ML20058K007

Denies 880425 Backfit Claim Per 10CFR50.109 Re NRC Interpretation of Sys Pressure Testing Requirements Under ASME Code,Section XI

ML20058K007
Person / Time
Issue date:	08/17/1988
From:	Shao L Office of Nuclear Reactor Regulation
To:	Reynolds N NUCLEAR UTILITY BACKFITTING & REFORM GROUP
Shared Package
ML20058A334	List: ML20058A331 ML20058A336 ML20058J967 ML20058K007 ML20058K097 ML20058K115 ML20058K125
References
NUDOCS 8809130282
Download: ML20058K007 (6)

Text

{{#Wiki_filter:* n asc /o uSe UNITED STATES l ' 3 s.,, NUCLEAR REGULATORY COMMIS$lON mamwovow. o. c.aosas 4,][ 8 August 17, 1988 Nicholas S. Reynolds i Counsel to Nuclear Utility Backfitting and Reform Group (NUBARG) Suite 800 1400 L Street, N.W. Washington, D.C..20005-3502

Dear Mr. Reynolds:

In your April 25, 1988, letter you submitted a backfit claim pur,suant to -10.CFR 50.109 regarding the NRC staf f's interpretation of the system pressure testing requirements under ASME Code, Section XI. the staff has changed its position to one of not permittinSpecifically, you claim that operation before or during hydrostatic or leakage testing.g core critical However, the. staff has never changed its position; the core must not be critical before or during these tests. Therefore, we must deny your claim. 1 following NRC documents and licensees' written connitments:Th (2) boiling-water reactor (BWR) technical specifications (3) NUREG-75/087,(1 the Standard Review Plan Reports by the licensees,(and (5) staff Safety Evaluation Reports 11censees. Sect, ion 50.5Sa of 10 CFR establishes that, throughout the service life of a BWR facility, reactor vessels are to be examined and inspected in accordance with Section XI of the ASME Code. Note 5 in Table IWB-2500-1 (B-P) Section XI of the ASME Cote (Reference 1), states that "the system leakage test shall be con-i ducted prior to elTmetartup following each reactor refueling outage

• 1.2 in the Technical 5phifications for the Hatch Unit 2 plant (Reference 2)

Table indicates that SWRs have five operational conditions, and the condition before <startup is hot shutdown. be-in the shutdown position.During the hot shutdown condition, the mode switch must indicates "when the mode switch is in shutdown, the reactor is to be with all control rods inserted." are typical of BU FSARs and Technical Specifications.The Hatch 2 FSAR and Tec these documents must perform the system leakage test with the control rodsThus, inserted and the reactor not critical. Section 5.3.2 in NUREG-75/087 (later NUREG-0800) establishes safety margins and cooldown, and core critical operations. required for nuclear plants durin As the safety conse operation increases, the staff requires greater safety margins. quence of the during core critical operations, thereby implying that'the cor SRP Section critical during the testing. Moreover, some BWR licensees' Technical Specifics-tions or FSARs indicate that the hydrotests are to be performed with these lower safety margins. licensees will perform hydrotests when the core is not critical.Thus it

a q Nicholas S. Reynolds Section 5,3 of some BWR FSARs indicates that the control of reactivity.will be in accordance with GE BWR Licensing. Topical Report.NED0-21778-A (Reference 4). The Nuclear Regulatory Comission's acceptance basis is includeLtherein. For example, in the Perry Safety Evaluation Report (Reference 5), the staff per-- mitted modifications to the requirements of 10 CFR 50 Appendix G, based on i NEDO-21778-A. The General Electric Company provided justification in NE00-21778-A kor modifying the safety margins identified in SRP Section 5.3.2 and 10 CFR $0, Appenoix G. Most of the roquested changes were later incorporated into 10 CFR-50, Appendix G. As part of the bases for this modification, General ~ Electric .) described the procedure BWRs were to follow during hydrostatic and leakage. i testing. These procedures preclude the core being critical before and during i hydrostatic and leakage testing. Section 3.1.2 of the NED0-21778-A report states: Prior to start after a refueling' outage, an operational hydrostatic. pressure leak check is conducted-to verify the integrity of RPV head seal and other vessel / appurtenance interfaces. Initially, all control rods ') are fully inserted and the reactor mode switch is in ' Refuel." Control rods cannot be withdrawn. The RPV. is vented, filled 10- the top using the - CR0 system, and pressurized to 1000 psig at a rate of 100 psig per minute along curve A of Figure 2-1. Pressure in the vessel-is maintained at. 1000 psig during the leak check by simultaneous filli19 with the CRD-l system and rejecting water through the RWCU system. Section 3.1.3 of the NE00-21778-A report states: The Section XI hydrostitic pressure test is perfomed in essentially.the same manner as the operational hydrostatic leak check except that the fomer is 'perfomed at a higher pressure along curve A of Figure 2-1. In its evaluation of this NEDO report the staff intJicated its position on whether hydrostatic or leakage testing may be perfomed with.the core critical. In the " Regulatory Position" section of the NEDO-21778-A report evaluatio% the staff stated: . All control rods shall be fully inserted during hydrotestt." "Hydrotests" in this statement means hydrostatic and-leakage tests. ] The staff's position was again stated in memoranda between A. F. Gibson and R. M. Bernero in April 1986. On April 17. 1986, Mr. Gibson referenced BWR Owners Group coments in NEDO-31140, December 1985, as.having raised the nuclear heat issue, and he asked for an interpretation of the testing requirements. Mr.'Bernero's reply on April 29, 1986, states: "The position of the staff is that System Pressure Tests (Leakage and Hydrostatic) are to be perfomed before the reactor goes critical from a refueling outage." In your. April 25, 1328, letter you state the following bases for your claim: (1) "For a number of years the Staff has allowed leakage and hydrostatic tests using nuclear heat, i.e., pulling control rods and using nuclear power to heat up and pressurize the primary coolant system..." 4

