Unit 2 (THI-2) at power levels up to 2772 megawatts thermal. The facility, which is . located in Township, Dauphin County, Pennsylvania.

is a pressurized water reactor previously used for the commercial generation of electriCity.

II. Following the accident of March 28. 1979. by Order for Modification of License, dated July 20, 1979. the licensee's authority to operate the facility was suspended and its authority was limited to maintenance of the facility in the present shutdown cooling mode (44 F.R. 45271, August 1, 1919). By further Order of the Director.

Office of Nuclear Reactor Regulation.

dated February 11, 1980. a new set of license requirements was imposed to reflect the post-accident condition of the facility and to assure the continued maintenance of-the current safe, l ( * . . stable, long-term cooling condition of the facility (45 F.R. 11282.

20, 1980). As a result of the accident, about 700.000 gallons of highly contaminated water are standing in the Reactor Building sump and an approximately 95,000 gallons of highly contaminated water are contained in the reactor coolant system. In addition to the highly contaminated water. approximately 100,000 gallons of level water is being held in Building tanks. Although the highly contaminated waste water is presently safely contained in the Reactor Building sump and reactor coolant system. its presence ,here constitutes a continuing risk of,leakage to the environment and prevents or hinders the performance the major decontamination activities.

The Commission has clearly stated its intent that the licensee proceed expeditiously with all decontamination activities consistent with protection of the public health and safety and the environment. (Statement of PolicYi Programmatic Environmental Impact Statement of the Cleanup of Three Mile Island Unit 2. 46 F.R. 24764 (May 1. 1981).) The licensee has constructed the Submerged Oemineralizer System (-SOS") for proceSSing (decontamination) of this highly contaminated water. After processing by the SOS, the water may, if necessary.

be further processed

(-polished-)

by the EPICOR-II system, which has been previously used to decontaminate water which was located in Auxiliary Building tanks *. The intermediate level water which has accumulated since processing of the Auxiliary Building water with the EPICOR-II system wi1' also be processed with the SOS in order to minimize generation of solid waste and to check out the operation of the SOS. The SOS slstem has been thoroughly reviewed by the NRC Staff and the conclusions of that review Ire set forth in the Staff's Safety Evaluation Report (NUREG 0796. June 1981). The EPICOR-II system had been previously reviewed and approved by the Staff for treatment of Auxiliary Building water and has been further "reviewed for its polishing application to the SOS effluent.

After processing.

the decontaminated water would be stored in onsite tanks and the filters and zeolite ion exchanger vessels used in the SOS decontamination process would be temporarily stored underwater in the TMI-2 spent fuel storage pool. The EPICOR II resin liners, if any, would be temporarily stored in the existing onsite storage modules. The Oepartment"of Energy has stated its willingness to utilize and retain for research and development purposes the high specific activity zeolite solid wastes resulting from operation of the SOS. Low specific activity wastes (including filters) resulting from these 'operations should be -suitable for disposal by shallow land burial. The processed water wi1l be stored on site in available tankage until its disposition is proposed by the licensee.

reviewed by the Staff and approved by the Commission.

On the basis of its review, the Staff his determined that proceSSing of the Reactor Building sump and reactor coolant system highly contaminated water with the SOS and EPICOR II system would (1) enable the decontamination of the TMI-2 facility to proceed and (2) place the radioactivity in the waste water into an immobilized state from which

• release to the environment is much less likely. Processing of the Auxiliary'Building tankage intermediate level water with the SOS minimize generation of solid waste and enable the licensee to theck out the operation of the SOS. The Staff has determined that the public health, safety and interest require that the licensee promptly commence and complete procesSing of the highly contaminated Reactor Building sump and reactor coolant system water and the intermediate level water in the Building tanks with the SOS and, if poliShing by the EPICOR-II III. Accordingl!, pursuant to sections 103, 16lb and 1611 of the Atomic Energy Act of 1954, as amended, and the Commission's Regulations in 10 C.F.R. Parts 2 and 50, IT IS ORDERED EFFECTIVE THAT: The licensee shall promptly commence and complete processing of the highly contaminated Reactor Building and reactor coolant system water and the intermediate level water in the Building tanks with the SOS and, if necessary, the EPICOR II system. IV. The lfcensee or any person whose interest may be affected by this Order may, within thirty (30) days of the date of publication of this Order in the Federal Register.

