ML20244E632

Informs of CRGR Meeting 109 on 870209 in Bethesda,Md to Discuss Proposed New SRP Section 6.5.5, Pressure Suppression Pools as Fission Product Cleanup Sys

ML20244E632
Person / Time
Issue date:	01/30/1987
From:	Sniezek J Committee To Review Generic Requirements
To:	Bernero R Office of Nuclear Reactor Regulation
References
NUDOCS 8702050069
Download: ML20244E632 (56)

Text

{{#Wiki_filter:' fri UNITED STATES f NUCLEAR REGULATORY COMMISSION ] WASHINGTON, D. C. 20065 e -) JAN 3 01987 REL EW" ~0 THE PDR j i 1 i PEMORANDUM FOR: Robert M. Bernero, NRR Richard W. Starostecki, IE 9 Richard E. Cunningham, NMSS { Denwood F. Ross, RES l Clemens J. Heltemes, Jr., AE00 Joseph Scinto, 0GC James H. Snierek, Chairman J FROM: Comittee to Review Generic Requirements

SUBJECT:

CRGR MEETING NO. 109 The Committee to Review Generic Requirements (CRGR) will meet on Monday, February 9, 1987, 1-3 p.m. in Room 6110 MNBB. T. Speis (NRR) will present for CRCR review the enclosed proposed new Standa Review Plan Section 6.5.5, entitled " Pressure Suppression Pools as Fissien Product Cleanup Systems." (Category 2 1. tem.) If a CRGR member cannot attend the meeting, it is his responsibility to assure that an alternate, who is approved by the CRGR Chairman, attends the meeting.- l PersonsnakingfresentationstotheCRGRareresponsiblefor(1)assuringthat ( the information required for CRGR review is provided to the Committee (CRGR Charter - IV.B), (2) coordinating and presenting views of other offices, (3) as f appropriate,) assuring that other offices are represented dur tion,and(4 contact (Walt Schwink, xP8639) and others involved with the presentation. Division Directors or higher management should attend meetings addressing agenda items under their purvie:r. In accordance with the ED0's March 29, 1984 memorandum to the Commission con-cerning " Forwarding of CRGR Documents to the Public Document Room (PDR), enclosure, which contains predecisional information, will not be released to the PDR until the NRC has considered (in a public forum) or decided the matter addressed by the information. F ?r ) O @W / tW gJ nes H. Sniezek, Chaban.;. C ittee to Review Generic ' Requirements

Enclosure:

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l ' (1 JAN 3 01987 2 cc: SECY Commission (5) V. Stello, Jr. Office Directors Regional Administrators 1 W. Parler f T. Speis 8 ( Distribution: JSniezek JRoe j RFraley TRehm j ROGR Staff DEDROGR cf GZwetzig BZaleman MLesar FHebdon WLittle RErickson Central File PDR(NRG/CRGR) JClifford JZerbe PRabideau WMcDonald W01mstead l .A ) j 1 4 i i 1 OFC :ROGR

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T 8" NUCLEAR REGULATORY COMMISSION n 5 E WASHINGTON, D. C. 20S55 HELEASED TO THE PDR %,,,,,+ JAN 3 01987 PEMORANDUM FOR: Robert M. Bernero, NRR Richard W. Starostecki, IE Richard E. Cunningham, NMSS Denwood F. Ross, RES Clemens J. Heltemes, Jr., AE00 Joseph Scinto, OGC FROM: James H. Sniezek, Chainnan Committee to Review Generic Requirements

SUBJECT:

CRGR MEETING NO. 109 The Committee to Review Generic Requirements (CRGR) will meet on Monday, February 9, 1987, 1-3 p.m. in Room 6110 MNBB. T. Speis (NRR) will present for CRGR review the enclosed proposed new Standard Review Plan Section 6.5.5, entitled " Pressure Suppression Pools as Fission Product Cleanup Systems." (Category ? item.) If a CRGR member cannot attend the meeting, it is his responsibility to assure that an alternate, who is approved by the CRGR Chairman, attends the meeting. Persons making presentations to the CRGR are responsible for (1) assuring that the information required for CRGR review is provided to the Committee (CRGR Charter - IV.B), (?) coordinating and presenting views of other offices, (3) as appropriate, assuring that other offices are represented during the presenta-tion, and (4) assuring that agenda modifications are coordinated with the CRGP contact (Walt Schwink, x28639) and others involved with the presentation. Division Directors or higher management should attend meetings addressing agenda items under their purview. In accordance with the ED0's March 29, 1984 memorandum to the Commission con-cerning " Forwarding of CRGR Documents to the Public Document Room (PDR)," the enclosure, which contains predecisional information, will not be released to I the PDR until the NRC has considered (in a public forum) or decided the matter addressed by the information. l QjwLJ /@ty ' J'a nes H. Sniezek, ChaWman .C.cittee to Review Generic Requirements l l

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yg, 'I <6R - g 4fg H" Ja I P a %h UNITED STATES 8 y NUCLEAR REGULATORY COMMISSION f_, o _y j WASHINGTON, D. C. 20555 DEC 2 21986 MEMORANDUM FOR: James H. Sniezek, Deputy Executive Director 1 for Regional Operations and Generic Requirements -FROM: Harold R. Denton, Director Office of Nuclear Reactor Regulation

SUBJECT:

PROPOSED NEW STANDARD REVIEW PLAN SECTION 6.5.5, " PRESSURE SUPPRESSION P0OLS AS FISSION PRODUCT CLEANUP SYSTEMS" The enclosed generic requirements review package contains a proposed new standard review plan section which, if approved, would add a procedure for establishing the fission product retention capabilities of BWR pressure suppression pools. This is a category 2 action.

• The proposed item is one of the near-term items discussed and scheduled in i

an information paper transmitted to the Commission on February 28, 1986 by the EDO entitled " Implementation Plan for the Severe Accident Policy State-ment and the Regulatory Use of New Source-Term Information (SECY 86-76). The proposed section does not place any requirement upon licensees, since no credit for fission product retention has previously been allowed in any operating license review. Licensees may opt for such credit, however, by appropriate license amendments. Its acceptance criteria and review pro-cedures contain three features: (1) stated values for pool decontamination factors, such that licensees or applicants claiming minimal credit need not perform computer calculations, (2) technical specification limits on drywell l l leakage which are reviewed under SRP 6.2.1.1'.C are also used to establish pool bypast: rates, and (3) when the proposed new section is used to set retention (redit, acceptance criterion 5 of SRP 6.5.1 is not to be applied. This last position is needed to prevent the use of SRP 6.5.1 to cowngrade an existing standby gas treatment filtration system from being an effective engineered safety feature as defined under the testing guidance of Regulatory Guide 1.52. t The net effect of the proposed new section for existing licensees is a possible relaxation in current staff positions which has no significant detrimental effect on safety, but which provides more flexibility and some potential cost savings to the industry in meeting the regulations. Since fission product i l cleanup credit for BWR suppression pools was reviewed and approved by the staff j for the GESSAR application, the effect of this SRP section will also be to provide uniform and consistent guidance to the staff for the review of this area. An earlier draft of the attached package has been reviewed by the ACRS, and their comments have been accommodated. After any further changes arising ~ from CRGR } considerations, the proposed new section will be published for public comment. l MllDMI h

..e .s DEC 2 2 M i James H. Sniezek 2 1 h 'the proposed new section reters to the SPARC code, which is a part of t e source-term code package developed by RES contractors. This code has only recently been updated to. treat iodine vapor in addition to aerosols. A ' Brookhaven National Laboratory report is enclosed as a technical t1nding - document. A reference appearing 'in the proposed section has not yet been J printed, and'a preprint has also been enclosed. These enclosures are intended i L to assist CRGR and ACRS members in their cons 1 aeration of the~ general subject of pool retention, and contain no new guidance or criteria. A regulatory 1 I analysis, as specified in NUREG/BR-0058, Rev. 1, 1s also encloseo. If,the proposed section were applied using current Regulatory Guide 1.3 source term i assumptions, tnere would be some net reduction in requirements. On the basis, j of the assessment of this reduction in the enclosed regulatory analysis, we l conclude that public nealth and safety would be adequately protectea 1t the proposed new section were implemented. The cost savings to an operating plant made possible by this net reduction in requirements, however, would be small. Committee consideration of this matter by January 15,198/, is requested. i hb e Harold R. Denton, Dire or Ottice of Nuclear Reactor Regulation i tnclosures: i As stated 4 l

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• Y Q DEC 2 21986 i

MEMORANDUM FOR: James H. Sniezek, Deputy Executive Director for Regional Operations and Generic Requirements FROM: Harold R. Denton, Director Office of Nuclear Reactor Regulation PROPOSED NEW STANDARD REVIEW PLAN SECTION 6.5.5,

SUBJECT:

~ " PRESSURE SUPPRESSION POOLS AS FISSION PRODUCT CLEANUP SYSTEMS" i The enclosed generic requirements review package contains a proposed new standard review plan section which, if approved, would add a procedure for establishing the fission product retention capabilities of BWR pressure suppression pools. This is a category 2 action. The proposed item is one of the near-term items discussed and scheduled in an information paper transmitted to the Commission on February 28, 1986 by the ED0 entitled " Implementation Plan for the Severe Accident Policy State-ment and the Regulatory Use of New Source-Term Infonnation (SECY 86-76). The proposed section does not place any requirement upon licensees, since no credit for' fission product retention has previously been allowed in any operating license review. Licensees may opt for such credit, however, by appropriate license amendments. Its acceptance criteria and review pro-cedures contain three features: (1) stated values for pool decontamination factors, such that licensees or applicants claiming minimal credit need not perform computer calculations, (2) technical specification limits on drywell leakage which are reviewed unoer SRP 6.2.1.1.t are also used to establish pool bypass rates, and (3) when the proposed new section is used to set retention credit, acceptance criterion 5 of SRP 6.5.1 is not to be applied. Thic last position is needed to prevent the use of SRP 6.5.1 to downgrade an existing standby gat treatment filtration system from being an effective engineered safety feature as defined under the testing guidance of Regulatory Guide 1.52. l The net effect of the prcposed new section for existing licensees is a possible i relaxation in current staff positions which has no significant detrimental effect on scfety, but which provides more flexibility and some potential cost savings to the industry in meeting the regulations. Since fission product cleanup credit for BWR suppression pools was reviewed and approved by the staff for the GESSAR application, the effect of this SRP section will also be to provide uniform and consistent guidance to the staff for the review of this area. An earlier draft of the attached package has been reviewed by the ACRS, and their comments have been accommodated. After any further changes arising from CRGR considerations, the proposed new section will be published for public comment. dA V 10ln. ao!L typ omc4 umwep DATEf """*""'""cu 2' OFFICIAL RECORD COPY

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Iy I 1 i* i+ UNITED STATES -~ 8 't. NUCLEAR REGULATORY COMMISSION g p WASHINoTON, D, C. 20555 %,*,I / 2 6 NOV 1986 MEMORANDUM FOR: Zoltan R. Rosztoczy,,,, Chief j Regulatory Improvements Branch l Division of Safety Review & Oversight i FROM: Karl Kniel, Chief Safety Program Evaluati' n' B/anch o Division of Safety Revision & Oversight I

SUBJECT:

CRGR PACKAGE - PROPOSED NEW SRP SECTION 6.5.5, " PRESSURE SUPPRESSION POOLS AS FISSION PRODUCT CLEANUP SYSTEMS" l In accordance with your requests, we have reviewed a parallel concurrence copy of the proposed CRGR Package. Based cn our review, we will concur on the original concurrence package. 4 Karl-Kniel, Chief Safety Program Evaluation Branch Division of Safety Review & Oversight cc: L. Soffer R. Frahm R. Emrit ~ J. Read R. Riggs () f l % % 8 .\\ D N hh { f. d O L (. O \\V

6 i f j 1 ) James H. Sniezek 2 j The proposed new section refers to the SPARC code, whicn 1s a part of the source-term code package developed oy RES contractors. This code has only recently oeen updated to treat iodine vapor in addition to aerosols. A Brookhaven National Laboratory report is enclosed as a technical finding document. A reference appearing in the proposed section has not yet been ] printed, and a preprint has also been enclosed. These enclosures are intended i to assist CRGR and ACRS memDers in their consideration of the general subject i of pool retention, and contain no new guidance or criteria. A regulatory analysis, as specified in NUREG/8k-0058 Rev. 1, is also enclosed. If the proposed section were applied using current Regulatory Guloe 1.3 source term assumptions, there would be some net reduction in requirements. On tne basis I of the assessment of this reduction in the encloseo regulatory analysis, we concluoe that public health and safety would be adequately protected if the { proposed new section were implemented. The cost savings to an operating j plant made possible by this net reduction in requirements, however, would j be small. Committee consideration of this matter by January 15, 1987, is requestea. 1 j Harold R. Denton, Director Office of Nuclear xeactor Regulation i' Enclosures

• As stated Distribution:

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[_. Q s u w ~ l MEMORANDUM FOR: James H. Sniezek, Deputy Executive Director for Regional Operations and Generic Requirements FROM: Harold R. Denton, Director Office of Nuclear Reactor Regulation

SUBJECT:

PROPOSED NEW STANDARU REVIEW PLAN SECTION 6.5.5, " PRESSURE SUPPRESSION P0OLS AS F1SSION PRODUCT CLEANUP SYSTEMS" r l 1he enclosed generic requirements review package contains a proposed new stanoard review plan section which, if approved, would add a procedure for establishing the fission product retention capabilities of BWR pressure suppression pools. This is a category 2 action. l The proposed item 1s one of the near-term items discussed and seneduled in i an information paper transmitted to tne Commission on February 28, 1986 by the EDO entitled " Implementation Plan for the Severe Accident Policy State-ment and the Regulatory Use of New Source-Term Information (SECY 86-76). l Tne proposed section does not place any requirement upon licensees, since l no crea1t for fission product retention has previously been allowed in any operating license review. Licensees may opt for such credit, however, by l appropriate license amendments. Its acceptance criteria and review pro-cedures contain three features: (1.)' stated values for pool decontamination factors, such that licensees or applicants claiming minimal credit need not performcomputercalculations,(2.) technical specification limits on drywell leakage which are reviewed under SRP 6.2.1.1.C are also used to establish pool bypass rates, and (3.) when the proposed.new section is used to set retention creait, acceptance criterion 5 of SRP 6.5.1 is not to be applied. This last position is nteded to prevent the use of SRP 6.5.1 to down grade an existing standby gas treatment filtration system from being an engineered safety feature. The net effect of the proposed new section for existing licensees is a possible relaxation in curr ent staft positions which has no significant detrimental effect on safety. Dut which provides flexibility and some potential cost savings to the industry in meeting the regulations. Since fission product cleanup crealt for BWR suppression pools was reviewed ana approved by the staff for the GESSAR application, the effect of this SRP section will also be to provide uniform and consistent guidance to the statt for the review of this area. An earlier dratt of the attached package has been reviewed by tne ACRS, and their ~ comments have been accommodated. After any further changes arising from CRGR considerations, the proposea new section will be published for public comment. --Y 0 }\\ M l

~ 2 ~ estabilsh whether or not fission product scrubbing of the drywell or reactor compartment atmospnere is claimed or. required for mitigation'of oft-site consequences following a postulated accident. 2.) Design Bases A comparison is made to establish that the design bases for the suppression. pool and the drywell or reactor compartment are consistent with the assumptions made in the accident evaluations of SAR Chapter 15. 3.)- System Design The information concerning the, suppression pool is revsewed to familiarize the reviewer with the expected temperature histories, depth of fission product entry expected during postulated accidents and potential leakage paths through' i 'drywell penetrations. 4.) Testing and Technical Specifications The' details of the. applicant's proposed preoperational tests, and, at tne operating license stage, the surveillance The requirements, are reviewed unde'r section 6.2.1.1.C. results of that review are examined to assure that pool depth and amount of leakage bypassing the pool are maintained con-I sistent with the assumptions usea in assessing the pool's effectiveness in fission product cleanup. 1 L__