d I -) Nicholas S. Reynolds.! (2) [Thestaffrecognizedthat):"'thereisminimaldifferenceinthe safety afforced' by the two methods" (and the staff had) "' permitted' the licensee to-conduct hydrostatic tests using nuclear heat 'for a number of years' and recognized the.' late arrival of the staff posiYT5n'" (May 5, 1986, letter from D.' R. Muller, NRC, to J. T. Beckham, Georgia Power Co). (3) "Section 1Y. A.5 of Appendix G scates... 'If there is fuel Jn the ' reactor during system hydrostatic pressure tests or leak tests,'the requirements of paragraphs 2 or 3 of this section apply, depending on whether the core is critical during the test.'" (4)' "The ASME Comittee stated that[c3 ore criticality during pressure - testing is not addressed by Section-XI. Division 1,'" 'and that. "...the Code pemits the use as a pressurizing medium of 'a mixture of steam, water,. and non-concensible gases in a proportion no greater than that present-during normal startup... '" The NUBARG position that:the NRC staff had previously allowed leakage and hydrostatic tests, using nuclear heat, appears to be based on the stated fact that Georgia Power had been using this practice since initial plant startups, presumably with no staff objection. practice until April 1986. _ Not ob However, the staff was not aware of this stitute permission or concurrence.jecting to en unknown practice.do2s not' con-Our refusal to pemit the practice when we-became aware of.it does not constitute a chsnge in position or interpretation and, therefore, does not constitute a-backfit. NUBARG has misinterpreted the May 5,1986, letter.- The subject of the letter was one time relief from the staff positiori. It was granted to avoid a delay of The justification given for granting temporary relief in this c interpreted as a generic staff positiolipermitting pressure tests with nuclear hea t. NUBARG has 41so misinterpreted paragraph IV.A.5 of 10-CFR_50 Appendix G.. Appendix G provides fracture toughness requirements, not-system requirements. The phrase.*oepending on whether the core is critical during the test" in Appendix G does acknowledge the use of nuclear heat, but'only when.special. circumstances arise, such as for Hatch 1 on a one-time basis. The technicalL issues that must be resolved before nucigar heat may be used during testing are addressed in an April'10,1987, J. P. O'Reilly, Georgia Power Company. letter from J. H. Sniezek, NRC, to i The remarks of the ASME Comittee do not support the claim that the staff's position on hydrotests is a backfit. It is true that Section XI Division 1 of the ASME Code does not explicitly address core' criticality during pressure testing. However, as discussed above, it' follows from the Code that licensees with FSARs and Technical Specifications which have restrictive conditions that 4 preclude testing with nuclear heat must perform the hydrotests without using nuclear heat.

Nicholas S. Reynolds 4, i For the reasons given above, your Apri1~ 25,1988 claim pursuant to'20 CFR 50.109-regarding'the NRC staff's interpretation of the system pressure testing i i requirements under ASME Code.- Section XI, is denied. - Sincerely, T N %./-4 Lawrence C. Shao, Director Division of Engineering and Systems. ~ Technology Office of Nuclear Reactor Regulation O e l l l f c 1 ~ -.. - -.

i 4 l References' \\ (1) American Society of Mechanical Engineers Boiler and Pressure Vessel Coot, Section XI, Division 1,1980 Edition, Winter 1981 Addenda. (2) Technical Specifications for Edwin I. Hatch Nuclear Plant, Unit 2. - ' NUREG-0395, June 13, 1978, Appendix A to License No. NPF-5. (3) Updated Final Safety Analysis Report: Edwin I. Hatch Nuclear Plant - Unit 2. July 28, 1975. (4 ') Licensing Topical Report NE00-21778-A, " Transient Pressure Rises Affecting Fracture Toughness Requirement-for Soiling Water Reactors,"- December 1978. (5) Safety Evaluation Report Related to the Operation of Perry Nuclear Power Plants, Units l'and 2. NUREG-0887, May 1982. ,. j / d e9 0 6 i G '3 O 1 l W w m--- . J ---.m-m--99mga, = w ev e-

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i), l; l NEno-21778-A 1 i 0 l l . /*% %g l-f \\ NUCLEAR RIOULATORY 00h8WS$10N i ummmm s i nasemoves.o.a.ames L i av u en u Central Files - Topical Reports , l 1 Dr. G. G. Sherwood f '.f. Manager Safety and Licensing . General Electric Company 175 Curtner Avenue 4 San Jose, California 95114 p'