file a request for a hearing with respect to this Order. pursuant to 10 C.F.R. 12.114. A request for a hearing shall be submitted to the Office of the U. S. Nuclear Regulatory Commission, washington.

D. C. 20555, Attention:

Docketing and

.. Service Section. by the above date. A copy of the request for a hearing should also be sent to the Executive Legal Director.

U. S.

Commission.

Washington, D. C. 20555. Any request for a hearing shall not stay the immediate effectiveness of this Order. If a hearing is requested by the licensee or other person who has an interest affected by this Order, the Commission will issue an order designating the time and place of any such hearing. If a hearing is held. the issue to be considered at such hearing shall be: Whether on the basis of the matters set forth in section II of this Order, th1s Order should be sustained.

The NRC Staff's Safety Evaluation Report (NUREG 0796) 1s ava1lable for inspection and copying, for a fee, at the Comm1ssions Public Document Room, 1717 H Street, N. W., Washington, D. C. 20555 and at the State of Pennsylvania, Government Publications Section, Education Building, Commonwealth and Walnut Streets, Harrisburg, Pennsylvania 17126. Copies are also available for inspection at the NRC's office at 100 Brown Street. Middletown, Pennsylvania 17057. Copies may be purchased for $4.50 directly from NRC by sending check or money order, payable to Superintendent of Documents, to Director, Division of Technical Infonnation and Document Control, U.S. Nuclear Comm1ssion, Washington, D.C. 20555. GPO Depos1t Account holders may charge their orders by call1ng (301) 492-9530.

Copies are also available for purchase through the National Techn1cal Infonnation Service, Springfield, Virginia 22161. . .

• . Dated at Bethesda, Maryland this 18thday of June, 1981. FOR THE NUCLEAR REGULATORY COMrlISSION aL Raro d R. litOfl:rrector Office of Nuclear Reactor Regulation

-....

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• PU Nuel,., CofporItlon o.t Offlc, Box.eo oute ... , South TMI-2 Cleanup Project Oirecto Attn: Or. W. O. Travers Director US Nuclear Regulatory c/o Three Mile Island Hue Middletown, PA 17057

Dear Dr. Travers:

co. r'. 17 a...*7621

LEX aA*2386 'rlt,r'. Direct Dial Number: (717) 948-8461 441D-86-1.-0207 Document 10 0141P As mbers of your st Nuclear understands that favorab e iew of Technical Specification Change Requests (TSCRs) 49 and 51, submit via GPU Nuclear letters 44lD-85-L-OllO dated ..lJne 18, 1985, and 44lD-85-L-35 dated ..lJly 31, 1985, respectively, require _exel1l'tion from 10 CFR SO, Appendix A, General Design Criteria (GOC) 17, "Electric Power Systems," and GDC 19, "Control Room." A request for said exemptions is SUbmi tted herein. Based on the justification provided in Attachments 1 and 2, GPU Nuclear has concluded that these exemptions from the GDC are warranted in accordance with 10 CFR 50.12(a)(2)(ii).

Due to the uniQue condition of TMI-2, compliance with the reQuirements of GOC 17 and 19 is not necessary to achieve the underlying intent of these criteria.

The safety evaluations submitted in support of the referenced TSCRs, as well as the analyses presented in lFU t<U:lear letters dated February 26, 1986, and 441D-86-L-008l dated May 20, 1986, provide the technical bases for this reQuest. GPU Nuclear Corporation Is

• lubsldiary of the General Public Utilities Corporation Or. Travers December 10, 1986 441Q-86-L-D207 Per the requirements of 10 aR 170, an application fee of $150.00 is enclosed.