's, 1' .+ Proposed New Standard Review Plan'Section ( 6.5.5 PRESSURE SUPPRESSION P0OLS AS FISSION PRODUCT CLEAN-UP SYSTEMS REVIEW RESPONSIBILITIES l l Primary - Plant Systems Branch Secondary - Reactor Systems 4 I A'REAS OF REVIEW

Pressure suppression pools are reviewed under this plan only when the applicant claims credit for fission product scrubbing and retention by-the suppression pool. The pressure suppression pool and the drywell, when considered as a barrier to the release of. fission products, are reviewed to assess the degree to which fission products released during postulated reactor accidents will be retained in the I

suppression pool. Leakage paths which allow fission products to s bypass the pool are identified and reviewed, and the maximum fractional bypass leakage is obtained, for use in the evaluation of l radiological dose consequences. j ]t i l 1.) Fission Product Control Requirement 3 \\ Sections of the SAR related to accident analysis, dose calculations, and fission product control are reviewed to f i i' i

j ~ a l The bypass leakage assumed for purposes of evaluating f1ssion 2. product retention must be no less than that accepted in the review under section 6.2.1.1.C, and must be demonstrated in ] periodic tests by the license technical specifications also reviewed under that section. 3. For plants wnich have already received a construction permit, the iodine retention calculated using this section must not be used to justify removal of the standby gas treatment or other filtered exhaust system from status as engineered safety For such reviews, criterion 11.5 of SRP 6.5.1 shall features. not be applied, ana tne charcoal absorbers must be at least maintained to the minimum level of Table 2 in Regulatory Gulae 1.52, Revision 2. Acceptable methods for computing fission product retention by the suppression pool are given in Subsection III, " Review Procedures." 111 Review Procedures The first step in the review is to determine whether or not the I suppression pool 1s to be used for accident dose mitigatico If no fission product removal credit is claimed in the purposes. accident analyses appearing in chapter 15 of the SAR, no further review is required. l l 1 l

3 A 11 ACCEPTANCE CRITERIA -The acceptance criteria for the fission product clean-up function of the suppression pool.are based on the.following requirements from Appendix A of 10 CFR 50: General Design Criterion 41 (Ref.1) as related to the control A. of fission products following potential accidents. B. beneral-Design Criterion 42 (Ref.1) as related to the periodic inspection of engineered safety features. General Design Criterion 43 (Ref.1) as related to the periodic C. functional testing of engineered safety features. Where they can.De shown to be in compliance with these criteria, suppression pools may De given appropriate credit for fission product scrubbing and retention (except. for noble gases, for which no pool retention is allowed) in the staff's evaluation of the radiological consequences of design basis accidents. i i Specific criteria which must ce met to receive credit are as follows: The drywell ano its penetrations must be designed to assure 1. that, even with a single active failure, all releases from the core must pass into the suppression pool, except for small bypass leakage.

s: 6 2. Pool bypass fraction The fraction of the drywell atmosphere bypassing.the suppression pool by leaking through drywell penetrations is i .obtained as a product of the review under section'6.2.1.1.C. If.B is the bypass fraction and DF is the: time-integrated pool decontamination factor,'then the overall decontamination, D, to be reported to the Reactor,dystems Branch for use in chapter 15 dose calculations may De taken as: DF g=_ 6 (DF-l} I t or 1--, G + 1 B_ 0 OF The reviewer should clearly distinguish that fraction'of B which may be further treated by the standby gas treatment system from that fraction of B which also bypasses secondary containment. 3, Other containment atmosphere clean-up systems Plants having dryweli or containment spray systems for which Tission product cleanup credit is claimed are reviewed l separately under section 6.5.2, and credit for noth suppression pool and spray cleanup can be given as a result of the separate reviews. J ) ] i \\ _ _:_J

5 If the suppression pool is intended as'an engineered' safety feature L tor.the mitigation of off-site doses, then the rev1 ewer estimates its effectiveness in removing _ fission products from fluids expelled from the drywell or directly from the pressure vessel through the 1;- depressurization system. 1. Pool decontamination factor The decontamination factor (DF) of.the pool is defined as the ratio of the amount of a contaminant entering the pool' to the. amount leaving. Decontamination factors for each fission form as functions of time can be calculated by the product SPARC code (Ref.2), and this calculation should be performed l 9 whenever the pool design.is judged by the reviewer to differ h significantly from those found acceptable as fission product cleanup systems in past reviews. If, however, the time-inte-gr,ated DF values claimed by the applicant are 10 or less for particulate and 100 or less for iodine vapor the applicant's values may oe accepted without any need to perform calculations (Ref. 3). A DF value of 1 (no retention) should be used for noble gases and, unless the applicant demonstrates otherwise, for organic locides as well. If calculation of fission product decontamination is done using the SPARC code, the review should be coordinated with the Reactor bystems Branch, which is responsible for establishing the accidentassumptionsneededtoassembletheinputf5rthecalcu-l lations.

8 Tunction can be accomplished assuming a single tailure. The l' applicant's proposed program for preoperational and surveillance tests will assure a continued state of readi-ness, and that bypass of the pool is unlikely to exceed the assumptions used in the dose assessments of Chapter 15. The staft concludes that the suppression pool is acceptable as a fission product cleanup system, and meets the requirements of General Design Criteria 41, 42 and 43. V IMPLEMENTATION Except in those cases in which. the applicant proposes an acceptable alternative method for complying with the specified portions of the Commission's regulations, the methods described nerepin are to be used by the staff in its evaluation of conformance with Commissions regulations. Implementation of the acceptance criteria of subsection 11 of this plan is as follows: (a.) Operat 1g plants and OL applicants need not comply w1th the provision of this review plan section. CP applicants will be required to comply with the provisions (o.) of this revision.

.p e 4. Technical Specifications The. technical specifications are reviewed to assure that they require periodic inspection to confirm suppression pool depth and surveillance tests to confirm drywell leak tightness consistent with the bypass fraction used in computing the overall decontamination. Technical specification review is coordinated with the Facility Operations Branen as provided in NRR Office Letter No. 51. IV EVALUATION FINDINGS The reviewer verifies that sufficient information has been provided by the applicant and that the review and any calculations support conclusions of the following type, to be included in the staff's Safety Evaluation Report: 4 We have reviewed the fission product scrubbing function of the pressure suppression pool and find that the pool will reduce the fission product content of the steam-gas mixture flowing through the pool following accioents which blow down througn the suppression pool. We estimate the pool will decontaminate the flow by a factor of for molecular iodine vapor and by a factor of tor particulate fission products. No significant pool decontamination from noble gases or organic iodides will occur. The system is largely passive in nature, and the active components are suitably redundant such that its fission product attenuation

3 9 VI' REFERENtEd'. 10 CFR Part 50, Appendix A, General Design Criteria 41, 1.. " Containment Atmosphere Clean-up", 42; " Inspection of Containment Atmosphere Cleanup Systems", and 43, " Testing of Containment Atmosphere Cleanup System". P.C. Owczarski, R.I. Shreck and A.K. 'Postma, " Technical 2. Bases and Users Manual for the Prototype of'a Suppression- [. Pool Aerosol Removal Code (SPARC)', NUREG/CR-3317,1985. P.C. Owczarski and W.K. Winegardner, " Capture of Iodine 3. in. Suppression Pools", 19th DOL /NRC Nuclear Air Cleaning Conference, Seattle, 1986. e I L

2 Standard Review Plan 6.5.3, " Fission Product Control Systems and Structures," contradicts Regulatory Guide 1.3 by stating that suppression pools may be considered as fission product control systems, although no guidance or reference is supplied as to methods to be used in their review. In NUREG-0979, supplement 4, " Safety Evaluation Report related to .the final design approval of the GESSAR II BWR/6 Nuclear Island Design," the staff agreed to consider suppression pool retention in any application referencing the approved design. Revisions prompted by new source term information and the replacement of TID-14844 by more realistic accident assumptions will result also in the revision of Regulatory Guide 1.3. Regardless of whether an accidental release is assumed using the current Regulatory Guide 1.3 or using the most modern methods, it is an undue conservatism to ignore the capability of the suppression pool to mitigate off-site dose consequences, provided that recognition of such capability does not degrade safety. The effectiveness of suppression pools in retaining gaseous iodine and particulate matter varies markedly with the conditions under which these materials are swept into the pool. While the overall effects of such variation can be calculated for any given postulated accident, this calculation would be uncertain in its predictions of the relevant conditions and would be very expensive to perform. It would be i l inappropriate to solve the problem of ignoring suppression pool - effectiveness by replacing it with a required set of calculations that are

REGULATORY ANALYSIS OF THE REVIEW 0F l SUPPRESSION P0OLS AS FISSION PRODUCT CLEANUP SYSTEMS i 1. Statement of the Problem Regulatory Guide 1.3, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors,". states, as. Regulatory Position C.1.f,'that "No credit is'given 'for retention of iodine in the suppression pool." Before the time this . guide was first published, November 2, 1970, experiments had demonstrated the efficacy of suppression' pools in removing iodine in several chemical forms from air-steam mixtures. The adoption of Regulatory Position C.1.f, therefore, was deliberately conservative. Factors which may have influenced its adoption are: (1) drywells are generally leaky, permitting significant bypass of the suppression pool. (2) suppression pool retention of fission products varies markedly with conditions pertaining during the accident, and would have required more complicated models than any being used in 1970. (3) because of heavy reliance on standby gas treatment systems, suppressionpoolcreditwasnotneededbyboilingwaterheactorsto meet the dose guidelines of 10 CFR 100.

4 SRP 6.5.3 could be revised to remove the statement allowing pool credit, and the GESSAR 11 SER could be amended to retract the earlier position. This course would remove the inconsistency between the SRP and Regulatory Guide but, in addition to ignoring the large volume of research data supporting pool credit, would provide an undue degree of conservation to the staff's review and be contrary to commission policy. (Goal 2.4, NUREG-0885, Policy and Planning Guidance, 1986) The alternative selected is to propose an additional review plan section which would prov1de a consistent use of the available data without de-gradation of safety. In preposing review procedures, two decisions were made concerning the means by which the review could be simplified. (a) time-averaged decontamination factors (DF) were introduced, (b) minimum DF values were stated, such that only applications claiming larger 0F's would require plant specific computer runs. These decisions were prompted by practical considerations in conducting reviews; not taking the proposed course would have required great computer i expense in any review. If a novel suppression pool feature were proposed, { 1 ) such as, for example, a chemical additive or increased submergence depth of the downcomers or quenchers, the Source Term Code Package computer codes could be run to quantify the* effectiveness of the pool. Use of the Source Term Code Package costs about 525,000 per accident sequence.

3 c. impractical'for use in assessing effectiveness. To avoid this further problem,'the present proposal takes a narrow interpretation by replacing the undue conservatism of omitting credit in-favor of moderately conser-vative simplifications. 2. Objectives The objective of the proposed action is to establish _the degree to which suppression pools can be considered as fission product cleanup systems and by revising the Standard Review Plan (SRP) to include procedures and criteria for, suppression pool design evaluation. 3.. Alternatives The existing SRP 6.5.3, " Fission Product Control Systems and Structures," in 11.5, states that " Fission product retention credit assumed by the applicant for other systems, e.g., pressure suppression pools, may be acceptable provided that justification is supplied by the applicant." i 'This provision has been applied, so far, only in the review of the GESSAR-II application. The existing SRP, however, contains no procedures for reviewing pressure suppression pools. One alternative to the proposed new'section, therefore, would be to continue to review pools on a case-by-case basis. This course would not consistently apply computer code and model valldation experiments which have been devised for pur-l poses of developing a means of calculating pool retention of fission products.

-6 ' Apart from.the containment buildings themselves, the most important acci-dent off-site dose mitigation features of boiling water reactor plants under the SRP.are the standby gas treatmr.nt systems (SGTS). These filtered exhaust systems are designed to have maximum effectiveness against the forms When reviewed of fission product iodine assumed to be released by TID-14844. 'against the fission product releases predicted by the new source term code package, however, suppression pools are capable of a high degree of retention of fission products. The proposed change will focus attention on sup-pression' pools-as dose mitigation features, and as a means of providing defense-in-depth in fission product mitigation capability. The development of regulatory requirements for suppression pools might lead existing licensees to upgrade the quality of drywell penetrations, as part of measures to minimize pathways bypassing the pools. Drywell pene-trations are already subject to leak testing at each refueling under SRP. 6.2.1.1.C. At present, Regulatory Guide 1.3 assumes that 22.75% of the core iodine inventory as molecular iodine, 1.25% as particulate and 1% as organic iodide are available for release from containment. Typical standby gas treatment systems (SGTS) serving BWR secondary containments as filtered exhaust systems are maintained at 99% efficiency against all of these forms, i.e., after one-hundred-fold decontamination 0.25% of the core iodine is exhausted into the environment as the sum of the primary containment and main steam line isolation valve leak rates following a DBA-LOCA. I l I

N. 4, a : 5 The minimum DF values chosen are designed to be sufficiently small that no. accident sequence is likely to be found to have a smaller time-averaged value, even allowing a margin of safety for uncertainties. 4. Consequences By resolving the contradiction between Regulatory Guide 1.3 and SRP 6.5.3 in favor of the former, the staff would be denying a large body of evidence proving the efficacy of suppression pools in retaining fission products. This might be defensible on the grounds of being conservative, but would not permit the realistic consideration of core melt accidents as they are currently being modeled to be used'1n licensing decisions. By continuing the present situation, i.e., taking no action, the con-tradiction would remain. Licensees could request suppression pool credit, based on the staff's GESSAR II statement, but the staff would have no consistent guidance for performing the review, and would be reduced to either accepting or rejecting the licensees' submittal, or running the source term code package repeatedly. l The consequences of the proposed new section would be the effect that increased pool credit would have upon the efficiency required of other fission product control systems in order to meet the dose guidelines of 10 CFR Part 100. A licensee could request pool credit to justify a relaxation ) of the maintenance and surveillance requirements placed on other systems. l