SUBJECT:

REVIEW 0F GENERAL ELECTRIC TOPICAL REPORT NE00 21778. "TRAN5!ENT a PRES 5URE RISES AFFECTING FRACTURE TOUGNNES$ REQUIREENTS FOR B0ILING WATER REACTOR 5" We have completed our review of the subject topical report and have concluded that the report provides an acceptable basis for an amendment to Appendix G of 10 CFR 50. Our proposed revision to paragraph IV.A.2.c of Appendix G is currently being reviewed within the NRC prior to . publication in the Federal Register for a public comment period. Our evaluation of NE00 21778 includes our proposed change to Appendix G and is attached. In accordance with established procedure, we request that General Electric issue a revised version of the subject topical report which includes any supplementary.information pr9vided for our review of the report, this acceptance letter, and the enclosed staff evaluation. Sincerely.- arr of Light We'.dr Reettors Branch No. 3 Division of Project Management

Enclosure:

-Topical Report Evaluation .3 cc w/ enclosure: Mr. L. Gifford General Electric Company 4720 Montgomery Lane 8ethesda, Maryland 20014 l E

4 s,,y., ,,.:m. ;;.,,.ery; e,m,... r,.p,,, ,y., n,., gggo.31773.A.v c l usant TOPICAL REPORT EVALUATION 4 Report No: NE00 21774 Report

Title:

Transient Pressure Rises Affecting Fracture Toughness Require-- ,2 seats for toiling Water Reettors Report Date: January,1978 originating Organization: General Electric Campany Reviewed by: Engineering Oranch and Reester Safety Branch. Division of-Operating Reacters: Reactor Systems tranch, Core Perf *sence ? Branch, and Meterials Engineering Sranch of the Division of Systems Safety. k SISMARY OF RPORT i The report contains justification for a request from the Seneral Electric Company for relief from a provision of Appendis 8 '10 CFR Part 60 which sets minimum temperature limits for reactor core criticality. Paragraph IV.A.2.c. reads: Whenever the core is critical. the metal temperature of the-l reactor vessel shall be high enough to. provide an adequate-margin of protection against fracture, taking into account such factors as the potential for overstress and thermal shock during anticipated operational occurrences in the control of reactivity. In no case when the core is critical-(other than for the purpose of low level physics tests) shall' the tempera. ~ ture of the reactor vessel be less than the minimum permissible temperature for the inservice system hydrostatic pressure test nor less than 40*F above that temperature required by section l IV.A.2.b. The primary change requested by GE is elimination of the criticality limit t'ased on the temperature required for an inservice hydrotes.. A secondary change reovested by GE is the substitution of a 25 psi pressure margin for the 40*F temperature margin in paragrapn IV.A.2.c. - The effect of tha two changes requested by GE on the pressure. temperature limitt'for-a BWR with a radiation sensitive beltline material near end of life is shown by the shaded region.in Figure 1 (taken from Figure 21 of the Topical 1 1 J

~ 4 0 g ..e ... a > gg.g... 1 f. ) .f Report). At present, the shaded region must be traversed by heating with non. nuclear heat, such as heat from the recirculation pumps, whereas the t requested changes would permit taking:the reactor critical to obtain power for bestup in this region. The Report points out that substantial savings i l could be realized in startup time and in power replacement costs, in the discussion of NEDO.21778 at meetings'with GE people, it was explained that BWRs cannot use pump heat during startup as effectiveiy as PWRs can, 1 because the elevation head of water 'n the reactor alone is insufficient to i meet the NpSH requirements of the peps at all but the lowest speeds. tience, l pop heat is low until there is steam pres.sure is, the reactor. . g., The report addresses the concern expressed ta peregraph !V.A.f.c. of Appena dix G to 10 CFR 50 that there is a potential for everstressing the reactor i vessel as a result of some transient during reactor teartup or when. critical i due to a subsequent malfunction in the control of react 1Vity. The centrol 'l rod drop accident (CRDA) was selected as the lietting case for study, t despite its low probability of occrrrence, because CR04 conditions bound those of the anticipated transients that would have to meet Appendix 0. f The CRDA increases in total energy output and peak energy density directly with the assumed reactivity worth of the dropped rod, Rod worths are limited such that the peak energy density is less than 280 cal /ge, the staff limit criterion for the event. For this energy density limit there is no prompt ? dispersal of the 00: fuel into the coolant and thus no prompt pressure pulse, { and heat transfer from fuel to water, and thus pressure increase, may be deterutned by conventional heat transfer methods. For a 280 cal /gm bounding J event, giving maximum total energy output ~ the G.E. analysis shows that the 1 l resulting pressure transient should,not exceed 12.5 psi. i { With regard to the potential for a water solid pressure transient that i would violate the revised Appendix G Ifmits reconnended in the Report, the only events discussed by GE were a rod drop accident during a hydrotest or i concurrent with an inadvertent filling of the vessel to water. solid conditions.. No detailed analysis was provided for either event, because GE believes that 3 each is of sufficiently low probability to not be considered.