". FRS/ROW/em!

Attacl'1'nents Sincerely, lsI F. R. Standerfer F. R. Standerfer Vice President/Director, TMI-2 Enclosed:

GPU Nuclear Corp. Check No. 000219

'1 ATIACf+ENT 1 44lo-86-L-0207 EXOfITION FRC14 OCt£RAL tl:SIGN 17 General Design Criterion (OOC) 17, "Electric Power Systems," to 10 a=R SO Appendix A requires that, "An on-site electric power system arel an off-site electric power system shall be provided to permit functioning of structures, systems, and components important to safety." Technical Specification Change Request (TSCR) 51 proposed, in part, deletion of the Class IE Oiesel Generators based on demonstrating that many original safeguard system loads requiring emergency diesel generator backup are no 10f'9!r required due to the current unique condition of 00-2. The safe evaluation attached to TSCR 51 demonstrates that "most of the loads automatically or manually sequenced on emergency diesel generators whose functions are no longer safety-related or functioning is no longer required to maintain safe plant Additionally, this SER demonstrates " *** all loads the emergency diesel generators ei of time conservatively assume restore supplied with back-up power sting stat rved by __ a length r or can be es."

I ATTAO+ENT 2 441D-86-L-0207 Fa:t [)(E)PTION FR()4 CD£RAL IESI(}o.I CRITERION 19 General Design Criterion (IJ)C) 19, *Control Room," to 10 a='R SO, Appendix A requires that, -A control room shall be provided from which actions can be taken to operate the ru::lear power unit safely under normal conditions and to maintain it*in a safe stlJtdown condition under accident Conditions, including 10ss-of-coo1ant accidents.

Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of S mrem whole body or its equivalent to any part of the body, for the duration of the accident." The analysis presented in (FlJ Nuclear letter 441D-86-L"(x)Sl dated 1986, specifically addressed this issue. This letter concluded t "Within its limits, the Control provide protection from a concurrent TMI-l LOCA and Control Room, tettpOrary

.. _roL .... would be required t TMI-2 in a safe s the public." , compliance with underlying purpose fely under normal er accident UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. c. 201156 IIIlIICD IIlI ",c. P -' February 9, 1987 III:

__ _

Mr. F. R. Standerfer Vice President/Director Three Mile Island Unit 2 GPU Nuclear Corporation P.O. Box 480 Middletown, Pennsylvania 17057

Dear Mr. Standerfer:

-,

Subject:

Three Mile Island Nuclear Station Unit 2 Operating License No. DPR-73 Docket Ho. 50-320 Approval of Exemption from 10 CFR 50, Appendix A General Design Criteria 17 and 19

Enclosures:

1. Exemption

2. Environmental Assessment and Notice of Finding of Significant Impact cc: See next Sincerely, William D. Travers, Director 1MI -2 Cleanup Project [lirectorate Office of Nuclear Reactor Regulation

.'

F. R. Standerfer cc: T. F. Demmitt R. E. Rogan S. Levin J. E. Frew J. J. Byrne A. W. Miller Service Distribution L;st (See attached)

" TMI-Z SERVICE-LIST Dr. Thomas Murley Regional Administrator U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia. PA 19406 Sheldon J. Wolfe. Esq., Chairman Administrative Judge Atomic Safety and licensing Board Panel U.S. Nuclear Regulatory Commission Washington.

D.C. 20555 Dr. Oscar H. Paris Administrative Judge Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington.

D.C. 20555 Dr. Frederick J. Shon Administrative Judge Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Washington, D.C. 20555 Dr. Judith H. Johnsrud Environmental Coalition Power 433 Orlando Avenue State College, PA 1680 atory CODmission . 20555 Frederick

• Rice, Chairman Dauphin County Board of Commissioners Dauphin County Courthouse Front* and Market Streets Harrisburg.