1 8 Guide 1.3, even if molecular and particulate iodine forms are totally absorbed by the pool. A'15-fold re' duction in SGTS effectiveness, i.e., from a penetration test of 1% or less to one of 15% or less, would reduce its organic iodide absorption efticiency from 99% to 85%. Unfortunately, the SRP section dealing with SGTS review, 6.5.1, states that systems requiring iodine f absorption efr1ciencies of less than 90% may be reviewed under SRP 11.3. I Charcoal absorbers reviewed under SRP 11.3 may follow Regulatory Guide 1.140 rather than 1.52, and are not built or maintained to engineered safety teature standards. To prevent use of suppression pool credit to justify not maintaining and testing SGTS absorbers to Regulatory Guide 1.52 criteria, prior to revision of SRP 6.5.1, explicit mention of SGTS surveillance tests has been added as a criterion. A typical SGTS contains about $20,000 of impregnated charcoal per train, with a comparable additional labor cost for renewing and testing if filter If the surveillance replacement is needed to pass a surveillance test. test criteria of a SGTS were relaxed, charcoal change-out would be 4 required less often, perhaps reducing maintenance by of the order of 10 dollars per year. It is also possible that a licensee might wish both suppression pool and maxim 6A 5GTS creWits, while requesting an increase in allowable containment leakage. Again, a very large saving in the costs of containment integrated leak rate tests would not be expected, since the i large degree of iodine fission product retention would not be associated I I

i 1 Against the same release to containment, the minimum decontamination ~ ~ factors in the proposed SRP 6.5.5 would reduce the 25% of the core iodine inventory available for primary containment leakage to 3.5%, assuming 10% suppression pool bypass leakage. For 1% pool bypass leakage, 1.6% of the core iodine inventory would be computed as available. { Using the assumed release in Regulatory Guide 1.3, and obtaining sup-pression pool scrubbing credit with 10% bypass, a typical BWR could meet 10 CFR Part 100 thyroid dose guidelines and still reduce the plant SGTS -efficiency from 99% to 95%. It should not, however, be assumed that by reducing bypass leakage and claiming suppression pool credit a lice'nsee i could greatly reduce the surveillance testing requirements of their SGTS. Other design basis accidents, for example the fuel handling and instrument line break accidents, also require the use of the SGTS to meet the acceptancegriteriaofSRP 15.6.2, 15.6.5 Appendix B, and 15.7.4. 131 For a typical plant, the release of 3000 Ci of 1 would lead to. dose This amount consequences in excess of the guidelines of 10 CFR Part 100. is equivalent to only a few parts per million of the core inventory of iodine fission products. For a typical BWR, a million-fold reduction in iodine 15 mostly achieved by a low leakage containment (0.5% per day in 2 hours leaks 4X10-4) and to a lesser extent by the SGTS (10"* penetration by iodine). Since suppression pools are virtually useless against organic iodide, and since it is not feasible to eliminate bypass completely, overall decontamination factors of more than about 15 cannot be practically achieved using the current iodine chemistry assumptions in Regulatory

10 5. Decision Rationale Strategic goal 2.4 of the NRC Policy and Planning Guidance,1986, lists as objectives the completion of the reassessment of source terms and the implementation of appropriate revisions ~in staff practices. The source term revisions will involve many related changes to the SRP and regulatory guidance, and may also include rulemaking and revision of existing regulations. The proposed action is perceived as an early step in this process, since it will put in place the review procedures and criteria necessary for considering the mitigation of new source terms by suppression pools. TheproposedsectionIsequallyapplicabletothesourcetermassumptions contained in Regulatory Guide 1.3 and to the fission product releases cal-culatedbythepourceTermCodePackage. For both applications the pro-z posed action ofters the following advantages: 1.) Suppression pool fission product retention can be assumed to be The described by conservatively chosen decontamination tactors. use of these factors avoids the large expense of computer analysis needed to quantify suppression pool response using the available computer codes. As discussed earlier, very large decontamination factors can be calculated, but the net effective decontamination j { achievable is limited by the possibility of pool bypass leakage. 1 .-- _ --- --____________ _ Q

.o. 1 with any change in the postulated noble gas releases during a LOCA. Licensees electing this course would be limited by the 10 CFR Part 100 guideline for whole body doses at the low population zone boundary over the course of the ar.cident. For most BWRs, a doubling of the containment I leak rate would bring the noble gas release consequences to the guideline, although for some plants having favorable meteorological parameters and large low population zones several-fold increases would still meet the guidelines. While granting credit for suppression pool scrubbing, as proposed, would allow the deterministic licensing calculations of accident dose to be more easily met, the primary thrust of the change will be to allow greater BWR containment leak rates and more noncondensible accident fission products past SGTS filters. That is, existing BWR containment leak rates of about 0.5 volume percent per day maybe increased to as much as 5 volume percent per day, and 99 percent elemental iodine filter efficiencies maybe reduced to 90 percent. The change, therefore, may result in increases in the quantities of fission products postulated to I be released during design basis accidents. However, regulatory guidelines would still be met, and the change in risk is expected to be very small since the bulk of public risk is attributed to accidents in which the containment fails or is bypassed (i.e., severe accidents not designbasisaccidents).

~11 2.) Existing. plants have the possibility of reducing maintenance costs for their charcoal absorbers by being able to retain the absorbent 4 for longer periods of time between changes. The cost savings of these advantages would vary with the degree to which licensees and aplicants. elected to claim suppression pool fission product cleanup credit, and the number and diversity of accident' sequences necessary to represent the effectiveness of' the pool. While releases of fission products as assumed in Regulatory Guide l'.3 are effectively reduced by filtered exhaust systems, the releases calculated for many accident sequences by the Source Term Code-Package are more effectively reduced by suppression pool scrubbing. By adding guidance'for the review.of suppression pools as fission product' cleanup systems in the form proposed, conservative but .s, appropriately realistic-credit would be assessed without significant loss of the safety afforded by existing filtered exhaust systems. 6. Implementation l The proposed action requires no action of existing licensees, except as j l-they might voluntarily elect to reanalyze the accident consequences and 1 l l This action submit an FSAR amendment to reduce reported iodine doses. .3 L would take effect upon publication of the proposed revision. I I L - - - - --_____

k, b I 19th DOE /NRC NUCLEAR AIR CLEANING CONFERENCE The SPARC simulation of pool scrubbing first relies on a descrip-tion of the hydrodynamics of gases entering a water pool at a sub-merged depth. The hydrodynamics of the vent exit region are very important as are the hydrodynamics of the bubble swarm rise to the pool surf ace. Particle scrubbino A number of phenomena have been identified as contributors to the particle scrubbing process. These are: particle inertia at the vent exit j e .e bubble inertia at the vent exit j .e steam condensation at the vent exit l temperature gradient at the vent exit l e steam formation during bubble rise l e e particle growth l e bubble circulation during swarm rise l e bubble coalescence /redispersion during swarm rise. The above phenomena are quantitatively modeled in SPARC for their roles in the particle capture mechanisms at the vent exits (centrifu-gal scrubbing, inertial deposition, steam condensation and thermo-I phoresis) and during swarm rise (centrifugal scrubbing, gravity set- ) tling and Brownian diffusion). The centrifugal scrubbing here refers l to deposition of particles at curved gas-liquid surfaces caused by the acceleration of particles in the radial direction as a result of tan-gential surf ace velocities. q Iodine Behavior A number of aspects of iodine behavior are related to its capture in suppression pools. These aspects can be identified in'three regions of the flow of gases. The first region is the flow of iodine ) species in the core-melt of f gases in the reactor primary system. The second is the vent exit region in the pool and the third is the bubble swarm rise region in the pool. In the primary system, where gases are hydrogen' and steam and iodine species can be I, organic lodides, HI, and particulate iodides 2 such as csI, conditions can exist that f avor the coniplete removal of the volatile inorganic species from the gas phase. These favorable conditions consist of a sufficiently low temperature so that alkaline aerosol particles can exist as a liquid or partially liquid phase. Alka hydroxides such as Cs0H have this property in the vicinity of This liquid phase can be highly reactive with the volatile 300 We speculate that solid CsOH can be reactive with species HI and I2 these species as well. The SPARC code has a subroutine that allows the user to switch on this iodine absorption process in the primary system. The process is modeled as a continuous plug-flow reactor where spherical aerosol particles absorb elemental iodine at a rate in the gas phase around the partic-controlled by the diffusion of I2 les. Although not modeled, HI would behave similarly to I2, but with a slightly higher diffusion coefficient. The results of using this subroutine are discussed in III. Accident Secuence Results.

7-4 19th DOE /NRC NUCLEAR AIR CLEANING CONFERENCE CAPTURE OF. IODINE IN SUPPRESSION POOLS P. C. Owczarski and W.. K. Winegardner Pacific Northwest Laboratory *, Richland, Washington Abstract The effectiveness of suppression pools in capturing airborne iodine species was investigated. A computer code was used to simulate the scrubbing of particulate iodide, vapor elemental iodine, and vapor organic iodides. For a typical postulated severe core damage accident sequence, suppression pools were effective scrubbers of elemental iodine if the pool was alkaline or dilute in iodine and of particles >l.5 um mass median diameter. Little scrubbing of organic iodide species occurred. An absorption model shows that elemental iodine can be absorbed by wet alkaline droplets before the droplets encounter the suppression pool. Thus, the iodine removal effectiveness of the pools is likely to be controlled by particle scrubbing. I. Introduction The estimation of airborne source terms in postulated severe core melt accidents required the evaluation of the. responses of nuclear reactor Engineered Safety Features (ESF) under accident conditions. As part of this evaluation, the Pacific Northwest Laboratory (PNL) has been studying the aerosol capture effectiveness of Boiling Water Reactor (BWE) pressure suppression pools.** The initial work assumed that fission product iodine would exist as CsI in the aerosol leaving the reactor primary system. Concern remains that other chemical spec-ies of iodine might exist, notably I2 and organic iodides (represented Continuing work reported here shows that the scrubbing by CH I). 3 ef f ectiveness quantified by decontamination f actors (DFs)' of the pool varies dramatically for the three chemical species. To estimate the pool g7 rubbing effectiveness on particles, {pp with existing published data $gich has been partially validated developed the SPARC code The SPARC code was then modified to 2 and CH I scrubbing. An additional function of SPARC com-include I 3 by particles containing deliquescent CsoH. putes the absorption of I2 II. Technical Bases Summary This section summarizes the technical bases for the models in previously discussed ( sost of the particle scrubbing models have been SPARC. The bases for and will be briefly repeated nere. Then the bases for iodine scrubbing will be discussed.

• Pacific Northwest Laboratory is operated for the U.S. Department of Energy by Battelle Memorial Institute.

j

• • Work supported by the U.S. Nuclear Regulatory Commission under Contract DE-AC06-7 6RLO 18 30, NRC FIN B2 4 4 4.

DF = mass flow rate of a fission eroduct into cool + mass flow rate of that fission product out of pool

o l l l 19th DOE /NRC NUCLEAR AIR CLEANING CONFERENCE scrubbing..No plans exist to measure volatile iodine scrubbing in large-scale experiments. Accident Parameters SPARC can be used to analyze pool scrubbing during the course of an accident scenario. A number of accident parameters must be defined for each time step when pool scrubbing is important. The most impor-tant set of parameters is the particle size distribution. The SPARC input parameters are listed below: Pool e noncondensable gas flow rate into pool e noncondensable gas composition e steam flow rate into pool e pool depth,. temperature e pool size, configuration e pressure above pool e pool composition (surfactants) e vent exit configuration Aerosol Particles e mass flow rate e aize distribution e density / shape factors e solubility in water (and fraction of soluble alkaline materials) Iodine e mass flow rates of each iodine species e temperature and pressure of primary system These parameters are defined for an example accident scenario in the next section. III. Accident Sequence Results To examine the behavior of iodine specie's in the pool, we used a specific postulated accident sequence to establish the pool, flopg)and fission product characteristics for this study. The Tc sequence for a Mark I BWR was' selected as a representative accident. In this accident, a transient event was followed by control rod insertion i f ailure, but emergency core cooling systems operated. However, the i reactor power level exceeded the cooling capability of the suppression pool. Overpressure f ailure of the containment occurred followed by stoppage of reactor vessel coolant flow. The core heated up and melted, releasing fission products into outflowing steam and hydrogen. During this melt release, these gases and fission products flow f rom ) the core through the primary system and suppression pool. It is for this period, f rom 134 to 168 min af ter the initiating transient, that we have analyzed the pool scrubbing ef festiveness of iodine species as

As the gases 1savo the primary cystcm, they enter the pool at a dspth through a spacific vant typa. In the ragion of this vent,.the . gases try to equilibrate with the thermodynamic conditions of the pool at the vent depth. This equilibration process frequently results in steam condensation and scrubbing of particles. In SPARC, this conden-sation results in some deposition of I 2 and CH I, but is limited by 3 l the species solubility at the interf ace. Af ter the initial gas globules at the vent break up into the ris- .ing bubble swarm, the SPARC code assumes that bubble circulation con-tinually renews the bubble interface and that the film theory of mass transfer resistance holds on both sides of the interface. The equili-brium boundary conditions at the interface for the two volatile iodine species are: [TI (aq))1 = H(I2)lI2(gas))1 2 ~ and. [CH I(aq)]i = H(CH I)(CH I(gas)]i 3 3 l where (TI2(aq))1 = total liquid molar concentration of iodine at the interface as I 2- [I2(gas))1 = interf acial gas molar concentration of I2 and H(I2) = iodine partition coefficient. Similar definitions hold for CH I controlled by the fast reactions:I5)The aqueous chemistry of iodine is 3 I2(gas) = I (aq) 2 I (aq) + I~ = I ~ 2 3 I2 (aq) + H O = H+ + I~ + HIO 2 I (aq) + H O = H 0I+ + I~ 2 2 2 H O = H+ + OH~ 2 By using the equilibrium constants for the above five reactions, the partition coefficient is quantitatively defined if mass balances of all lodine species and H+ and OR~ are maintained. The value of H(CH I) is(gtained in a simpler way using solubility and vapor pres-3 sure data. SPARC validation The particle capture models in SPARC e been ' partially vali-dated with data as they become available. The iodine capture mod g are validated by small-scale tests. The data of Diffey et al. compare favorably with SPARC calculations for both I2 and CH I 3

~- 19th DOE /NRC NUCLEAR AIR CLEANING CONFERENCE 8 10 j i I l I l~ Residence Time I 'i l l I l I I iO' r l l 1 1 I M I 1 l [ l I .E l I i e st EI l 3 10,."- { l le .g is

=

a 9 I i l, I,,, g 2: t o 12 I 1 dg lu$ ./ ~ i l l t o*' - [.5 1 I Il .E [ l Half Life l* e I 4 I I } l I 1 l 1 10*8 E l l 2 I l l I I 1 I l 10~' 'I I I I I I I I - 130 135 140 145 150 155 160 165 470 Accident Time, min FIGURE 1 COMPARISON OF THE HALF-LIFE OF I 2 EXISTENCE IN THE PRIMARY SYSTEli WITH THE RESIDENCE TIME OF GASES IN.THE PRIMARY. SYSTEM (CASE 1) The scrubbing of iodine species is portrayedin Figures 2 and 3. In the first figure, the instantaneous I csI, and CH I DFs are plot-ted versus time during the melt release.2,In case 1, where CsI is the 3 lodine species, the scrubbing of csI generally increases in time - because of the gradually increasing particle size until 154.5 min, where particle size stabilizes until the end of the melt release. However, steam and. hydrogen gas flow increase dramatically at this y point, and as a result the inertial particle capture mechanism at the vent exit increases the DF. In case 2, where the iodine is. elemental and the pool receives no alkaline particles, the iodine se' rubbing is represented by the I2 curve. Here the I2 flow rate is fairly high until 148.5 min, then the rat'e (and incoming I2 concentration) decreases. These decreases cause the pool scrubbing to become less effective at the iodine concentrations of the pool. However, in Case 3 the pool was allowed to increase in pH from incoming Cs0H particles, the time history of iodine DFs yas dif f erent. Then the instantaneous DFs were never less than 3 x 10 during the accident. The CH I curve 3 a .________.___.___._______________w