SUMMARY

OF THE REGULATORY EVALUATION-5 Backcround l During the evaluation of the Topical Report. consideration was given to additional infonnation obtained from meetings and correspondence as given in the attached chronological listing. Paragraph IV.A.2.c. was written into Appendix G during the final review by the ACRS in 1973.- to satisfy a concern that the reactor. vessel material be warm enough to be in a ductile condition before undertaking any startup l V

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.t + operations that could overstress the vessel if some malfuncti ecturred. Core criticality was chosen as the specific startup operation to make the - l requirement definite. The reevired taperature for an inservice hydrotest was used for the criticality limit, because that would represent the lowest i l temperature at which vessel integrity at pressures near the relief selee j settings had been esmonstrated on the vessel in its irradiated sendition. g Control Rod Deee Accident f The staff has accepted the use of the CRD4 as a limiting case with regard d to overstress of the reacter vessel resulting from a selfunction during the l control of reactivity. We also accept the reported estimate of 12.5 psi pressure rise, for the following reasons: First, by having previously-i accepted an estimate of ISO cs1/gm peak feel enthal , we have agreed that 1 a certain physical model of fwl danese is reasonab e.. ] The time constant for the process of heet transfer to the water is estimated et greater than 6 seconds, much too long for there to be significant nuclear-to. U -I mechanical energy conversion. Second, even if the calculated pressure rise i (12.5 psi) were off by a factor of 10, the effect en the vessel would not be i significant. In this calculation, water level was assumed to be normal, d: 1.e., the large vapor space in the vessel and in the steam lines cushions the pressurt rise. If the criticality limit is modified as requested, it is possible that the reactor could be taken critical to warm up the vessel for a hydretest. To further reduce the possibility of a.CR04 while the vessel is water solid for a hydrotest it will be necessary to add a requirement to Appendda G that all control rods mus*. be fully inserted during hydrotest. Despite general agreement that the 12.3 psi pressure rise estimate is reasonable the criticality margin of 40'F required by Appendia G was: retained. The GE reconnendation to substitute a 25 psi pressure margin, a = minor part of the requested change, was not accepted because: (a)25psiis a much scialler margin than 40*F at the temperatures of interest, and (b) it is a tradition of the transition temperature approach to state this sort of-margin in terms of temperature. Water Solid Pressure Transients Despite the generally good operating history of OWRs with regard to water solid pressure transients. the staff explored typical existing startup I procedures and systems aspects of potential vessel overpressure, goine well l-beyond the coverage in the Topical Report. As Can be seen in Figure 1, startup operations that normally occur af ter achieving criticality now take l place at temperatures above the shaded area whereas the requested rule ~ change would permit these operations to occur at lower temperatures where the pressure Ilmits are reduced and more easily exceeded. Violation of ' pressure limits in a BWR during startup could orly occur if water level was not adequately controlled. v1 -...................m..-. . _. -,. _.. -... - -.. _,.. ~..

NED0-21778-A I, In the meeting at the BWR Slaulator, in Morris Illinois, the staff received thefollowingadditionalinformationonwaterIevelcontrol. 1. .For the vessel to go water solid, the level would have to rise about 300 inches above normal (a volume of 1700 to 5600 cu. ft, depending on theplant)- i 2. There are three high level alarms. at 10,18 and 25 incher, stove l noma). - 3. During startup, there are several pimps that could supply water to the l msea Condensate booster W.,.:(shutoff head 400 - 600 psig, typically a. wov1d be running could not be tripped by hign level conditions.) b. Control rod drive pumps (shutoff head 1500 psig, typical flow L r 40 80 gpm) would be running. Feedwater pumps (shutoff head + 1500 psig) are not normally c. started until steam pressure reaches 300 to 600 psig, according to NE00 21778. steam driven pumps could not start at the temperatures of interest. The feedweter ' flow control valve closes on a high-yater signal if in automatic level control mode. The high level trip vou' d trip the feeduster pumps,

d. ' (he low pressure coolant-injection (ECCS) system (375 psig shutoff 4

head) could fill the vessel in 2 or 3 minutes if inadvertently i (No high level trip.) actuated. 1 The high pressure coolant injection system is steam driven and e. thus is not of concern here. l f. The high pressure core spray system trips automatically on high I water level. Yarious let-down paths are available during startup a: follows: Drain flow to the reactor water cleanup system is manually con-s. trolled as part of level control during startup. L b. The head vent is open to the dry well sump until a temperature of 'E about 190' is reached. at which time it is vented to the steam id l lines.

j The main stvarn line isolation valves and drain valves are open c.