PA 17101 Thomas M. Gerusky. Director Bureau of Radiation Pratp.ction Department of Environmental Resources P.O. Box 2063 Harrisburg.

PA 17120 Ad Crable lancaster New Era 8 West King Street lancaster, PA 17601 u.S. Department of Energy P.O. Box 88 Middletown.

PA 17057-0311 David J. McGoff Office of LWR Safety Technology NE-23 U.S. Department of Energy Washington.

D.C. 20545 William lochstet 104 Davey laboratory Pennsylvania State University University Park. PA 16802 Jane lee 183 Va lley Road Etters. PA 17319 Walter W. Cohen. Consumer Advocate 1632 Department of Justice Strawberry Square. 14th Floor Harrisburg.

PA 17i27 Edwin Kintner Executive Vice Presider.t General Public Utilities Nuclear Corporation 100 Interpace Parkway Parsippany.

NJ 07054 Us Enviror.mental Prot. Agency Region III Office ATTN: EIS Coordinator Curtis Building (Sixth Floor) 6th and Walnut Strep.ts Philadelphia.

PA 19106 UNITED STATES NUCLEAR REGULATORY 11'\ the Matter of GENERAL PUBLIC UTILITIES NUCLEAR CORPOPATION (Three Mile Island Nuclear Station Unit 2) ) ) ) ) ) ) EXEMPTION I. GPU Nuclear Corporation, Metropolitan Edison Company, in Londonderry Towns reactor previously used By Order for Modificatio was Docket No. 50-320 see) (TMI-2) located a pressurized water licensee's authority (44 Director, Office of Nuclear dated February 11, 1980, a new set of formal license requirements was imposed to reflect the post-accident condition of the facility and to assure the continued maintenancp.

of the current safe, stable, long-term cooling condition of the facility (45 Fed. Reg. 11292). The license provides, amol'\g other things, that it is subject to all rules, regulations and Orders of the Commission now or hereafter in effect.

-.' II. By letter dated December 10, 1986, the licensee requested exemptions from the requirements of 10 CFR 50, Appendix A, General Design Criteria (GOC) 17 and 19, concerning electric power systems and control room Specifically, GOC 17 requires that an onsite electric power system and an offsite to provide sufficient acceptable fuel desia pressure boundary are n occurrences and (2) the vital functions terns and should be ated operational other ite electric power systems ncy to perform their request, shall be provided from which actions can nuclear power unit safely under normal conditions and to maintain it 1n a safe condition under accident conditions, including loss-of-coolant accidents.

GOe 19 further reQuires that adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures 1n excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident.

Under the TMI-2 operating license, control room habitability during accident conditions has been assured through the operability of the control room emergency air cleanup system. In the unlikely event of an accident with concurrent lossef offsite power (lOOP), the diesel generators would provide onsite emergency backup power to assure the operability of the system. III. May 20. 1986 y letters dated June February 26, 1986 and reviewed the safety evaluation amendments.

which also pr exemption requests.

requirements for generators.

As a result of se to reflect the unique post-accident

-2 facility.

the control room emergency air cleanup system 1s status .,aining system requiring power from the onsite diesel generators.

Consequently.

the licensee also proposes to delete the license requirement for onsite emergency backup power to this system. TrI-2 is currently in a long-teTm cold shutdown accident recovery and defueling.

Short.lived fission products which make up the preponderance of the source term in operating reactors have dpcayed to negligible levels.

-. -...... ---. ----_. -_ ... -----._------4 -Decay heat is less than 10 kilowatts and forced cooling of the core has not been required or used since 1981. Core cooling and criticality control are provided by maintaining a sufficient volume of borated water in the RCS. Natural convective heat loss the RCS directly to the reactor building atmrsphere provides sufficient decay heat removal capability.