] 19th DOE /NRC NUCLEAR AIR CLEANING CONFERENCE ) well as any prior reaction of vapor elemental iodine with alkaline aerosol droplets in the primary system upstream of the pool. The data pertinent to the SPARC analysis are summarized here: Aerosol flow rates ranged from 110 g/s at the beginning of core melt to less than 1 g/s at the end. CsOH ranged from 60 to 10 wtt of this aerosol. Iodine flow rates cent of this iodine was exam. ranged from 9 to 1 g/s. Ninety-nine per-ined as either I cles and 1% of the lodine was assumed to be as CH I2 vapor or as CsI parti-i began at 1.5 pm mass median diameter' and finishe2 a. Particle sizes t 2.7 um. Geomet-i ric standard deviation's 'of the aerosol distribution gnd the aerosol particle density remained constant at 1.7 and 3 g/cm, respectively. l maintained a temperature range of 340 to 360 C. Cases from the steadily ) Steam flow began at 1300 g/s and ended at 8 g/s. Hydrogen flow began at 170 g/s and ended at 110 g/s. These gas and aerosol flows entered the suppression pool through 13 t-quenchers at 12 f t submergence. With the above aerosol and flow specifications, the SPARC code was run as three independent cases: Case 1 where the elemental iodine was allowed to be absorbed by alkaline aerosol droplets in the primary l system, Case 2 where the alkaline materials were absent in the aero-sol, and Case 3 where the CsOH was not allowed to react with the I2 in the primary system. In Case 1, lodine was present as CsI in the par-ticulate mass. In case 2, the iodine remained as vapor I 2 Also, the pool did not have the benefit of becoming alkaline from CsOH parti-cles, so Ig scrubbing was affected by an initially neutral pool (pH = 6.5 at 100 C) that became slightly acidic at the end of core melt (pH = 5.9). In Case 3, the iodine also remained as I became alkaline during core melt and reached pH =2 vapor, but the pool 8.3. The first of the SPARC results examined is the behavior of ele-mental iodine in the primary system in case 1. Here,I2 was absorbed by wet alkaline particles in the short, once-through pass of gases through the primary system. The SPARC subroutine for these calcula-tions computes the instantaneous absorption rate for the entire aero-sol cloud in terms of the half-life of I2 existence as the elemental form. The gas residence time in the Mark I primary systems, whiph splits the flow into two parallel streams, has a value of 2 x 10 /0 secongs where Q is the total primary system exit volumetric flow rate in cm /s. Figure 1 compares the half-life with the' residence time for I the melt release period of the TC sequence.. Here it is evident that sufficient residence time exists from the beginning of the melt to nearly its end to absorb virtually all of the I 2 Only at the end of fuel melt does the residence time equal the half-life of 1 2, which indicates that only one half the I2 vapor is absorbed by droplets at 1 that time. The iodine half-life increased with time because the con-decreasing the area available for absorption. centration of particles decre The gas flow rate dra-matically increased at 163 min resulting in the insufficient residence time for reaction. Use of the primary system absorption model, Case 1, was done solely for demonstrating that elemental iodine (or HI) is not likely to exist as a species in the presence of particles containing CsOH in a moist environment. 9

19th DOE /NRC NUCLEAR AIR CLEANING CONFERENCE The integration period starts at the beginning of the core melt. The initial DFs in both Figures 2 and 3 are identical. Cases 1 and 2 are again represented by the CsI curve and the I2 curve, respectively. The important observation here is that even though the case 2 pool is slightly acidic, the integrated DF is one order of magnitude greater than the Case 1 integrated DF. ge final integrated DF for I2 in Case 3 (alkaline pool) is 2 x 10 , which is more than seven orders of magnitude larger than the corresponding Case 2 DF. 10' :: 5 l l I ~ I I 1 1 I I l 10' r '8 I N l T I i .I I E1 I. ci 10' 3 =l ly E I $l I Csl f I 12 3 - 01 le e si 16 5 10' r"El l} I lw I 1.

• l 1

I 1 t l .10' :- l I i CHal i I ~ l I o ,i e i i i g 130 135 140 145 150 ,'155 160 165 170 Accident Time, min FIGURE 3 DECONTAMINATION FACTORS FOR IODINE SPECIES INTEGRATED OVER THE CORE MELT PERIOD. THE CsI CURVE REPRESENTS CASE 1, THE I 2 CURVE REPRESENTS CASE 2 AND THE CH I CURVE IS THE SAME IN ALL CASES 3 Some general conclusions can be drawn from the results of'the core melt sequence above: Pool scrubbing of iodine can be very effective when iodine is I o vapor if the pool lodine concentration is low or if the pool is 2 j alkaline.

19th DOE /NRC NUCLEAR AIR CLEANING CONFERENCE shows scrubbing initially and around 148 min, when incoming airborne concentrations are sufficient to drive CH I into the pool. Otherwise, 3 the pool is stripped of CH I (DFs (1) during periods of low CH I con-3 3 centration in the incoming gas. It should be noted that the 16 assumption does not affect the DFs for CLI I. 3 ' '! j \\ i l 1 l I 10' l I 1 l T I Csl l .I '8 1. la 10' -{ll lj j ~E lC 7 5l li a. 3 g -2 is el e c l3 E ~ $l l l

• to'

-e e E E ~61 5 -El 1 l 1 ~ l ~ l i I I. 10 :- l l ~ 1 I l CHal 1 I i I l l ll l l l l I 1 130 135 140- 145 150 155 160 165 170 Accident Tinie, min l FIGURE 2 8 INSTANTANEOUS POOL DECONTAMINATION FACTORS FOR IODINE SPECIES DURING CORE MELT IN TC ACCIDENT SEQUENCE. CsI CURVE REPRESENTS CASE 1, I2 CURVE REPRESENTS CASE 2 AND THE CH I CURVE IS THE SAMF IN ALL CASES 3 j Another representation of pool scrubbing is portrayed in t Figure 3. Here, the time-integrated DF is portrayed ove'r the core-melt period. This DF is defined (over the time period 4t) as l DF(time-integrated)1 = total mass of soecies i entering the cool in 6t. total mass of species i leaving the pool in at 4

~~ I f o.- 19th DOE /NRC NUCLEAR AIR CLEANING CONFERENCE Pool scrubbing of CH I is poor. e 3 l Pool scrubbing of iodine as particulate CsI can be fairly effec-e tive for large particles (>1.5 pm mass median diameter) e I 2 vapor cannot exist long in the presence of large numbers of wet alkaline droplets 4 The limiting pool DF would be that of particulate CsI unless sig- { e nificant core iodine (>0.1%) is converted to CH I' 3 I i References (1) Owczarski, P. C., A. K. Postma and R. ' I. Schreck. Technical 4 Bases and User's Manual for SPARC - A Suppression Pool Aerosol Removal Code. NUREG/CR-3317, PNL-4742, U.S. Nuclear Regulatory Commission, Washington, D.C.,1985. (2) Owczarski, P. C. and W. K. Winegardner. " Validation of SPARC, A Suppression Pool Aerosol Capture Model." Paper IAEA-SM-281/29 presented at IAEA International Symposium on Source Term Evaluation for Accident Conditions, October 28-November 1,1985, Columbus, Ohio, 1985. (3) ' Cunane, J. C., H. R. Kuhlman and R. N. Oehlberg. of Fission Product Aerosols in LWR Water Pools Under Severe"The Scrubbing . Accident Conditions - Experimental Results." In Proceedings: American Nuclear Society Meetino on Fission Product Behavior and- { i Source Term Research.. NP-4113-SR, Electric Power Research Institute, Palo Alto, California,1985. 4 (4) Rollet, A.-P., R. Cohen-Adad and C. Ferlin. "Le systeme eau-hydroxyde de cesium. " Comptes Rendu, Vol. 256, Pt. 6,

p. 5580 (1963).

(5) Eggleton, A. E. J. A Theoretical Examination of Iodine-Water _ Partition Coefficients. AERE-R 4 8 87. Atomic Energy Research i Establishment, Harwell, England,1967. g (6) Glew, D. N. and E. A. Hoelwyn-Hughes. " Chemical Statics of the Methyl Halides in Water." Discussions of the Faraday Society, No. 15, pp. 150-161 (1953). (7) Diffey, M. R. et al. " Iodine Clean-up'in a Steam Suppression System." In International Symoosium on Fission Product Release and Transport Under Accident Conditions. CONF-650407 (Vol. 2), Oak Ridge National Laboratory, Oak Ridge, Tennessee,1965. (6) Gieseke, J. A. et al. Radionuclides Release Under Specific LWR Accident Conditions, Volume II, BWR, MARK I Desien. BMI-2104, Volume II, Battelle Columbus Laboratories, Columbus, Ohio, 1984. l

TECHNICAL REPORT A-3788 8-1-86 EFFECTIVENESS OF BWR PRESSURE SUPPRESSION POOLS IN RETAINING FISSION PRODUCTS H. P. Nourbakhsh, R. Davis, and M. Khatib-Rahbar Accident Analysis Group Department of Nuclear Energy Brookhaven National Laboratory Upton, New York 11973 August, 1986 Prepared for U.S. Nuclear Regulatory Comission Washington, DC 20555 Contract No. DE-AC02-76CH00016 FIN No. FIN A-3788

1 -iv-J ACKNOWLEDGEMENTS i I The authors are grateful to W. T. Pratt (BNL), J. Read. L. fof fer, and Z. Rosztocsy (USNRC) for their review and many helpful remarks on this manu- - sc ript. The work reported herein was conducted under the auspices of. the - United States Nuclear Regulatory Comission (USNRC), Office.of Nuclear Reactor Regulation. 4 4 e S

.L -iii-ABSTRACT l l The effectiveness of BWR suppression pools in retaining fissiott products released during severe accidents is assessed. Scrubbing models are reviewed and sensitivities to input parameters of SPARC Computer Code used in' Source l-Term Code Package (STCP) are also discussed. An assessment of the effective l-decontamination factors for the suppression pools based on the results of l recent STCP calculations performed by BNL and BCL is also presented. .e e 8

-vi-i LIST OF FIGURES Figure Title Page 1 BWR Ma rk I con t a i nme nt sys t em.................................. 4 2 BWR Ma rk 11 cont a i nme nt sy s t em................................. 5 3 BWR Ma rk 111 cont ai nme nt sy s t em................................ 6 i 4 Comparison of SPARC calculated decontamination factors with BCL experimental values for 1/2 in diameter horizontal injector....................................................... 9 5 Variation in bubble aspect ratio with mean bubble diameter..... 11 i 6 Ef fect of' ~bubbl e di ameter and aspect ratio on DF............... 12 7 Ef f ect of bubbl e swa rm ri se vel oc i ty on DF..................... 15 LIST OF TABLES Tabl e Title Page 1 The Input Parameter Values to SPARC (Calculated by Preceding Codes in the STCP.............................................. 13 2 Effective (Time Averaged) In-Vessel Release Decontamination Factors for the Suppression Pool (Peach Bottom Mark 1)......... 18 3 Effective (Time Averaged) In-Vessel Release Decontamination Factors for the Suppression Pool (Grand Gulf Mark 111)......... 21 4 Effective (Time Averaged) Ex-Vessel Release Decontamination Factors for the Suppression Pool (Grand Gulf Mark 111)......... 21 4 r

-v-I CONTENTS Page ABSTRACT.........................................................*...~... 1ii -ACKNOWLE0GEMENTS........................................................ iv ~ LIST OF FIGURES......................................................... vi LIST OF TABLES.......................................................... vi 1. INTRODUCT10N........................................................ 1 2. DESCRIPTION OF PRESSURE SUPPRESSION P00L5........................... 3 3. PREDICTIVE METH0DS.................................................. 7 3.1 SPARC (Suppression Pool Ae rosol Removal Code ).................. 7 3.2 Expe riment al Val id ati on of the SPARC Code...................... 8 3.3 SP A RC Se n s i t i vi ty An aly s e s..................................... 8 3.4 Sol u bl e'.Ga s Sc ru bb i n g.......................................... 14 4 EFFECTIVE DECONTAMINATION FACTORS FOR THE SUPPRESSION P00L.......... 17 4.1 Peach Bottom (Mark 1)........................ ................. 17 4.2 G r a nd Gul f ( Ma r k 11 1 ).......................................... 19 5.

SUMMARY

AND CONCLUSIONS............................................ 23 6. REFERENCES.......................................................... 25

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m 1 1. INTRODUCTION The radiological source terms resulting 'from postulated severe reactor accidents have important implications regarding health and publjc,, risk. To assess the radiological consequences of reactor accidents, an evaluation must be made of the quantities and characteristics of releases of radionuclides from the fuel pins to the environment. The fission product release and transport is strongly influenced among othr things by reactor type, containment design and the engineered safety features. In a boiling water reactor (BWR), the pressure suppression pool is de-signed to serve as a passive heat sink. In most accident sequences involving severe core damage, soluble gases and aerosol-laden gases vent through the suppression pool prior to escape to the outer containment building. The pas-sage of these materials (gases, vapors, and particulate materials) through the water in the pool results in the removal of certain fission products. This report presents information for the mitigative potential of pressure suppression pools in order to develop a technical basis for changes in regula-tory requirements for such engineered safety features.

4-q l I i l l I 1 l l l 1 ( SHIELD PLUGS I TOP OF SHIELD WALL I Y EEi9'u Q-Y M k..t'~~-[' ":\\ REACTOR 9 VESSEL x

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~ DRYWELL i DRYWELL SUPPORT I ~* . SKIRT .J y' DRYWELL iI{ VENT VENT; I j/ C.. J (. M.t. m.9 Q ~' y w .w-a a g_{.t ..a PRESSURE PRESSURE SUPPRESSION CHAMBER SUPPRESSION POOL (TORUS) Figure 1 BWR Mark I containment system. F e

1 2. DESCRIPTION OF PRESSURE SUPPRESSION P0OLS j The pressure suppression pool is primarily designed to reduce the primary containment pressure following a design basis accident. The thr.ee basic BWR containment designs (Mark 1, !!, and !!!) are illustrated in Figures 1 through 3. These three types of designs are similar in concept. The Mark I design has a separate toroidal pool (wetwell) that is con-nected to the main part of containment (drywell) by large vent pipes. Typi-3 cally, the suppression pools contain approximately 120,000 ft of water. The torus containing the water has a major diameter of about 110 feet and a minor diameter of 30 ft. Ducts several feet in diameter connect the drywell to the wetwell torus. The large ducts branch through a vent header into multiple (typically 2-ft-diameter) downcomers that have their open lower ends submerged in the water. Steam can also be directed into the pool by separate lines from the safety / relief valves on the reactor's primary system. The Mark 11 design is called the "over-under" design because the drywell is located directly' above the wetwell. Steam released during an accident to the drywell is conveyed into the suppression pool by multiple vertical steel downcomer pipes. The riowncomers penetrate the diaphragm floor separating the drywell and the wetwell. Vent valves in the floor allow free flow from the top of the wetwell back intoTe~drfwell.~ In the Mark 111 design the wetwell is an annular region at the periphery of the containment. Water is retained by the weir wall (height approximately 20 ft) and steam discharges into the pool from the drywell through submerged, horizontal vents in the lower drywell wall in the event of a steam system rup-ture in the drywell. The safety relief valves on the primary reactor system discharge directly into large pipe headers that terminate at spargers sub-merged in the sppression pool. The suppression pool volume is typically about 160,000 ft, similar to a Mark 11 plant. 9

STEEL CONTAINMENT ~ \\ ' SHIELD BUILDING DRYWELL HEAD \\ g PER CONTAINMENT n r f C .i Q) c l'I -REACTOR ,l _~. II l c - BIOLOGICAL DRYWELL I L =: --- DRYWELL WALL i:= ~ .a =1 s t = WEIR WALL hh[]l l l HORIZONTAL VE SUPPRESSION POOL [_i __ j -~~~" 3 .T T -'~'~ -E* 1 I Figure 3 BWR Mark 111 containment system. i

r O -. . m. k ('. l

• p'z 3

REACTOR VESSEL DRYWELL .g w / L k ' i[", ~'..,, p A{ VENT PRESSURE f.;, SUPPRESSION CHAMBER s. ~- ' -:-:-:+: e-l' = ~:1:::22:7 3: c -:J:,a - - ~ ~ -~ g % ((3 SUPPRESSION Q i POOL .;* j \\ e Figure 2 BWR Mark 11 containment system.