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-{ 1-NE00-21778-A l l In sumary, we conclude that the probability of a water solid pressure transient in SWRs is sufficiently low and tMt the proposed lowering - of the criticality limits will not increase the probability significantly. For a water solid pressure transient to occur, (a) there would have to te an error or as1 function that caused an unanticipated inflow from some pue, (b) the operator would have to ignore three high. water alarms (in some cases, there would have to be failure of automatic protection feature), and (c) the normal let.dowe paths would have to be reduced through error or malfunction. One positive aspect of the proposed lowering of the criticality limits is t the reduction in time spent at temperatures where fracture is of concern, t 4 Such a reduction of time spent in this region would imply less chance of F occurrence of a maintenance or testing operation that might inadvertently flood up and pressurize the vessel. Regulatory Position The Regulatory Staff agrees that it is desirable to amend Appendix G,10 CFR $0 to concur in the requested deletion of the criticality limit based on the minimum permissible temperature for the inservice hydrotest. In its place, the criticality limit at very low pressure would be based on considers. tion cf fracture prevention in the flange regions that are highly stressed by the bolt preload. The Regulatery Staff does not concur in the requested change of the criticality margin from 40*F to 25 psi. The staff-recomended pressure temperature Ifmits are illustrated by the dashed Ifnes in Figure 1. A proposed package of amendments revising Appendices G and H is now being y reviewed within NRC, Paragraph IV.A.2.c. has become paragraph IV.A.1.f. in [ the current draf t. It reads as follows: 1, - When the core is critical (other than for the purpose of low-level ~~ physics tests) the temperature of the reactor vessel shall not be less i-than 40'T above (22 C above)-the minimum permissible temperature of l paragraph e. of this section nor less than 60'F above (33 C above) the referes.ce temperature of the closure fiansa regior.: that a e high*y . stressed by the bolt preload. All control rods shall be fully inserted during hydrotests. f viii _x_ _. - _. --_. - - - - - - -. _ _ _ _ _,,. _ _ _ _ _ _ _,. _, ,,_,,4 4 we,. _,,, ,yws,,_,,. np .-,y7p-y. ,gn,mgq-p.q,,4._,w,,,y+,,,,--g

ptDo=21778-A L i i { 1 ATTAC M NT t i CHRONOLOGICAL LISTING 0F MEETIN85 AND CORRESPONDENCE RE44R0!N8 NE00.fl178'- [ l. July 13, 1977 Letter MFN 270 77) to the Secretary of the Commission. U.S. N, from 81 den S. Sherwood, tienager, Safety and Licensi Nuclear Systems Divison, General. tiestric .. Subjects. ts en Proposed Change to. j i, 10 CFR 50 is8,'FractureToughness trements", s !! noted in 1 Register of March 3,1977, lume 42. { ' No. 62. Thir letter contained the first su from GE regarding the change to Appendia 8.ggestion-i., [ December 6,1477 Meetine at San Jose, CA, at which an early draft'of l' Nt00-21778 was discussed.

j ;;

January 24,1978 Letter to the Comeission, Atte: -01an D. Parr from j O J. F. Quirk, Manager SWR standardiastion. Transmitting the Licensing Top cal-Report, 'Transtant Pressure Rises - Affacting Pr4cture Tovehness Requirements for toiling. Water Reactors " Nt00 21778 F. E. Cooke. L. A. Kelley.. C. J. Paone, R. L, Shingleton and H. T. Watanabe.. March 16,1978 Informal meeting in Bethesda with. ze of the authors of the report. t l April 6. 1978' Li Mer (NTW 78-016) to' U.S. NRC. Attn: Mr. W. $. Hazelton from H. T. Watanabe. Principal Licensing Engineer. Operatin Plant Licensing General Electric Co., subject:- ) NE00 217 8 1 May 26. 1978 Letter to H. T. Watanabe from W. $. Hazelton..

Subject:

Nt00 21778, which transmitted informal questions for use as agenda for upcoming meeting... June 7,1978 Meeting in Morris, Illinois at location of BWR Simulater.. 2.

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I i ~-- .,,,.....mm,,.,,,,, STDC-31140 l 1 ABSTRACT 1 An evaluation was done to determine the impact of tb2 proposed Revision 2 to Regulatory Guide 1.99 or BWR operation. The guide 1 provides methods to predict shift in brittle fracture transition temperature due.to material irradiation. Operating limits for pressure test conditions were developed considering the snift ~ l predicted by Revision 2 compared to the Revision 1 shift. Very serious operational concerns related to higher vessel temperatures for. l l pressure tusts were obtained from the participating SWR Owners' Croup ) utilities and cosmon operational hardships were summarized. The concerns indicate that meeting the higher pressure test temperatures may endanger plant personnel and damage equipment. Estimations of the safety margin inherent in the current operating limits methods were made and alternative approaches were developed which lower the required veset1 temperature for pressure tests while maintaining adequate safety margin. This study was funded by the BWR - Owners ' Croup. The participating BWR owners are listed in Appendix B of the report. F 5; i. I h 1x/x l