In the unlikely event of the maximum credible TMI-2 loss of coolant accident, analyzed by the staff, sufficient borated water would be provid ive, for a longer period. The staff and the spill s, from th reactor. trum of potential accident These included liquid coolant accidents.

The source terms t are much smaller than those associated with cidents It operating power reactors.

Additionally, none of these accidents would be caused by a LOOP and thus are extremely unlikely to occur simultaneously with the unavailability of the control room emergency air cleanup system. 1MI-1. which is adjacent to TMI-2, is in a normal operating cycle for power reactors with periods of power operation periodically interrupted by variable length shutdowns for refueling, maintenance and repairs. A severe accident at while it is at power could generate a source term which could affect TMI-2 control room habitability.

It is very improbable that this type of accident would occur and even more unlikely that it would be coincident with a loss of offsite power to TMI-2. If there were no coincident TMI-2 lOOP, the control room emergency air cleanup system function norm While an accident at TM No active components are required is all that is required.

previously determined that can be restored within five hours. With the restoration of the TMI-2 control room emergency air cleanup system would again become operable and personnel could again monitor activities from the control room. Although not required, short-term access to the control could be provided by use rf self contained breathing apparatus.

The staff has evaluated the potential accident discussed above relative to the requirements for control room habitability specified in GOC 19. We conclude that it is highly probable that in the event of an accident at TMI-1 or TMI-2, the TMI-2 control room emergency air cleanup system will be operable without relying on onsite backup emergency power sources, and thus habitability of the TMI-2 control room will be ensured in accordance with GOC 19.

in the extremely unlikely event that a severe accident at Unit 1 occurs coincident with loss of offsite power to Un" Unit 2 control room if necessary, be evacuated without af in a safe shutdown condit Except provisions of GOC 19. the requirement

room, backup emergency longer safety of the facility in its present condition. xemption from GOC 17 is also justified.

This is based on the fact that the control room emergency air cleanup system is the only remaining load on the emergency diesel generators still required by the facility license; the emergency diesel generators (the onsite electric power system) are not to assure core cooling, containment integrity or other safety functions at TMI-2 in the current post-accident, cold $hutdown condition. IV. Accordingly, the Commission has determined that pursuant to 10 CFR 50.12, these exemptions are authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security.

The Commission further determines that special circumstances, IS pro IS noted above, neith onsite emergency backup at Design result in any significant environmental fmpact. In light of this determination and as reflected in the Environmental Assessment and Notice of Finding of No Significant Environmental Impact prepared pursuant to 10 CFR 51.21 Ind 51.30 through 51.32 (February 9, 1987, 52 FR 4067), it is concluded that the instant action is insignificant from the standpoint of environmental impact and an environmental impact statement need not be prepared. These exemptions are effective upon issuance of the corresponding changes to facility Technical Specifications.

sections 3.7.4. 3.7.7, 3.7.10, 3.8.1, 3.8.2, 3.9.12.1.

3.9.12.2. 3/4.7.4. 3/4.7.7.

4.3. 4.7.4, 4.7.7, 4.8.1, 4.8.2, 4.9.12.1 and 4.9.12.2.

Dated at Bethesda.

Maryland this 9th day of February 1987 FOR THE NUCLEAR REGULATORY COMMISSION d-Frank J.

irettor Division of PWR licensing-B Office of Nuclear Reactor R .'

UNITED STATES NUCLEAR REGULATORY GENERAL PUBLIC UTILITIES NUCLEAR COPPORATION DOCKET NO. 50-320 ENVIRONft1n 1 TAL ASSESSMENT AND NOTICE OF FINDING OF NO SIGNIFICANT ENVIRONMENTAL The U.S. Nuclear Regulatory Commission (the Commission) is plannin two Exemptions to the requirements of 10 CFR 50, Appendix A, Criteria 17 and 19, relative to to General Public Utilities N of the Three Island Township, Dauphin Coun qn issued dated July 20, 1979, the e faci lity was 45271). By further gulation, dated February 11, was imposed to reflect the nd to assure the continued maintenance safe, stable, long-term cooling condition of the facility (45 Fed. Reg. 11292). The license provides, among other things, that it is subject to all rules, regulations and Orders of the Commission now or hereafter in effect. ASSESSMENT of Proposed Action: The actions being considered by the Commission are exemptions from 10 CFP 50. Appendix A, General Design Criteria (GOC) 17 and 19 relating to requirements for electric power systems and nuclear station control rooms. Specifically, GDC 17 requires that an onsit offsite electric power system shall be provided to permit functi normal conditions and conditions.