. In addition to the three removal mechanisms modeled by Fuchs, the SPARC code considers the following additional mechanism. Steam condensations (no particle size dependence assumed)., Convection caused by vapor flux to or from the bubble walls. T.he non-vection velocity is added algebraically to the deposition velocities calculated for other deposition mechanisms. Inlet impaction during gas injection into the pool. Particle growth in the bubble from water acquisition by deliquescing material in the particles. This is not specifically a removal mechan-ism, but it will enhance removal of larger particles by larger parti-cle dominant mechanisms and degrade removal of smaller particles by small particle dominant mechanisms. 3.2 Experimental Validation of the SPARC Code The phenomenological models included in the SPARC code are well supported by separate effects testing as found in the extensive literature on bubble dynamics and mass transfer between rising bubbles and liquid media. Experimental studies of pool scrubbing have been conducted at 'Battelle Columbus Laboratories. The available data base consists of particle scrubbing measurements taken in a pool using a 0.5 in diameter horizontal injector. The following conditions were varied during 56 dif ferent experiments: inert gas composition (air or helium), steam composition, gas flow rate, injector depth, pool temperature (ambient or near boiling), and aerosol (Csl, Te0, or 2 Sn), size, solubility, density, and aerosol concentration. Decontamination factor (DF) measurements for each experiment consist of the time-integrated particle mass flow rate into the pool divided by time-integrated particle mass flow rate out of the pool. Figure 4 presents a com-parison of experimental values and calculations by the SPARC code as used in the STCP. These comparisons correspond to an underprediction by SPARC by an -average factor of 6.2.8 3.3 SPARC Sensitivity Analyses The sensitivity study of the SPARC code involves variations in the fol-lowing important input parameters: Particle size of aerosols borne through the pool by gases The size of gas bubbles passing through the pool (DIAM) The aspect ratio of the gas bubbles (RATIO) The swarm rise velocity of the gas bubbles (VSWARM) The volume fraction of steam in inlet gas

3. PREDICTIVE METHODS Several models have been developed for predicting aerosol scrubbing effi-ciencies in BWR suppression pools. The Fuchs' model of particle removal from 2 L singleespherical bubbles is the basis of all particle scrubbing models and codes currpntly in use for nuclear reactor analysis. These models. include; (1)SPARC, developed under NRC sponsorship by Battelle Pacific Northwest Lab-oratory, (2) SUPRA, developed developed by-General Electric.,under EPRI sponsorship by SAIC and (3) a mode Several models have also been developed for scrubbing efficiency of solu-ble gases. Dif fey. et al.6 proposed a model for scrubbing efficiency of ele ' mental iodine based on the assumption that iodine in the gases leaving the pool is in equilibrium with iodine in the water pool. The experimental mea-surement reported by giffey et al. seems.to support the plausibility of their in. water at 100*C and model. Devell et al. carried out experiments with 12 concluded that iodine in gas bubbles did not necessarily reach equilibrium with iodine in the ' liquid and thus, extended the Diffey et al. model. to account for the degree of saturation. SUPRA also includes models for scrub- ~ bing gaseous-fissibn products. More recently, models for elemental and organic iodine scrubbing have also been added to the SPARC code.8 In the following sections, the SPARC (as used-in the STCP) aerosol scrub-bing model will be discussed and the code results will be used to. illustrate the variation in scrubbing decontamination factors over a range of input parameters _ selected to reflect the current uncertainty in their values. 3.1 SPARC (Suppression Pool Aerosol Removal Code) 2 The SPARC computer' code has been developed to calculate the behavior of aerosol ' particles in the pressure suppression pool under conditions that may be predicted to result from a postulated accident. The code calculates the. scrubbing of the aerosol particles from the gas mixture bubbling through the ' pressure suppression pool. This calculation is handled in terms of a decon-- tamination' f. actor- (DF) per particle size. . Fuchs' model of particle removal from single, spherical bubbles is the basis of particle scrubbing models used in the SPARC code. In the Fuchs' model, the cominant scrubbing processes take place inside rising bubbles. This model identifies three mechanisms of particle removal. They are: 1. Brownian diffusion of particles to the bubble wall (dominant for smaller particles). 2. Gravitational settling of particles to the lower bubble wall (dominant for larger particles). 3. Inertial deposition of particles on the bubble wall driven by the I centrifugal acceleration produced in the internal circulation of the gas in the bubble (dominant for larger particles). h

There are other less important input parameters to SPARC such as pool temperature, pool depth and percent of soluble material in particles. Aerosol particle size is a parameter obtained from the result of calcula-tions with the VANESA and TRAPMERGE model s of Source Term Lode Package (STCP). The sensitivity of the SPARC analyses to this parameter is reduced as the breadth of the particle size distribution is increased.10 The volume fraction of steam in inlet gas is a sequence-dependent quantity calculated by Lhe MARCH code. In this section decontamination curves calculated by SPARC will be presented to illustrate the importance of the user input parameters, namely, the bubble size, the bubble shape, and the bubble rise velocity. Tne parameter ranges for these variables that are chosen reflect a reasonable j range of uncertainty. studies of gas liquid hydrodynamics have been conducted at BCL.I{xperimentalTests have been conducted using mixtures of condensable (superheated steam) and noncondensable (air, helium, or hydrogen) gases injected into water pools through single hole and multihole configurations typical of those found in BWR quencher pipes. In an actual accident situation, swarms of bubbles ) rather than single ' bubbles will be encountered. The bubble size in these j swarms is a distribution. The bubble size distribution has been found to be { independent of the injection-flow rate and injection angle. There is, how-i ever, a dependence on condensable steam fraction. The distribution is well described by a lognormal distribution with mean diameters of 0.55 cm and 0.35 cm for low and high steam volume fraction respectively, and with a constant standard deviation of 1.E. The bubble diameters selected for the sensitivity study are 0.3 to 0.9 cm to reflect the range of uncertainty associated with this parameter. Aspect ratio and bubble diameter are related, the larger bubble being ( more elliptical. Tnis relationship also depends on water purity. Figure 5 shows the aspect ratio of bubbles as a function of their equivalent spherical k diameter. Two correlations from Clif t et al.12 are given. One is for pure water and the other for contaminated water. Impurity levels of parts per mil-I lion range are sufficient to produce more nearly spherical bubbles. Figure 5 also shows a correlation developed by BCL based on their experimental resul ts. The aspect ratio selected is based on the Clif t correlation for con-taminated systems. The resulting SPARC sensitivity to bubble diameter and aspect ratio is presented in Fig. 6. The important input parameters for these cases are pre-sented in Table 1. The input parameters to SPARC which are calculated by preceding codes in the STCP are taken from a recent BNL calculation for a typical time frame in the Peach Bottom TC2 sequence (TIME = 90 min). The bubble swarm rise velocity determines the residence time for I scrubbing. For a single bubble rising in an infinite pool, terminal velocit measurements for a large number of gas - liquid systems have been performed.1z For an air bubble rising in a stagnant water pool, the experimental data I reported by Haberman and Morton shows th6t the terminal rise veloc.ity is nearly constant at about 0.24 m/s for bubbles with equivalent diameters between 2 and 20 mm. For a swarm of bubbles the drag force between the bub-bles and the surrounding liquid will create significant circulation current in the liquid. Inside the bubble column the rising gas bubbles pump liquid from

.g. 10' o o o o 10 .b o

== O o m o o C o $10 2 __ y 9 mC L o 2 tf ~ , #I o to 10' gIf o te o t i 4/ 10 L-i....I 4,,,I ,,i.,,,1 2 10 10' 10 10 10' ' Calculated DF Figure 4 Comparison of SPARC calculated decontamination factors with BCL experimental values for 1/2 in diameter horizontal injector, i

eo. a. DIAM RATIO 0.s 1.s ... 0. 3....... 1 2 5.... i s..... i L. O \\ l l ce i. l l e .i o~O - s. 1 H t .c i U"- r' l l' CC u. \\ l o i 5: i 1 r H 5 i cc 1 a n i .z r i l ct n - z. e o l Q u l o. .o. l ~. s, l l s, l ~.,'~ .,...'~.......... ~, - .o 4 i i i e i6 4 i e e e a .....i., o i 10, i 10 10, 10 PARTICLE D!RMETER IMICRON) 1 Figure 6 Effect of bubble diameter and aspect ratio on DF. I l l 4

11 8 1 4 a i i i 4.0 ~ E 3li PURE 3.5 LIMIT a o a. O 3.0 p 10 ~ CONTAMINATED 5 f 2.5 100* C p \\ ~ 8 i 2.0 g o wg 1.5 .f p#'w ~ ~ BCL Correlation 1,o ? I i t i i i 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 1.1 De (cm) i figure 5 Variation in bubble aspect ratio with mean bubble diameter. ?

the bottom of the pool to the surf ace. The local. liquid velocity inside the bubble column increases the bubble rise velocity " relative to a stationary observer outside the pool. in the BCL experiments, the bubble rise veloci-ties have been measured relative to a stationary observer outside the tank for different gas injection rates. The typical spatial distributtom of bubble rise velocities 'in the bubble column varies between 20 cm/sec at the outer - edge of the bubble column to 100 cm/sec at the centerline of th6 bubble col umn. The values ranging from 20 to 116 cm have been chosen for the purpose of the present analyses. The higher value of 116 cm/sec was chosen becaus it ' corresponds to the value used in the BNL and the BCL STCP calculations.16 q8 The result of sensitivity of SPARC with respect to bubble. swarin rise velocity is presented in Fig. 7. The important input parameters for these cases are also presented in Table 1 and are based on BNL STCP calculations of Peach Bottom TC? Sequence (TIME = 116 min). The various uncertainties identified in the sensitivity study of the SPARC model are estimated to lead to an order of magnitude uncertainty in the pool decontamination. This conclusion is consistent with the QUEST study for the Grand Gulf TC 'sequ2nce performed by Sandia National Laboratory.10 It shoul d be noted that the SPARC underpredicts the DF values due to both unmodeled phenomena such as fragmentation and coalescence of bubbles as well as uncertainties associated with the code input parameters. As indicated earlier, the DF values in the BNL and BCL recent STCP calculations are similar to the lower bound estimates of the present sensitivity study. 3.4 Soluble Gas Scrubbing Mechanistic models for elemental and organic iodine scrubbing have been added recently to the SPARC computer code. A good comparison between avail-able experimental data and the SPARC prediction has been observed.8 An inte-gral decontamination factor of the order of 7000 for glgental iodine (1 ) has 2 been ' calculated for the Peach Bottom TC1 sequence.

• Due to high sol u-bility of H1 in water (relative to I ), a higher integral decontamination 2

factor -for hydrogen iodide is expected. An exact quantification of pool scrubbing efficiency for various soluble gases requires detailed calculations using any of the available models discussed previously. 4 i e .P ~

13-Table 1 The input Parameter Values to SPARC (Calculated by Preceding Codes in the STCP) TIME = 90 min TIME = 116 min Pool Temperature (*C) 113. 115 Pressure Above Pool ( ATM) 4.75 5.58 Inlet Gas Flow Rate (G/SEC) H0 1.27E+4 4.03E+4 2 H 5.12E+2 1.06E+2 2 CO2 0. O. C0 O.

0..

Ai r 0. O. Inlet Gas Temperature ('C) 517 451 Inlet Gas Pressure (EMS N6h 49.5 61. Particle Material Density (G/CC) 3. 3. Percent Soluble Material 0.35 0.37 i i l \\

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50. CM/SEC 116. CM/5EC

' 3E0.M/.6CP.".".. 0 l b: 's O i. 8 s8~ \\. ./ L. t 8 i ~.o. +-x l .z x r n - z OouO e - ? is. e. i ~o i iiii,3 i i iii i i i i i.,, i i 10,, 10 10, 10 PRRTICLE O!AMCTER (NICRON) I t' Figure 7 Ef fect of bubble swarm rise selocity on DF_.

t Table 2-Effective (Time Averaged) In-Vessel Release Decontapiqation factors for the Suppression Pool (Peach - Bottom Mark 1) l y ~ Fission Product Group TC2 Sequence TB1 Sequence CSI 200 Large* CSOH 300 Large 8 250 Large

• Due to inconsistencies in the reported values, an exact quantification is not possible.

l f 0 4 ^

17 4 EFFECTIVE DECONTAMINATION FACTORS FOR THE SUPPRESSION P00L One of the major considerations in predicting the scrubbing effectiveness of suppression pools is the definition of the environment and conditions that could conceivably challenge' the pool. This section of the report presents a sequence-based assessment of the effective decontamination factors for the BWR suppression pools. The information is based on the results of Sour'ce Term Code Package (STCP) calculations performed by BNL and BCL 17,1s for Peach 16 Bottom (Mark 1) and Grand Gulf (Mark 111) plant subject to a postulated severe accident condition. 4.1 Peac,h Bottom (Mark I) Peach Bottom Unit 2 Power Plant was included in the BCL and BNL STCP radionuclides release calculations. Peach Bottom, which is a General Electric BWR 4/ Mark I design, has been in operation since early 1970. The accident sequences selected for BCL detailed source term analysis includes of (1) TC, an anticipated transient without scram, (2) TB, a station blackout scenario, and (3) V, an interfacing system LOCA sequence. These sequences were selected on the basis of preliminary ASEP results on accident sequence probabilities as well as preliminary SARRP containment event tree quantification. In this sec-tion, the results of Source Term Code Package Calculations r one variation of TC (TC2) sequence and one variation of TB (TB1) sequence ~ used to assess the effective decontamination factors for the BWR/ Mark I suppression. pools. In the TC2 sequence the failure to scram is accompanied by the failure to achieve early power reduction as well as the failure to achieve emergency I depressurization. The primary coolant inventory is maintained by the combina-tion of the HPCI, RCIC, and the CRD systems. As the suppression pool heats up due to the continuing large steam input through the safety / relief valves, failure of the safety systems could take place due to loss of lubrication oil cooling, seal overheating, etc. In the present analysis the HPCI was assumed, to fail at a suppression pool temperature of 200'F, and the RCIC was assumed to fail at a containment pressure of 25 psia, due to high turbine exhaust back pressure.- The CRD system, which takes.its suction from the condensate storage tank, would continue to operate as long as the water in the latter was j available. The CR0 flow is insufficient to keep the core covered and cooled, and eventual core melting would take place. The containment would oe intact during the initial core melting in this sequence, but would fail shortly after the reactor vessel f ailure. For the TB1 scenario loss of all.off-site and on-site AC power leads to the loss of all active engineered safety features except the steam powered emergency core cooling systems. The latter, however, require DC power for operation and would fail when the station batteries are depleted; the latter has been estimated at six hours after the start of the accident. In such an event, core uncovery and melting takes place with the containment initially intact; containment fes:best is assumed to fail late in the accident sequence. 1 In both sequences considered, the in-vessel fission product elease due ] to core degradation and melting which consists primarily of Csi, Cs0H, and Te are free to pass down the safety relief lines and into the suppression pool through the quenchers, and these are subject to pool scrubbing. Table 2 pre-sents the implied decontamination factor (DF) for the in-vessel phase, for i

Containment. failure in this case ~ would. be expected due to the buildup of non- ~ ,condensables during the attack of the concrete foundation by the core debris. For the-early containment-failure. variation of the station' blackout scenario,

containment f ailure was assumed ' to occur, immediately~' af ter reactor ; vessel

' failure due to a large hydrogen-burn. The expulsion of the hot csre debris from the_ primary system is the. obvious ignition source. In the analysis-of-this scenario, a.1arge leakage between the drywell and the. outer containment was assumed after ~ vessel and containment f ailure; this implies some degrada- ~ tion of - the. boundary between the drywell - and containment due. to.the events associated-with primary system. failure or the hydrogen burn. In all three' sequences considered, the in-vessel fission product release due to core degradation and melting, primarily of Cs!,-Cs0H, and Te, are sub-e. ject to pool scrubbing. Table 3 presents the implied decontamination factor- , (DF). for the in-vessel release phase for the three calculated sequences. j In the TC and TB2 sequence, the fission product released during core / con-crete interactions can' bypass the suppression pool. However, in TB1 sequence it' is assumed that most of. the Te and refractory fission products during the j ex-vessel release ' phase' pass through the suppression pool. These fission pro-i ducts consist primarily of Ba, Sr, La, and Ce-with lesser quantities of Te (a proportionately larger fraction of Puff release at the time of pressure vessel failure is Te due' to its later release time, during the melt release phase). Table 4 presents the effective ex-vessel release decontamination factors for. l TB1 sequence. The DFs corresponding to the ex-vessel release phase are smaller than the DFs for the in-vessel release phase because:

1),The gases evolved ex-vessel contains less condensable gas (steam)

2) The pool temperature is higher later in the accident sequence

3) The ex-vessel aerosol particle sizes are smaller

4) The depth of the suppression pool during the release under water is smaller for the ex-ressel release.