NEDC-31140 { 1 2. SL'MMARY A0 CONCLt'SIONS As directed by the BWROC Committee on Regulatory cuide 1.99, a group of plants representative of most BWRs was examined to determine the impact of the proposed Rev 2 on operating' limits. A review of the available BWR surveillance data showed that the data agree fairly well with Rev 2. but the data base is rather limited. 11ntil a significant study of BWR data can be done. Rev 2 is a reasonable method to use in predicting BWR vessel shif t. The limiting beltline plate and weld RT shifts were calculated as a function of fluence using Rev 1 and WT Rev 2 methods to estimate the relative impact of each revision. Hydrostatic pressure test operating limits were identified as an area of operational concern; thus. EOL pressure test temperatures were j developed for each plant. including the E0L shif ts due to Rev 1 and Rev 2. Information was gathered f rom the utilities on the Cosunittee regarding operational difficulties that would result from more severe operating limits for pressure tests. The safety margins present in the current ASME Code Appendix C analysis were-determined and alternative analysis methods were evaluated for safety margin. The conclusions drawn from that evaluation are susnarized below. a. For almost every plate and weld chemistry examined. Rev 2 predicted higher shifts than did Rev 1. For typical BWR chemical compositions. the Rev 2 EOL shif ts were as such as 40'T higher for plates and 97'T higher for velds, b. The area of interest for the pressure test operating limits is the minimum temperature required at hydrostatic or leak test conditions. At reference prssoures of 1000 pais and 1100 pais. 200'T was used as the temperature at which.he transition f rom Cold Shutdown to Hot Shutdown or Startup conditions would exist. Based on the EOL shif t required by Rev 2, over half of the 35 BWRs evaluated would be forced over the 200'T transition. 1 1 2-1 l l _.. - _.. ~ _.

n .I NEDC-311I.0 i c. In order to perform a hydrostatic or leak test in Hot Shutdown or 4 l Startup rather than Cold Shutdown, most plants would have to make major changes to their Technical Specifications and outage schedules. Safety systems such as ECCS would need to be operational and primary containment would have to be isolated during pressure testing. As a result, the pressure test would probably become a critical path outage event. Inspection would take longer, requiring more inspectors and shorter stay times. Some regions, such as the vessel flange region.' would become dif ficult or lapossible to inspect once the primary containment i i was established. Other operational concerns related to pressure testing i.bove 200'T include pump operation problems (adequate NnR; heating capability, premature seal failure) and conflicts between -l hydrotest system requirements and-Hot Shutdown systes requirements. I Other means of heating the vessel, such as auxiliary boiler use, would be dif ficult and expensive to implement from a procedural and a hardware point of view. Nuclest heating to heat the vessel prior to pressure testing was identified as an option. However. for many plants this option would require. major revisions to 'echnical Specifications, operating procedures and outage l schedules and thus might not be practical. d. Appendix G of the ASME Code requires a safety f actor of 1.5 on stress intensity factor when determining hydrotest operating limits. However, there are inherent safety margins due. to several factors: (1) The Code uses a lower bound crack arrest (K3) relationship instead. of a lower bound crack initiation ( c) relationship. The difference represents a safety facter of about.'. 2-2 l

NEDC-31140 l [ 5. ETTECT OF SHITT CORRELATIONS ON PRESSURE TEST i OPERATING 1.IMITS Most 3VRs perform a hydrostatic pressure test (102% to 110% of operating pressure) once per ten years or after an ASME Code pressure boundary repair.- A few perform the hydrostatic pressure test ' af ter every outage. All BWRs perform at least a leak test (100% of operating pressure) after each outage. For the purposes of discussion here.-the hydrostatic pressure test and the leak test are considered together as a "hydrotest". Hydrotests are typically conducted between 1000 psig and 1100 psig. j '.. ~ Minimum permissible metal temperatures-for these pressures were caleviated using the meth W in Reference 4 assuming a 1/4 T flaw size and a hydrotest safet, factor of 1.5. The relationship between stress intensity f actor and temperature in Reference 4 is the crack arrest relationshipt IR = 26. 78 + 1. 233 e [0.014 5 (T - RTNDT * (5-1)' K The minimum permissible metal temperatures for the plants examined are ) shown for 1000 psig in Table 5-1 and for 1100 psig in Table 5-2. The -] j vdues of RT use in quati n 5-1 were the predicted Rev 1 and 2T Rev 2 E01, adjusted reference temperatures for the most limiting j beltline material, either veld or plate. The temperatures are calculated on a generic basis and may not reflect the plant-specific values in some plants' Technical Specifications. The hydrostatic pressure test is divided into t two. basic categoriest (1) the Cold Shutdown hydrotest. nd (2) the Hot Shutdown ' or Startup hydrotest. Cold Shutdown is generally considered the condition when the reactor is not generating steam. For the purposes of'this report.- l a Cold Shutdown hydrotest is one performed where the minimum vessel eetal temperature is below 200'T. The operational differences between i hydrotests in Cold Shutdown and Hot Shutdown or Startup are discussed in Section 6. 5-1 -..-_y.--, ...-wv. ~. .~r ---,r

I ............ - ~. ~,.. - NEDC-31140 The hydrotest temperatures in Tables 5-1 and 5-2 are censistently higher with the Rev 2 shift instead of the Rev i shift included. The i tables categorire each plant as " unchanged below". " changed". or " unchanged above". The additional shift resulting from Rev 2 increases the hydrotest metal temperature for sees reactors to a value above 200'T. These plants, which must change from Cold Shutdown { hydrotest to Hot Shutdown or Startup hydrotest, are categorised as " changed". Some reactors maintain a Cold Shutdown status for hydrotest given either the Rev 1 or Rev 2 shift (unchanged below), and s,me will have to perform a hydrotest in Hot Shutdown or Startup by l EOL given either the Rev I or Rev 2 shift (unchanged above). Plots of hydrotest limits, constructed according to References 3 and 4, are shown in Tigures 5-1, 5-2 and 5-3 as examples of typical plants that are " unchanged below", " changed" and " unchanged above", respectively. For the 37 reactors evaluated, the categories are summarized below: At 1000 paig - 16 unchanged below 14 changed $ unchanged above At 1100 psig - 13 unchanged below 15 changed 7 unchanged above Based on these results, over half of the operating BWF.a will be required to perform hydrotests in Hot Shutdown or Startup mode before reaching EOL. l l I ..I 5-2 l l l