GDC 19 furt provided to permit access .........

...:under accident cold assured to the reactor building atmosphere.

In case of a loss of coolant event, sufficient borated makeup water will be provided by a passive gravity feed system to maintain the level of the ReS above the damaged fuel for a minimum of 10 days. Dedicated equipment is available and procedures have been established to provide recirculation to assure long-term core coverage, if necessary, even 1n the unlikely event that the Unit 2 control room is temporarily uninhabitable.

Consequently, the facility can be maintained for the short term in a safe, stable shutdown condition without relying upon action from control room personnel and manning of the TMI-2 control room is not necessary under worst-case accident conditions to achieve the underlying purpose of the requirement.

For this reason, a partial exemption to the reauirements of GOC 19 is justified.

Since continuous control room manning is no longer necessary for generators is not need load on the diesel gener shutdown of the plant is power source. Thus, emergency diesel unnecessary benefit in terms of Environment 1 Impact of the Proposed Action: The staff has evaluated the subject exemptions and concludes that in of the current and future condition of the facility described above, there are no significant logical or nonradiological impacts to the environment as a result of this action. The exemptions remove specific features of the Commission's ments to provide an onsite electric power system and a control room to maintain the nuclear power unit in a safe condition following an accident. to the Proposed Action: Since the Commission has concluded that there is no significant environmental impect associated with the proposed exemptions, any alternatives to this action will have either no significant environmental impact or greater environmental impact. This would not reduce significant environmental impacts of plant operations and would result in the application of unnecessary regulatory requirements.

Agencies and Persons Consulted:

and did not consult other agencie Alternative Use of Reso resources not previously c Programmatic for the e an environmental impact statement xemptions.

Based upon the foregoing environmental assessment, not have a significant effect on the quality of the human environment.

For further details with respect to this action see: (1) letter from F. R. Standerfer, GPUNC to W. D. Travers, USNRC, Exemption from 10 CFR 50 Appendix A, General Design Criteria 17 and 19, dated December 10, 1986. This document is available for inspection at the Commission's Local Public Document Room, 1717 H Street, N.W., Washington, D.C., and at the Commission's local Public Document Room at the State library of Pennsylvania, Government Publications Section, Education Building, Commonwealth and Walnut Streets, Harrisburg, Pennsylvania 17126. Dated at Bethesda, Maryland, this 3rd day of February 198;. FOR THE NUCLEAR REGULATORY COMMI

GPU Nucl.ar Corporation Post Office Box 480 Route 441 South Middletown, Pennsylvania 17057*0191 717 944*7621 Mr. Victor Stella, Jr. Executive Director of Operations US Nuclear Regulatory Commission Washington, DC 20555

Dear Mr. Stella:

Noventler 20, 1986 TELEX 84*2386 Writer's Direct Dial Number: 441D-86-L-D181 DoclJTlent 10 OHaP Three Mile Island Nuclear Station, Unit 2 (TMI-2) Operating License No. DPR-73 Docket No. 50-320 10 CFR Part 171 Exemption Request ** 2 WrwWfifi; Al: .1I.1f' ,pertS ?b,t ,'"7, fo .... ,i ** J <£c,' d' aamc 1V:;--=:;.l.=,J

,;.;2>;....-.