The values for in-vessel release decontamination factors for the TC sequence shown in Table 3 are of ghe same order of magnitude as the lower L. for the Grand Gulf TC sequence performed bound estimates in the QUEST study by Sandia National Laboratories. (In-vessel release DF values for Cesium, Iodine and Tellurium reported in the QUEST study is 111). The lower bound ex-vessel release DF value for Tellurium reported in the QUEST study is 5 com- . pared to 10 obtained in the present study. i 9 .h e

ig. 1 g- 'both BNL TC2 and BCL TB1 calculated sequences. The variation of DF with fis-sion product species is due to the fact that the various species are released at;dif ferent times and thus experience different' conditions in the pool. in the TC2 sequence,.the containment was assumed to fail at tte time of pressure. vessel f ailure. This ensures that the; fission products releas,ed dur-ing core / concrete interactions can. bypass the suppression pool. However, in' the TB1 sequence,.it is assumed that the containment failure occurs late and therefore most of _ the Te and refractory fission products released ex-vessel are passed' through the suppression pool. These fission products consist pri-marily.of Ba, Sr, La,. and Ce with lesser quantities of Te (a' proportionately. i Te)ger fraction of the puff release at the-time of pressure vessel failure is lar In this case, the DFs corresponding to the ex-vessel release phase were found to'be negligible as compared with DFs for the in-vessel release phase. j The BCL STCP results for another variation of TC, TC3, which is identical 1 to TC2 except for inclusion of containment venting, was also studied. With ] venting all; the releases pass through the suppression pool but due to incon - j sistencies in, the. reported values, no quantification of ex-vessel release decontamination factors was possible at this time. j i 4.2 Grand Gulf (Mark III) i Selected severe accident scenarios for the Grand Gulf Unit 1 Power Plant were included in the BCL STCP radionuclides release calculations. Grand Gulf I Unit 1,.which is a General Electric BWR 6 with Mark III containment, began operations in June 1982. The accident sequences selected for BCL detailed source term analysis consists of (1) TC, an anticipated transient with scram, q and (2) TB, a station blackout scenario. These sequences were selected on the basis of preliminary ASEP results on accident sequence probabilities as well j as preliminary SARRP containment event tree quantification. q 'For'the TC sequence, the containment was assumed to fail by overpressure-zation prior to core melting due to the elevated power input to the suppres- 'sion pool associated with the failure to scram; containment failure was assumed to lead to failure of the emergency core cooling system pumps. It was also assumed that the Automatic Depressurization System -(ADS) would be acti-i 1 vated ' prior. to containment failure and subsequent core uncovery. In the analysis of the containment response, nominal leakage between the drywell and the outer containment bypassing the suppression pool was assumed, l Two variations of the station blackout (TB) scenario were considered. In l the first, late containment failure was considered and in the second, the con-tainment was assumed to fail at the time of reactor vessel f ailure. With the complete loss of electric power in this sequence, all the active engineered safety systems, witn the exception of the steam turbine driven emergency core cooling systems, would be unavailable. The turbine driven pumps would operate as long as the station batteries were available. The latter were assumed' to be depleted at six hours af ter the start of the accident. Also, in the ' absence of electric power, the ADS, upper pool dump, and the hydrogen igniters would not be able to perform their functions. Thus, core overheating and melting would take place with the primary system at elevated pressure. For the late containment failure variation of the station blackout sequence, nomi-nal leakage between the drywell and the outer containment was assumed. i \\

S e o 22 e BLANK PAGE e l l e 4 e e m p,d 1

t :e -21 I Table 3 Effective (Time Averaged) In-Vessel Release Decontamination Factors for the Suppression Pool (Grar.o Gul f Mark 111) i Fission Product Group TC Sequence TB1 Sequence TB2 Sequence l Csl 85 50 60 Cs0H 80 55 65 Te 40 40 50 i Table 4 Ef fective (Time Averaged) Ex-Vessel Release Decontamination Far. tors for the Suppression Pool (Grand Gulf Mark 111) Fission Product Group D-Vessel 0F Sr 25 Ba 20 La 15 Ce 30 Te 10 i l f l L___________

y- 's' P -24 BLANK PAGE ) e 4 6 l l 4 I 1 I 4 l l l l f I l

4 23 5.

SUMMARY

AND CONCLUSIONS Tne scrubbing models and sensitivity to input parameters of 'SPARC com-puter code used in Source Term Code Package (STCP) have been discussed. The various uncertainties identified in the sensitivity study of SPARC ~model were estimated to lead to an order of magnitude uncertainty in the decontamination factor by the suppression pool. An assessment of the effective decontamination factors for the suppres-sion pools based on the results of Source Term Code Package (STCP) calcula-tions performed by BNL and BCL has also been presented. The DF values in these calculations correspond to the lower bound estimates of the sensitivity study. It is seen that variation in pool decontamination factors are a func-tions of sequence and system being considered and the DFs corresponding to the ex-vessel release phase are smaller than the DFs for the in-vessel release phase. l e 9 9

26- ~ 16. M. Khatib-Rahbar et al., " Independent Verification of Radionuclides Release Calculations for Selected Accident Scenarios," NUREG/CR-4629, BNL/NUREG-51998, July 1986. 17. R.. S. Denning et al., " Radionuclides Release Calculations

• fo~r Selected Severe Accident Scenarios, Vol.1 BWR, Mark 1 Des 1 n," Battelle Columbus 9

Laboratories, Draft, November 1985. 18. R. S. Denning et al., " Radionuclides Release Calculations for Selected j Severe Accident Scenarios, Vol. IV, BWR, Mark Ill Design," Battelle Columbus Laboratories, Draf t, February 1986. i e l 1 1 1 1 l i I l

, ) l 6. REFERENCES 1. " Technical Basis for Estimating Product Behavior During LWR Accidents," NUREG-0772, U. S. Nuclear Regul atory Commission, Office of Nuclear Reactor Regulation,1981. 2. N. A. Fuchs, "The Mechanics. of Aerosol s," Pergammon Press, New York, 1964, p. 240. 3. P. C. Owczarski, R.1. Schreck, and A. K. Postma, " Technical Bases and User's Manual for the Prototype of a Suppression Pool Aerosol Removal Code (SPARC), NUREG/CR-3317, PNL-4742, 1985. 4. A. T. Wassel et al., " Analysis of Radionuclides Retention in Water Pools," J. of Nuclear Engineering and Design, Vol. 90, pp. 87-104,1985. 5. F. J. Moody, " Derivation of an Elliptical Suppression Pool Scrubbing Model," General Electric Company,1983. 6. H. R. Dif fey et al., " lodine Clean-Up in a Steam Suppression System," AERE-R-4882, UKAEA, Harwell, United Kingdom,1965. 7. L. Devell et al., " Trapping of lodine in Water Pools at 100"C," Pro-ceedings of IAEA Symposium on Containment and Siting of Nuclear Power Plants, CONF-67042,1967. 8. P. C. Owczarski, Private Communication, August 6, 1986. 9. B. C. Owczarski and W. K. Winegardner " Validation of SPARC, A Suppres-sion Pool Aerosol Capture Model," Proceedings of an International Sym-posium on Source Term Evaluation for Accident Conditions, Columbus, Ohio, October 28 - November 1,1985. 10., P. K. Mast et al. " Uncertainty in Radionuclides Release Under Specific LWR Accident Conditions," Volume IV TC Anslysis, SAND 84-0410, December 1985. 11. D. D. Paul, et al., " Radionuclides Scrubbing in Water Pools Volume 1: Gas-Liquid Hydrodynamics," EPRI NP-4154, August 1985. 12. R. Clift, J. R. Grace, and M. E. Weber, Bubbles, Drops, and Particles, Academic Press, New York, 1978, 13. G. B. Wallis, "Tne Terminal Speed of Single Drops or Bubbles on an Infi-nite Medium," Int. J. Multiphase Flow, Vol.1,1974, pp. 491-511. 14 W. L. Haberman and R. K. Morton, "The David Taylor Model Basin," Repcrt No. 802, 55715-102,1953. 15. P. C. Osczarski and W. K. Winegardner, " Capture of lodine in Suppression Pools," To be presented at the 19th DOE /NRC Nuclear Air Cleaning Confer-ence, Seattle, Washington, August 19, 1986. I

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f FEB 121987 RELEASE 0 yo 7pg p9R MEM0PANDUM FOR: Victor Stello, Jr. Executive Director for Operations FROM: James H. Sniezek, Chairman Committee to Review Generic Requirements

SUBJECT:

MINUTES OF CRGP MEETING NUMBER 109 The Conrittee to Review Generic Requirements (CRGR) met on Monday, February 9, 1987 from 1-3 p.m. A list of attendees for this meeting is enclosed (Enclosure 1). B~ Sheron (NRR) and L. Soffer (DSR0) presented for CRGP review the proposed new Standard Review Plan 6.5.5, " Pressure Suppression Pools as Fission Product' Cleanup Systems." Enclosure 2 summarizes the meeting. The'vugraphs used for the presentation are attached to enclosure 2. contains.predecisional information and, therefore, will not be released to the Public Document Room until the NRC has considelred (in a public forum) or decided the matter addressed by the information. In accordance with the ED0's July 18, 1983, directive concerning " Feedback and Closure on CRGR Reviews," a written response is required from the cognizant office.to report agreement or disagreement with CRGR recommendations in these minutes. The response, which is required within 5 working days after receipt i of these meeting minutes, is to be forwarded to the CRGR Chairman and if there is disagreement with the CRGR recommendations, to the EDO for decisionmaking. Questions concerning these meeting minutes should be referred to Walt Schwink (492 8639). Original signed by James H. 24r9 l W ") 9 4 James H. Sniezek, Chairman g Committee to Review Generic ~" 8 Requirements

Enclosures:

As stated Distribution: f JSniezek JRoe cc: Commission (5) JZerbe PRabideau SECY JClifford GZwetzig Office Directors FHebdon WLittle i Regional Administrators PErickson MLesar i CRGR Members-EFox DEDR0GR cf J W. Parler PDR (t'PG/CDGP) Central File B. Sheron W01mstead R0GR Staf# 'I( L. Soffer BZalcman VMcDonald OFC 1ROGR

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2/ } /87 0FFICIAL PECORD COPY

LIST OF ATTENDEES CRGR MEETING NO. 109~ February 9, 1987 CRGR MEMBERS J.H. Sniezek D. Ross R.E. Cunningham R.M. Bernero R.W. Starostecki S. Rubin (for C. Heltemes) J. Scinto OTHERS W. Schwink J. Zerbe J. Conran P.- Polk B. Sheron Z. Rosztoczy J. Mitchell L. Soffer J. Read M. Miller T. Cox W. Shields J. Clifford

2 to the Minutes of CRGR Meeting No. 109 Review of the Proposed New SRP 6.5.5 B. Sheron (NRR) and L. Soffer (DSRO) presented for CRGR review the proposed new Standard Review Plan (SRP) 6.5.5. The vugraphs used for this presentation, as well as a revised SRP 6.5.5., are attached. In essence, SRP 6.5.5. allows credit in the 10 CFR 100 dose calculation for suppression pools as fission product cleanup systems. Such credit is a recognition of the present state of knowledge regarding fission product retention in water. Allowing credit for suppression pool retention is not intended to reduce plant safety. However, the proposed SRP did envision SGTS filtration system efficiencies as low as 90 percent or containment allowable leakages as high as 5 percent. Other than plants applying for a new Construction Permit (CP), of which there presently are none, compliance with the SRP is voluntary. Furthermore, if a licensee uses conservative decontamination factors (DFs), then an analysis is not required. Conservative DFs for Mark I designs are equal to or less than DFs for Mark II/III designs due to Mark I smaller pool inventory and smaller downcomer submergence. ACRS did not object to issuance of SRP 6.5.5. for comment. However, ACRS advised that they wanted to re-look at the revised SRP after comments have been received and evaluated. The CRGR recommended that the following issues be addressed: (1) If increased fission product concentrations in the suppression pool are acknowledged, then the effect on the environmental qualification of eauip-ment and access to equipment during an accident should be addressed. (?) SRP 6.5.5. and R.G. 1.3 are inconsistent. They should both be revised in final form at the same time and, by so doing, be made consistent. This correlation should be indicated when the SPP is issued for comment. (3) ALAPA considerations vs. system safety requirements should be balanced. As a minimum, there should be a recognition of ALARA concerns. (4) It should be made clear that the burden of proof should be on the appli-cant if DFs above the conservative allowables are used. (5) Revisions to allowable containment leakage rates should be handled separately as part of the siting, source term or containment performance criteria efforts. CRGP recommends that the staff issue the proposed SRP for comment after appropriate revision to reflect issues 2 through 5. Issue 1 may be addressed subsequent to issuing the SRP for comment.

i. OVERVIEW PROPOSED NEW SRP 6.