NEDC-31140 i Table 5-1 i MINIMLH FERMISSIBLE TEMPERATURES AT EOL TOR HYDROTEST AT 1000 PSIC y h Hydrotest Hydrotest Temp. Temp. with Rev 1 with Rev 2 U -hanged Unchanged Plant Shift ('F) Shif t ('F) Below Changed Above Lrowns Terry 1 206 252 x. Browns Terry 2 113 210 x Browns Terry 3 173 210 x Brunswick 1 189 220 x Brunswick 2 219 253 x Clinton 134 228 x Cooper 199 243 x Dresden 2 158 192 x Dresden 3 155 217 x Duane Arnold 198 227 x Fermi 2 170 235 x l Fit: Patrick 165 199 x Crand Gulf 1 127 162 x crand Gulf 2 122 162 x Hanford 2 174 211 x Hatch 1 171 239 x Hatch 2 152 240 x Hope Creek 1 141 168 x LaSalle 1 190 227 x LaSalle 2 166 191 x l l Millstone 167 234 x 5-3 l

+ ~- NEDC-31140 I Table 5-1 (continued) Hydrotest Hydrotest Temp. Temp. with Rev 1 with Rev 2 Unchanged Unchanged Plant Shift ('T) Shift (*F) Below Chansed Above Monticello 205 226 x Nine Mile Pt. I 244 251 x Nine Mile Pt. 2 132 167 x Oyster Creek 174 188 x Peach Bottos E 148 182 x Peach Bottom 3 !$3 189 x + Pilgrim 146 182 x Quad Cities 1 145 176 x Quad Cities 2 133 166 x River Bend 1 145 201 x Shoreham 227 281 x Susquehanna 1 139 174 x r Susquehanna 2 136 166 x Vermont Yankee 134 175 x i 5-4

---

.------A 3-

-n-. -w w p-e.-- n w p

~ j NEDC-31140 Table 5-2 MINIMUM PERMISSIB1.E TEMPERATURES AT EOL TOR HYDROTEST AT 1100 TSIG I Hydrotest Hydrottst I Temp. Temp. with Rev 1 with Rev 2 Unchanged Unchanged Plant Shift ('T) Shift (*T) Below

Changed, Above Browns Terry 1 215 261 x

Browns Terry 2 182 219 x Browns Terry 3 182 219 x Brunswick 1 199 230 x Brunswick 2 229 262 x clinton 144 238 x Cooper 209 253 x Dresden 2 168 202 x Dresden 3 165 227 x Duane Arnold 208 237 x Termi 2 179 244 x Tit: Patrick 175 209 x Grand Gulf 1 136 171 x Grand Gulf 2 131 171 x Hanford 2 183 220 x Hatch 1 181 249 x Hatch 2 162 250 x Hope Creek 1 150 177 .x LaSalle 1 199 236 x LaSalle 2 175 200 x Millstone 176 243 x l 5-5 l.'

hTDC-31140 Table 5-2 (continued) Hydrotest Hydrotest Temp. Temp. with Rev 1 with Rev 2 Unchanged Unchan8ed Plant Shift ('F) Shift ('F) Below Chanted Above i i Monticello 215 236 x Nine Mile Pt. 1 254 261 x Nine Mile Pt. 2 141 176 x oyster Creek 185 199 x Peach Botton 2 157 191 x Peach Bottom 3 162 198 x Pilgrim 155 191 x Ouad Cities 1 !$5 186 x Quad Cities 2 143 176 x River Bend 1 155 211 x Shoreham 236 290 x Susquehanna 1 149 184 x Susquehanna 2 146 176 x Vermont Yankee 144 185 x l c \\ t I 5-6 .-r -~~'

_ _~ __ l NEDC-31140 l [ 6. OPERATIONAL CONCERNS The larger shift in beltline RTNDT inherent in Rev 2 e apared to Rev 1 i vill cause operational problems for some BWRa. The area of concern is the hydrostatic test operating limits. - The heatup/cooldown and core critical operating limits are not of concern because the BWR operates - under these conditions at saturation temperature for a given pressure, and saturation temperatures are well above the limiting tempetatures.. j The results from Section 5 show that many BWRs will be required to perform a hydrostatic pressure test at or above 200'F. In order-to qualify the operational concerns. this ' situation creates, CE sent a questionnaire (Table 6-1) to the BWR00 participants on the Reg. Ouido 1.99 Committee. The BWR00 coal. tee aesbers' responses are summarized in Tables 6-2, 6-3 and 6-4 The responses for individual plants are listed in Appendix A. 6.1 HYDROTEST DURING COLD SHUTDOWN A hydrotest at 200*T as summarized in Table 6-2, could be performed by most plants in the Cold Shutdown condition. Cold Shutdown is defined for most vessels as having the coolant below 212'T. Most of a f reactor's safety systems (about 701) are not. needed. -so a hydrotest and related repair work can be done in parallel' with other outage activities. Leakage would be visible during inspection as dripping or puddling water. The concerns raised in Table 6-2 which apply to the Cold : Shutdown hydrotest basically-state = that the higher the minimus test temperature, the harder it is to perform the test.. Pues heat and decay heat are usually relied upon to heat the vessel metal prior to hydrotest. The higher the minimum temperature, the longer the heating prreess, which may impact the critical path time of an outage and cost the utility between $100,000. and $400,000 per outage, compared to currently perforned hydrotests at temperatures of about 150'T. The higher the vessel metal temperature during leakage inspections, the grester the risk of burns and heat exhaustion for personnel. 6-1