__ _ III: l1li.: _______ _ ----IIEVZDIS----

NA CPS _____ _ The purpose of this letter is to request an exell1'tion from the 10 CFR Part 1711WJC11lj

_ .. fJ_,, ____ _ annual fee for Three Mile Island Unit 2 (TMI-2). This letter is submitted pursuant to Section 171.11, "Exemption", which states: III;

____ _ The Commission may, upon application, grant an exell1'tion, in part, from the annual fee required pursuant to this part. An exemption under this provision may be granted by the Commission taking into consideration the following factors: (a) Age of the reactor; (b) Size of the reactor; (c) Number of customers in rate base; (d) Net increase in KWh cost for each customer directly related to the annual fee assessed under this part; and (e) Any other relevant matter which the licensee believes justifies the reduction of the annual fee. Based on the justification provided in the attachment relevant to the above considerations, GPU Nuclear believes that a total exemption for TMI-2 is GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation l1li _______ _

Mr. Stello November 20, 1986 4410-86-L-0l8l warranted.

Granting this exemption would permit the maximum resources available to TMI-2 to be properly directed to funding cleanup activities.

Sincerely, lsi E. E. Kintner E. E. Kintner Executive Vice President EEK/FRS/eml At tachnent cc: Director -TMI-2 Cleanup Project Directorate, Dr. W. D. Travers ATTACHMENT 4410-86-L-018l Justification for Exemption from the Annual Fee Requirement of 10 CFR Part 171 1. 10 CFR 17l.1l(b) states the size of the reactor is a consideration in granting an exemption from the annual fee. In its present condition, TMI-2 is unable to operate; GPU Nuclear receives none of the benefits attendant to an operating plant. Therefore, for purposes of the annual fee, TMI-2 should be considered a "zero" power reactor and considerations relative to size and benefits received should apply. 2. 10 CFR 17l.1l(c) stated that the number of customers in the rate base is a consideration in granting an exemption.

Since TMI-2 is not in the rate base, fees associated with TMI-2 are not passed on to customers.

Regulatory fees paid by GPU Nuclear for TMI-2 are provided from designated cleanup funds and are not available to fund cleanup activities.

Therefore, it appears that consideration of the lack of a rate base should be considered.

3. 10 CFR 17l.ll(e) states that any other relevant matter which the licensee believes justifies the reduction of the annual fee will be considered.

GPU Nuclear opines that there are several factors unique to TMI-2 which warrant a TMI-2 exemption from the annual fee. a. As stated in the proposed rulemaking, published in the Federal Register (5lFR24078) on July 1, 1986, the annual fee is based on the several regulatory services provided by the NRC to persons applying for or holding operating licenses.

In addition, in response to comments on the proposed rule which were published in the Federal Register (5lFR33224) on September 18, 1986, the NRC stated that a review of these services was performed " *** to ensure that only generic costs associated with all power reactors, with operating licenses, regardless of types, were included in the cost basis." Based on review of the information provided in those notices, it appears that the unique condition of TMI-2 was not considered in the review. In consideration of the unique condition of TMI-2, it should not be encompassed in the broad categorization of " *** all power reactors, with operating licenses *** ". b. GPU Nuclear acknowledges that TMI-2 does receive special consideration from the NRC in that a separate "Cleanup Project Directorate" has been established to manage NRC oversight of TMI-2 activities.

This oversight is provided by review of safety evaluations prepared to support major recovery activities and inspections performed by an on-site staff of NRC inspectors which exceeds the normal resident inspector staff assigned to nuclear power stations.

However, the cost of these activities is recovered via the fees paid by GPU Nuclear in accordance with 10 CFR Part 170. For example, 10 CFR Part 170 Licensing fees paid thus far in 1986 approximate

$500,000.

Therefore, GPU Nuclear believes that the fees paid pursuant to 10 CFR Part 170 meet the intent of Congress in that these fees are " ..* reasonably related to the regulatory service provided by the Commission and fairly reflect the cost to the Commission of providing such service."