5.5 BACKGROUND

o.0NE OF THE SHORT-TERM CHANGES DISCUSSED IN S MAJOR ASPECTS o PERMITS' CREDIT FOR SUPPRESSION POOLS AS CONSERVATIVE DECONTAMINATION FACTORS (

SYSTEMS, NO' APPLICANT ANALYSIS, SUPPRESSION P0OL BYPASS LEAKAGE TO BE AC O

CULATIONS, EXISTING 5SFFILTRATIONSYSTEMSNOTTOBEDEG o VALUE (90%) 0F REG, GUIDE 1,52 (REV 2). OTHER ASPECTS NOT DEPENDENT ON PARTICULAR SOURCE TERM IN o CAN BE USED WITH TID-14844 OR POTENTIAL REVISION NO LICENSEE ACTION REQUIRED. o

SUMMARY

- REPRESENTS RELAXATION WITH NO

SAFETY,

-8 .e. STANDARD REVIEW PLANS INV0KED BY DBA LO DOSE EVALUATION _ CONTAINMENT SPRAY AS A FISSION PRODUCT CLE 'o 6.5.2 FISSION PRODUCT CONTROL SYSTEMS AND S o 6.5.3 ICE CONDENSER AS A FISSION PRODUCT C o 6.5.4 ( PROPOSED PRESSURE SUPPRESSION P0OLS AS FIS o 6.5.5

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i SUPPRESSION POOL PARAMETERS TYPE VOLUME SUBMERGENCE OF 3 (FT ) DOWNCOMERS, i FT MARK 1 120,000 3-4 MARK Ii 160,000 10 - 15 MARK III 160,000 8.5 - 13 1

V.' l i SUPPRESSION P0OLS AS FISSION PRODUCT CLEANUP SYSTEMS - PRESENT STATUS REGULATORY GUIDE 1,3 - NO POOL CREDIT TO BE GIVEN, o SRP 6.5,3 STATES THAT POOL CREDIT MAY BE GIVEN, BUT GIVES NO o PROCEDURES OR CRITERIA FOR DOING S0, SRP 6.5.1, CRITERION V, PERMITS CHARC0AL FILTRATION UNITS o TO BE NON-ESF IF ( 90% 10 DINE EFFICIENCY, GESSAR-Il REVIEW ALLOWED POOL CREDIT FOR SEVERE ACCIDENT 0 RISK EVALUATION. I l

4 BASES FOR PROPOSED SRP SPARC CODE TIME-AVERAGED DECONTAMINATION FACTOR CALCULATIONS o FOR NUREG-1150, o PNL EXPERIMENTS ON 17 POOL SCRUBBING, o KNOWN CHEMISTRY OF HYDROGEN 10DIDE, A

DEFAULT DF VALUES IN PROPOSED SRP'6.5.5 DF = 1 FOR N0BLE GASES, ORGANIC 10DIDES (XE, KR, CH 1) 3 DF = 10'FOR_ PARTICULATE 10 DINE, OTHER AEROSOLS (MARK II AND III) 5 FOR PARTICULATE 10 DINE, OTHER AEROSOLS (MARK 1) = (Csl, TE, SP, ETC.) DF = 10 FOR ELEMENTAL IODINE (1 ) (MARK II AND III) 2 5 FOR ELEMENTAL 10 DINE (1 ).(MARK I) 2 = e _2______________________________ ]

i ~. Ef fective (Time Averaged) In Vessel Release Decontamination l Factors for the Suppression Pool (Peach Bottom Mark 1) l l In-Vessel DF TB1 Sequence Fissica Product Group TC2 Sequence .Large* 200 CSI Large 300 J C50H Large 250 Te act

• 0ue to inconsistencies in the reported values, an ex quantification is not possible.

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) i 4-POTENTIAL IMPACT OF SilPPRESSION POOL CREDIT ON OTHER ESF"S l SINCE SUPPRESSION P0OLS AND ESF FILTERS BOTH ATTENUATE 10 DINE, - l CREDIT FOR POOL SCRUBBING MIGHT BE TRADED OFF FOR FILTER ) n l REQUIREMENTS STAFF AND ACRS CONSENSUS NOT TO PERMIT CREDIT FOR SUPPRESS O POOLS TO ALLOW LARGE RELAXATION OF FILTERS (REMOVAL OR C STATUS TO NON-ESF), SINCE BOTH SYSTEMS YIELD DIVERSITY IN FISSION PRODUCT MITIGATION AND MINIMlZE BYPASS CONDITIONS, i THEREFORE, STAFF POSITION THAT EXISTING ESF"S CAN BE RELAXED o SOMEWHAT, BUT NOT BELOW MINIMUM VALUE OF R.G. 1,52, a

1 Proposed New Standard Review' Plan Section i 6.5.5 PRESSURE SUPPRESSION POOLS AS FISSION PRODUCT CLEAN-UP SYSTEMS REVIEW RESPONSIBILITIES Primary - Plant Systems Branch Secondary - Reactor Systems I AREAS OF REVIEW Pressure suppression pools are reviewed under this plan only when the applicant claims credit for fission product scrubbing and retention by the suppression pool. The pressure suppression pool and the drywell, when considered as a barrier to the release of fission products, are reviewed to assess the degree to which fission products released during postulated reactor accidents will be retained in the suppression pool. Leakage paths which allow fission products to bypass the pool are identified and reviewed, and the maximum fractional bypass leakage is obtained, for use in the evaluation of radiological dose consequences. 1.) Fission Product Control Requirement Sections of the SAR related to accident analysis, dose calculations, and fission product control are reviewed to I l __ 9

2 establish whether or not fission product scrubbing of the 1 drywell or reactor compartment atmosphere is claimed or required for mitigation of off-site consequences following a ) postulated accident. 2.) Design Bases A comparison is made to establish that the desion bases for the suppression pool and the drywell or reactor compartment are consistent with the assumptions made in the accident i evaluations of SAR Chapter 15. I 3.) System Design The information concerning the suppression pool is reviewed to ~ familiarize the reviewer with the expected temperature histories, depth of fission product entry expected during postulated accidents and potential leakage paths through drywell penetrations. 4.) Testing and Technical Specifications The details of the applicant's proposed preoperational tests, and, at the operating license stage, the surveillance requirements, are reviewed under section 6.2.1.1.C. The results of that review are examined to assure that pool depth and amount of leakage bypassing the pool are maintained con-sistent with the assumptions used in assessing the pool's effectiveness in fission product cleanup. i

7 3 ] { 'II ACCEPTANCE CRITERfA l l The acceptance criterie for the fission' product clean-up function of I the suppression pool are based on the-following requirements from j Appendix A of 10 CFR 50: l: A. General Design. Criterion 41 (Ref.1) as related to the control .i I of fission products following potential accidents. l I B. General Design Criterion 42 (Ref.1) as related to the periodic inspection of engineered safety features. 1 C: General Design Criterion 43 (Ref.1) as related to the periodic functional testing of engineered safety features. l Where thsy can be shown to be in compliance with these criteria, suppression pools may be given appropriate credit for fission product scrubbing and retention (except for noble gases, for which no pool retention is allowed) in the staff's evaluation of the radiological consequences of design basis accidents. Other assumptions concerning the release of radioactivity are to be taken from Regulatory Guide 1.3, except for Position C.I.f. which this section replaces. 1 Specific criteria which must be met to receive credit are as follows: ) 1. The drywell and its penetrations must be designed to assure that, even with a single active failure, all releases from the core must_ pass into the suppression pool, except for small bypass leakage. 1

?*. 2. The bypass leakage assumed for purposes of evaluating fission product retention must be no less than that accepted in the review under section 6.2.1.1.C, and must be demonstrated'in periodic tests by the. license technical specifications also reviewed under that section. 3. For plants which have already received a construction permit, the iodine retention calculated using this section must not be used to justify removal of the.etandby gas treatment or other filtered exhaust system from status as engineered safety features. For such reviews, criterion II.5 of SRP 6.5.1 shall not be applied, and the charcoal absorbers must be at least maintained to the minimum level of Table 2 in Regulatory Guide 1.52, Revision 2. Acceptable methods for computing fission product retention by the suppression pool are given in Subsection III, " Review Procedures." III Review Procedures l The first step in the review is to determine whether or not the suppression pool is to be used for accident dose mitigation purposes. If no fission product removal credit is claimed in the accident analyses appearing in chapter 15 of the SAR, no further review is required.

' 4: 5 - If.the suppression pool is intended as an engineered safety feature for the mitigetion of off-site doses, then the reviewer estimates its effectiveness in removing fission products from fluids expelled from the drywell or directly from the pressure. vessel through the depressurization system. l 1. Pool decontamination factor The decontamination factor (DF) of the pool is defined as the ratio of the amount of a contaminant entering the pool to the amount leaving. Decontamination factors for each fission product form as functions of time can be calculated by the SPARCcode(Ref.2),andthiscalculationshouldbeperformed whenever the pool design is judged by the reviewer to differ significantly from those found acceptable as fission product cleanup systems in past reviews. If, however, the time-inte-grated DF values claimed by the applicant for removal of parti-culates and elemental iodine are 10 or less for a Mark II or a Mark IIIBWR; and are 5 or less for a Mark I BWR, the applicants values tray be accepted without any need to perform calculations (Refs 3,4). (Ref. 3). A DF value of 1 (no retention) should be used for noble gases and, unless the applicant demonstrates otherwise, for organic iodides as well. If calculation of fission product decontamination is done usirp the SPARC code, the review should be coordinated with the Reactor Systems Branch, which is responsible for establishing the 4 accident assumptions needed to assemble the input for the calcu-lations.

-a 6 2. Pool bypass fraction The fraction of the drywell atmosphere bypassing the suppression pool by leaking through drywell penetrations is obtained as a product of the review under section 6.2.1.1.C. If B is the bypass fraction and DF is the time-integrated pool decontamination factor, then the overall decontamination, D, to be reported to the Reactor Systems Branch for use in chapter 15 I dose calculations may be taken as: OF D= H %(DF-Q or .= il + \\ S \\ DF D The reviewer should clearly distinguish that fraction of B which may be further treated by the standby gas treatment system from that fraction of B which also bypasses secondary containment. 3. Other containment atmosphere clean-up systems Plants having drywell or containment spray systems for which fission product cleanup credit is claimed are reviewed separately under section 6.5.2, and credit for both suppression pool and spray cleanup can be given as a result of the separate reviews.

l 4. Technical Specifications } 4 l The technical specifications are reviewed to assure that they I req &e periodic inspection to confirm suppression pool depth-and surveillance tests to confirm drywell leak tightness i consistent with the bypass fraction used in computing the overall decontamination. Technical specification review is coordinated with the Facility Operations Branch as provided in NRR.0ffice Letter No. 51. IV EVALUATION FINDINGS The reviewer verifies that sufficient information has been provided by the applicant and that the review and any calculations support conclusions of the following type, to be included in the staff's Safety Evaluation Report: ) We have reviewed the fission product scrubbing function of the pressure suppress' ion pool and find that the pool will reduce the fission product content of the steam-gas mixture flowing through the peol following accidents which blow down ) through the suppression pool. We estimate the pool will decontaminate the flow by a factor of for molecular iodine vapor and by a factor of for particulate fission products. No significant pool decontamination from noble gases or organic iodides will occur. The system is largely passive in nature, and the active components are suitably redundant such that its fission product attenuation

o ,s function can be accomplished assuming a single failure. The applicant's proposed program'for preoperational and surveillance tests will assure a continued state of readi-ness, and that bypass of the pool is unlikely to exceed the assumptions used in the dose assessments of Chapter 15. The staff concludes that the suppression pool is acceptable as a fission product cleanup system, and meets the requirements of General Design Criteria 41, 42 and 43. V IMPLEMENTATION Except in those cases in which.the applicant proposes an ~ acceptable alternative method for complying with the specified portions of the Commission's regulations, the methods described here in are to be used by the staff in its evaluation of conformance with Commissions regulations. Implementation of the acceptance criteria of subsection II of this plan is as follows: (a.) Operating plants and OL applicants need not comply with the provision of this review plan section. (b.) CP appi; cants will be required to comply with the provisions j j of this revision. I j: l-1 Q-_________-_

.~ 9 VI REFERENCES 1. 10.CFR Part 50, Appendix' A, General. Design Criteria 41,. " Containment Atmosphere Clean-up", 42, " Inspection of Containment Atmosphere Cleanup Systems",Jand 43, " Testing of Containment Atmosphere Cleanup System". 2. P.C. Owczarski, R.I. Shreck and A.K. Postma, " Technical Bases and Users Manual for the Prototype of a Suppression . Pool Aerosol Removal Code (SPARC)', NUREG/CR-3317, 1985. .3. P.C. Oweiarski and W.K. Winegardner, " Capture of Iodine .in Suppression Pools",19th DOE /NRC Nuclear Air Cleaning Conference, Seattle, 1986. 4. R.S. Denning et al, " Radionuclides Release Calcaations for Selected Severe Accident Scenarios",.NUREG/CR-4524, Vol. 1 i ) k ) L L i - = -__ -- -_____ _ _ _ _ _ _ - _

FEB 121987 P'EM0PANDUM FOR: Victor Stello, Jr. Executive Director for Operations FROM: James H. Sniezek, Chairman Committee to Review Generic Requirements i

SUBJECT:

MINUTES OF CPGP MEETING NUMBER 109 The Committee to Review Generic Requirements (CRGR) met on Monday, February 9, 1987 from 1-3 p.m. A list of attendees for this meeting is enclosed (Enclosure 1). B. Sheron (NRR) and L. Soffer (DSRO) presented for CRGP review the proposed new Standard Review Plan 6.5.5, " Pressure Suppression Pools as Fission Product Cleanup Systems." Enclosure ? summarizes the meeting. The vugraphs used for the presentation are attached to enclosure 2. contains predecisional information and, therefore, will not be released to the Public Document Room until the NRC has considered (in a public forum) or decided the matter addressed by the information. In accordance with the E00's July 18, 1983, directive concerning " Feedback and Closure on CRGP. Reviews," a written response is required from the cognizant office to report agreement or disagreement with CRGR recommendations in these minutes. The response, which is required within 5 working days after receipt of these meeting minutes, is to be forwarded to the CRGR Chairman and if there is disagreement with the CRGR recommendations, to the EDO for decisionmaking. Questions concerning these meeting minutes should be referred to Walt Schwink (492-8539). Originai signed by James H. D*9 m wm James H. Sniezek, Chairman 'f 10 LL -Jf Committee to Review Generic Requirements

Enclosures:

As stated Distribution: I JSniezek JPoe cc: Commission (5) JZerbe PRabideau SECY JClifford GZwetzig Office Directors FHebdon WLittle Regional Administrators PErickson MLesar CRGR Members EFox PEDR0GR cf W. Parler PDR RPG/CDGP) Central File B. Sheron W0lmstead P0GR Staf# /fh. L. Soffer BZalcman WcDonald 7 0FC !ROGR

ROGy

DED 0GR

...__:..________..:.._w 4 ____:.cc 4 NAME :PPolk

JZer e

Jh

>zek DATE :P/ lt /P7

2/// /87

2/ I /87 0FFICIAL PECORD COPY

I LIST OF ATTENDEES' CRGR MEETING NO. 109 February 9, 1987 l CRGR MEMBERS 'J.H.-Sniezek D. Ross R.E. Cunningham R.M. Bernero 'R.W. Starostecki - S. Rubin (for C. Heltemes) J. Scinto OTHERS W. Schwink ' J. Zerbe J. Conran P. Polk B. Sheron

2. Rosztoczy l

J. Mitchell j L. Soffer l J.. Read j M. Miller T. Cox W. Shields J. Clifford' l ) I i

1 . to the Minutes of CRGR Meeting No. 109 Review of the Proposed New SRP 6.5.5 B. Sheron (NRR) and L. Soffer (DSRO) presented for CRGR review the proposed new Standard Review Plan (SRP) 6.5.5. The vugraphs used for this presentation,'as well as a revised SRP 6.5.5., are attached. l In essence, SRP 6.5.5. allows credit in the 10 CFR 100 dose calculation for suppression pools as fission product cleanup systems. Such credit is a recognition of the present state of knowledge regarding fission product retention in water. i Allowing credit for suppression pool retention is not intended to reduce plant safety. However, the proposed SRP did envision SGTS filtration system efficiencies as low as 90 percent or containment allowable leakages as high as 5 percent. Other than plants applying for a new Construction Pennit (CP), of which there presently are none, compliance with the SRP is voluntcry. Furthermore, if a licensee uses conservative decontamination factors (DFs), then an analysis is I not required. Conservative DFs for Mark I designs are equal to or less than DFs for Mark II/III designs due to Mark I smaller pool inventory and smaller downcomer submergence. ACRS did not object to issuance of SRP 6.5.5. for comment. However, ACRS advised that they wanted to re-look at the revised SRP after comments have been received and evaluated. The CRGR recommended that the following issues be addressed: (1) If increased fission product concentrations in the suppression pool are acknowledged, then the effect on the environmental qualification of eouip-ment and access to equipment during an accident should be addressed. (2) SRP 6.5.5. and R.G. 1.3 are inconsistent. They should both be revised in final form at the same time and, by so doing, be made consistent. This correlation should be indicated when the SPP is issued for comment. (3) ALAPA considerations vs. system safety requirements should be balanced. As a minimum, there should be a recognition of ALARA concerns. (4) It should be made clear that the burden of proof should be on the appli-cant if DFs above the conservative allowables are used. (5) Revisions to allowable containment leakage rates should be handled separately as part of the siting, source term or containment performance criteria efforts, i CRGP recommends that the staff issue the proposed SRP for comment after i appropriate revision to reflect issues 2 through 5. Issue 1 may be addressed subsequent to issuing the SRP for comment.

l L, OVERVIEW PROPOSED NEW SRP 6.