ATTACHMENT 4410-86-L-0181

c. The Final Rule provides that applicants for operating licenses are not subject to the 10 CFR Part 171 Annual Fees. GPU Nuclear suggests TMI-2 is similar to an applicant for an operating license with regard to benefits received from generic regulatory services provided by the NRC. While regulatory services from which the generic costs in the annual fee are derived may be of some future benefit, no benefit is currently realized.

Thus, TMI-2 should not be subject to an annual charge based on generic regulatory costs; rather, costs associated with TMI-2 support should be recovered in accordance with 10 CFR Part 170, as discussed above. d. The Final Rule acknowledges that certain classes of license receive limited benefit from the NRC generic programs and should not be subject to the annual fees. Thus, operating licenses subject to the annual fee do not include licenses for "possession only" based on a licensee request to amend a license to permanently withdraw authority to operate or those for which the Commission has permanently revoked the authority to operate. Both cases are closely analagous to TMI-2. GPU Nuclear presently receives the same limited benefits from NRC generic programs as "possession only" licenses or operating licenses whose authority to operate has been permanently revoked by the NRC and should be similarly excluded from the rule. Further, sufficient uncertainty exists concerning the feasibility of returning the TMI-2 plant to an operational status to warrant consideration as permanently non-operating within the context of this rulemaking until a determination to the contrary has been made. e. The cleanup and evaluation of the TMI-2 accident is providing information that is of benefit to the entire industry.

The research programs funded by the Department of Energy (DOE) and others provide a significant contribution to many of the NRC generic programs.

Of special significance is the evaluation of the radioactive source term which, as stated in the July 1, 1986 proposed rulemaking, " .*. lies at the very heart of the regulatory process." Thus, it is the GPU Nuclear viewpoint that TMI-2 is contributing significantly to NRC generic programs and payment of an annual fee to further support these programs is not appropriate.

UNITeD ITAT ** NUCLEAR REGULATORY COMMISSION WAI"INOTON.

D. c. _ Dock.t No. 50-320 GPU Nucl.ar Corporation ATTN: Mr. E. E. Kintner Executive Vice President 100 Interpace Parkway Parsippany, NJ 07054 Gentlemen:

_181987 11:,_,.:;.,11/,;;.,;./._,,;.

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2e. 19a&, for a total and permanent exemption from the annual fee requirements of 10 CfR 171 for Three Mile Island Unit No.2 (TMI-2). The basic just i. fication for your request was that TMI-2's licen,e ;, analogous to a npossession only" lieense and the reactor shou1d be considered a " zero" power reactor since 1t 15 incapable of producing electrical energyi the authority to operate it was revoked by the NRC in 1979. Therefore, it is your position that GPU receives the same limited benefits for TMI-2 from NRC's generic: programs IS "possession only*' licensees who are already exempted from the annual fees of 10 CFR 171. Based on our eva1uet1or.

of your request in accordance with the visions of '0 CFR 171.11, we have determined that an exemption from the annual fee requirements of '0 eFR 171 for TMI-2 should be granted. Th1s txempt10n is limited to FY 1987 only and is not a permanent exemption for TMI-2. Our evaluation took into consideration the fact that TMI*2 is a unique case for which the authority to operate was reduced in 1979 to taining the reactor in a "shutdown condition".

In addition, your p'a" for monitored storage after cleanup has been submitted for NRC review. This plan provides for removal of the reactor fuel from the site and plant systems and placing it in dry storage and hiS no provisions for refurbishment of the reactor for restart. Consistent with the requirements of 10 CFR 171.19, your Corporation paid the FY 1987 first quarterly payment of $237,500.

Since you hive been exempted from the annual f.es for FY 1987, we are taking the necessary steps to refund to you the $237,500 payment. This refund wi" be accomplished by electronic transfer within a week after your receipt of this letter. SinCerelY.

(;1 ctor StelJo. v: Executive 01rector for Operations