5.5 BACKGROUND

ONE OF THE SHORT-TERM CHANGES DISCUSSED IN SE o MAJOR ASPECTS PERMITS CREDIT-FOR SUPPRESSION POOLS AS F CONSERVATIVE DECONTAMINATION FACTORS (D o

SYSTEMS, NO APPLICANT ANALYSIS, SUPPRESSION POOL BYPASS LEAKAGE TO BE AC o

CULATIONS, EXISTING ESF FILTRATION SYSTEMS NOT TO BE o VALUE.(90%) 0F REG, GUIDE 1,52 (REV, 2), OTHER ASPECTS NOT DEPENDENT ON PARTICULAR SOURCE TERM IN o CAN'BE USED WITH TID-14844 OR POTENTIAL REVISION, NO LICENSEE ACTION REQUIRED, o

SUMMARY

- REPRESENTS RELAXATION WITH N0 Sl i

SAFETY, I

i

., I'. ^* STANDARD REVIEW PLANS INV0KED BY DBA'LOC DOSE EVALUATION j CONTAINMENT SPRAY AS A FISSION PROD l L o 6.5.2 h FISSION PRODUCT CONTROL SYSTEMS AND ST o 6.5.3 ICE CONDENSER AS A FISSION PRODUCT CLEANUP SYS o '6.5.4 PROPOSED PRESSURE SUPPRESSION P0OLS AS FISSI o 6.5.5

SYSTEMS, RADIOLOGICAL CONSEQUENCES OF A DBA o

15.6.5A MENT LEAKAGE. LEA AGE FROM ENGINEERED SAFETY CO o 15.6.5B CONTAINMENT. LEAKAGE FROM MAIN STEAM LINE ISOLA o 15.5.5D CONTROL SYSTEM (BWR) . - _ _ - _ = _ - - - - - - _ _ _ i

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a SUPPRESSION POOL PARAMETERS TYPE VOLUME SUBMERGENCE OF 3 (FT ) DOWNCOMERS, FT l MARK I 120,000 3-4 i MARK II 160,000 10 - 15 MARK III 160,000 8,5 - 13 f

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l SUPPRESSION P0OLS AS FISSION PRODUCT CLEANUP SYSTEMS - PRESENT STATUS REGULATORY GUIDE 1,3 - NO POOL CREDIT TO BE GIVEN, o i SRP 6,5,3 STATES THAT POOL CREDIT MAY BE GIVEN, BUT GIVES NO o PROCEDURES OR CRITERIA FOR DOING S0, 4 SRP 6.5.1, CRITERION V, PERMITS CHARC0AL FILTRATION UNITS o TO BE NON-ESF IF ( 90% IODINE EFFICIENCY, GESSAR-Il REVIEW ALLOWED POOL CREDIT FOR SEVERE ACCIDENT O RISK' EVALUATION, i

i 1 BASES FOR PROPOSED SRP SPARC CODE TIME-AVERAGED DECONTAMINATION FACTOR CALCULATIONS o FOR NUREG-1150, o PNL EXPERIMENTS ON 12 POOL SCRUBBING. o KNOWN CHEMISTRY OF HYDROGEN 10DIDE. l l

Jy DEFAULT DF VALUES IN PROPOSED SRP 6.5.5 DF = 1 FOR NOBLE GASES, ORGANIC 10DIDES CH 1) (XE, KR, 3 DF = 10 FOR PARTICULATE IODINE, OTHER AEROSOLS (MARK II AND III) 5 FOR PARTICULATE-IODINE, OTHER AEROSOLS (MARK 1) = (Csl, TE. SP, ETC.). DF = 10 FOR ELEMENTAL-IODINE'(1 ) (MARK II AND.'III) 2 5 FOR ELEMENTAL IODINE (1 ) (MARK I) 2 =- e L 2

'l q m. i i Ef fective (Time Averaged) In Vessel Release Decontap nat on f actors for the Suppression Pool (Peach Bottom Mark 1) In-Vessel 0F-T81 Sequence Fission Product Group TC2 Sequence -1 .Large* 200 CSI Large 300 CSOH targe 250 Te ct

• 0ue to inconsistencies in the reported values, an exa quantification is not possible.

Effective (Time.' Averaged) Ex-Vessel Release Decontamination Factors for the Suppression Pool- (Grand Gulf Mark 111) Fission Product Group TC Sequence TB1 Sequence TB2 Sequence 85 50 60 Csl 65: 55 80 1 - 'CsDH. 40 50 40 Te Ex-Vessel 0F Fission Product Group 25 Sr 20 Ba 15 La 30 Ce 10 Te 1

0 0 8 J 0 b 7 S 'o R b O 6 TC A F 0 b ~ N S O )n I T i A m o N b( I 4 E M M \\ A \\ T I \\ T 1 N o O b 3 C ED / 1 0 b 2 ES f A 1 C u\\( 0 b f [ h! l I .l 0

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• , e mE S ls C.

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N ~ _ _ - - r- ~ _ 2- =5 = 1- =i 1 T 3 O u '0 '0 o '0 0 I T 1 e 1 1 i 1 AD O 30 E3FE~ NO II M AE ~ TP NOT CLS EE DM ~ I ll

a SUPPRESSION POOL BYPASS l i NOT ALL FISSION PP0 DUCTS G0 FROM DRYWELL THRU SUPPRES SMALL FRACTION (TYPICALLY, FEW PERCENT) BYPASSES P00L. I o l 1 FRACTION THAT BYPASSES POOL IS UNSCPUBBED, o NEED TO ACCOUNT FOR POOL BYPASS IN STAFF ASSESSMENT OF j o i CONSEQUENCES. I 1 L-____

, f; ~' POTENTIAL IMPACT OF SUPPRESSION POOL CREDIT-ON OTHER ESF"S SINCE SUPPRESSION POOLS AND ESF FILTERS BOTH ATTENUATE 10 DINE, o L CREDIT FOR POOL SCRUBBING MIGHT BE TRADED OFF FOR FILTER REQUIREMENTS STAFF AND'ACRS CONSENSUS NOT TO PERMIT CREDIT FOR SUPPRESSIO o POOLS T0' ALLOW LARGE RELAXATION OF FILTERS (REMOVAL.0R CHj STATUS TO NON-ESF), SINCE BOTH SYSTEMS YlELD DIVERSITY IN FISSION PRODUCT MITIGATION AND MINIMlZE BYPASS CONDITIONS. I o THEREFORE, STAFF POSITION THAT EXISTING ESF"S CAN BE RELAXED SOMEWHAT, BUT NOT BELOW MINIMUM VALUE OF R.G. 1.52., 1 ) _ =_____-_-________ __ _ _ _ - _ - _ - _ - _ _ _ _ _

4 c,. r 3 Proposed.New Standard Review Plan Section. 6.5.5 PRESSURE SUPPRESSION P0OLS AS FISSION PRODUCT CLEAN-UP SYSTEMS. REVIEW RESPONSIBILITIES Primary - Plant Systems Branch Secondary - Reactor Systems AREAS OF REVIEW Pressure suppression pools are reviewed under this plan only when the applicant claims credit for fission product scrubbing and retention.by the suppression pool. The pressure suppression pool and the drywell, when considered as a barrier to the release of fission products,.are reviewed to assess the degree to which fission products released during postulated reactor accidents will be retained in the suppression pool. -Leakage paths which allow fission products to bypass the pool are identified and reviewed, and the maximum fractional bypass leakage-is obtained. for use in the evaluation of l radiological dose consequences. l t 1.) Fission Product Control Reau1rement l Sections of the SAR related to accident analysis, dose calculations, and fission product control are reviewed to bo

I j 2 establish whether or not fission product scrubbing of the j .drywell or reactor compartment atmosphere is claimed or. l required for mitigation of off-site consequences following a-l postulated accident. 2.) Design Bases A comparison is made to. establish that the design bases for'the suppression pool and the drywell or reactor compartment are consistent with the assumptions made in the accident evaluations of SAR' Chapter 15. 3.) System Design The information.concerning the suppression pool is reviewed to familiarize the reviewer with the expected temperature l histories, depth of fission product entry expected during j postulated accidents and potential leakage paths through-drywell penetrations. 4.) Testing and Technical Specifications The details of the applicant's proposed preoperational tests, l and, at the operating license stage, the surveillance requirements, are reviewed under section 6.2.1.1.C. The results of that review are examined to assure that pool depth i and amount of leakage bypassing the pool are maintained con-sistent with the assumptions used in assessing the pool's effectiveness in fission product cleanup. i

3 l II ACCEPTANCE CRITERIA The acceptance criterie for the fission product clean-up function of L the suppression pool are based on the following requirements from Appendix A of 10 CFR 50: 3 A. General Design Criterion 41 (Ref.1) as related to the control of fission products following potential accidents. B. General Design Criterion 42 (Ref.1) as related to the periodic inspection of engineered safety features. C. General Design Criterion 43 (Ref.1) as related to the periodic functional testing of engineered safety features. Where they can be shown to be in compliance with these criteria, suppression pools may be given appropriate credit for fission product scrubbing and retention (except for noble gases, for which no pool retention is allowed) in the staff's evaluation of the radiological consequences of design basis accidents. Other assumptions concerning the release of radioactivity are to be taken from Regulatory Guide 1.3, except for Position C.I.f, which this section replaces. Specific criteria which must be met to receive credit are as follows: 1. The drywell and its penetrations must be designed to assure that, even with a single active failure, all releases from the core must pass into the suppression pool, except for small l bypass leakage.

I I p f

• W 4

I . g m 2. The bypass leakage assumed for purposes of evaluating fission product retention must be no less than that accepted in the review under section 6.2.1.1.C, and must be demonstrated in periodic tests by the license technical specifications also reviewed under that section. 3. For plants which have already received a construction permit, the iodine retention calculated using this section must not be used to justify removal of the standby gas treatment or other filtered exhaust system from status as engineered safety features. For such reviews, criterion 11.5 of SRP 6.5.1 shall not be applied, and the charcoal absorbers must be at least maintained to the minimum level of Table 2 in Regulatory Guide 1.52, Revision 2. Acceptable methods for computing fission product retention by the suppression pool are given in Subsectica III, " Review Procedures." III Review Procedures The' first step in the review is to determine whether or not the suppression pool is to be used for accident dose mitigation If no fission product removal credit is claimed in the purposes. accident analyses appearing in chapter 15 of the SAR, no further review is required.

F p a 5 1 If the suppression pool is intended as an engineered safety feature for the mitigetion of off-site doses, then the reviewer estimates its effectiveness in removing fission ~ products frem fluids expelled ~ from the drywell or directly from the pressure vessel through the depressurization system. 1. Pool decontamination factor The decontamination factor (DF) of the pool is defined as the ratio of the amount of a contaminant entering the pool to the amount leaving. Decontamination factors for each fission product form as functions of time can be calculated by the SPARC code (Ref.2), and this calculation should be performed whenever the pool design is judged by the reviewer to differ significantly from those found acceptable as fission product cleanup systems in past reviews. If, however, the time-inte-grated DF values claimed by the applicant for removal of parti-culates and elemental iodine are 10 or less for a Mark II or a MarkIIIBWR;andare5orlessforaMarkIBWR,theapplicants values may be accepted without'any need to perform calculations (P-" 3,4). (Ref. 3). A DF value of 1 (no retention) should be used for noble gases and, unless the applicant demonstrates otherwise, for organic fodides as well. If calculation of fission product decontamination is done using the SPARC code, the review should be coordinates with the Reactor Systems Branch, which is responsible for establishing the accident assumptions needed to assemble the irput for the calcu-lations. r

m j j w-2. Pool bypass fraction The fraction of the drywell atmosphere bypassing the 1 suppression pool by leaking through drywell penetrations is i obtained as a product of the review under section 6.2.1.1.C. If B is the bypass fraction and DF. is the time-integrated pool decontamination factor, then the overall decontamination, D, to be reported to the Reactor Systems Branch for use in chapter 15 dose calculations may be taken as: DF 0= l-A %(bF-Q =l+t1 or \\ M D The reviewer should clearly distinguish that fraction of B which may be further treated by the standby gas treatment system from that fraction of B which also bypasses secondary containment. 3. Other containment atmosphere clean-up systems Plants having drywell or containment spray systems for which fission product cleanup credit is claimed are reviewed separately under section 6.5.2, end credit for both suppression pool and spray cleanup can be given as a result of the separate reviews.

j 7 ] 4. Technical' Specifications The-technical specifications are reviewed to assure that they-require periodic inspection to' confirm suppression pool' depth and surveillance tests to confirm drywell leak tightness consistent with the bypass fraction used in computing the overall.. i decontamination. Technical specification review is coordinated 4 with the Facility Operations Branch as provided in NRR Office Letter No. 51. IV EVALUATION FINDINGS ' 1 The reviewer verifies that sufficient information has been provided by the applicant and that the review and any calculations support conclusions'of the following type, to be included in the staff's Safety Evaluation Report: We have reviewed the fission product scrubbing function of the pressure suppression pool and find that the pool will reduce the fission product content of the steam-gas mixture flowing through the pool following accidents which blow down through the suppression pool. We estimate the pool will decontaminate the flow by a factor of for molecular iodine vapor and by a factor of for particulate fission products. No significant pool decontamination from noble gases or organic iodides will occur. The system is largely passive in nature, and the active components are suitably redundant such that its fission product attenuation l l t

}.= 8 4 function can be accomplished assuming a single failure. The 1 applicant's proposed program for preoperational and surveillance tests will assure a continued state.of readi-ness, and that bypass of the pool is unlike_y to exceed the l assumptions used in the dose assessments of Chapter 15. The staff concludes that the suppression pool is acceptable as a fission product cleanup system, and meets the requirements of General Design Criteria 41, 42 and 43. V IMPLEMENTATION Except in those cases in which the applicant proposes an acceptable alternative method for complying with the specified portions of the Commission's regulations, the methods described here in are to be used by the staff in its evaluation of conformance with Commissions regulations. Implementation of the acceptance criteria of subsection II of this plan is as follows: (a.) Operating plants and OL applicants need not comply with the provision of this review plan section. (b.) CP applicants will be required to comply with the provisions f of this revision. ---_-__.________m__

( C 9 VI REFERENCES 3. 10 CFR Part 50, Appendix A, General Design Criteria 41, " Containment Atmosphere Clean-up", 42, " Inspection of Containment Atmosphere Cleanup Systems", and 43, " Testing j 1 i of Containment Atmosphere Cleanup System". l 2. P.C. Owczarski, R.I. Shreck and A.K. Postma, " Technical Bases and Users Manual for the Prototype of a Suppression Pool Aerosol Removal Code (SPARC)', NUREG/CR-3317,1985. 3. P.C. Owczarski and W.K. Winegardner, " Capture of Iodine in Suppression Pools",19th DOE /NRC Nuclear Air Cleaning Conference, Seattle, 1986. 4. R.S. Denning et al, " Radionuclides Release Calculations for Selected Severe Accident Scenarios", NUREG/CR-4524. Vol. 1 ..}}