1 i l l l NEDC-31140 L i

n cases where there is excessive decay heat. it may be impossible to provide adequate heat removal during hydrotest to keep the reactor in the Cold Shutdown condition.

This is only a problem if the target hydrotest temperature is 10'T to 20'T below boiling. Maintaining vessel metal temperature at or sbove 200'T is difficult or impossible without the coolant temptrature ever exceeding 212'T. Furthermore. some plants' Technical Specifications allow Cold Shutdown only for coolant temperatutes below 200'T. Therefore, some plants would be forced to perform a 200'T hydrotest ~ in the Het Shutdown condition or the Startup condition. I 6.2 HYDROTEST DURING HOT SHUTDOWN OR STAT. TUP Operational concerns related to performing a hydrotest at a minimum metal temperature of 22C'T are susemarized in Table 6-3. The significant differences between a hydrotest at 220'T and 200'T are due to the additional heating needed and the extra requirements placed on the reactor systems during Hot Shutdown or Startup. The operational concerns ~ on heating to hydrotest temperature applicable to Hot Shutdown er Startup are similar to those discussed in Subsection 6.1. but more severe. The issue of heating with pump heat is worse because of the higher test temperature. Most plants responding to the questionnaire believ. chey could reach 220*T with pump heat, but it would take longer. A test temperature of 220*T may be difficult for some plants to achieve.on pump heat and decay heat alone. Pump cavitation would probably occur before 220'T was reached, so the vessel vould have to be pressurized to an intermediate pressure that would prevent cavitation without violating the operating limits. The sectreulation pumps are designed to operate best at. ')00 pais. Operattin with the vessel at low pressures is an off-design condition. i so seals are not loaded as designed. While this off-design condition is not considered severe to the extent that it should be avoided. eperation at low pressures should be minimited to improve seal life. CE issued Services Information 1.etter 203 in October 1976, in which it _J sayst 6-2

I NEDC-311!.0 i "In restarting the plant or in using the pumps for raising the l primary coolant temperature (the practice at some BW plants), consideration should be given to minimiting the number of pump starts at low pressure (which in almost all cases vill be j atmospheric pressure)." Supplemental heat from auxiliar) boilers or available plant steam are options to consider for heating the reactor coolant, but either option would require major changes in operating procedures and probably to plant equipment, such as the condensate return. In the Hot Shutdown or Startup conditions, the ECCS systems must be operational, other safety system surveillances must be complete, and the primary containment must be isolated. Some' of the operable conditions conflict with the necessary conditions for the hydrotest. For example, the safety-relief valve low setpoint may be lover than the required hydrotest pressure. Major. Technical Specification revisions would be necessary to resolve these conflicts. The Hot Shutdown and Startup requirements may not add time to the outage schedule, but as a-result the hydrotest must be scheduled as a critical path activity in the startup sequence. During hydrotest, no other work can be done in the isolated primary containment, so scheduling becomes very dif ficult. I.eak detection is note dif ficult because any laaks vould be steam, not water. Inspection is more dangerous for personnel and radiation exposure would probably be worse in the event of steam laskage. The presence of steam would cause more protective gear, such as respirators and forced air equipment. to be required for the inspectors, increasing the chance of heat exhaustion and slowing the process of inspection. Cost impact figures in Table 6-3. ranging from $150.000 to S1.330.000. refer to the extra time needed for the pumps to heat the vessel to 20'T. There is no estimate of additional cost that vould be incurred if the now critical path hydrotest indicated that some repairs were needed. 6-3

i 1 h*IDC-31140 i \\ 6.3 USE OT h'UCLEAR HEATING BETORE HYDROTEST i t Since many of the operational concerns relate to the use of pump heat i to resch hydrotest temperature, the possibility of using nuclear heating was investigated. The latest revision of 10CFR50 Appendix G i (Reference 3) allows BWRs to initiate core critical operation at temperatures as low as the bolt preload temperature, as long as normal ater level is maintained. The questionnaire responses on nuclear _ ating are susmarized in Table 6-4 Most plants believe that nuclear heating would be too complex fros a licensing and operational viewpoint to justify reductions in the time to heat to hydrotest temperature. Some plants stated that there would be no significant time savings resulting from the use of nuclear heating. One of the plants surveyed performs the hydrotest at the end of the Startup sequence. The reactor is at about 1000 psip and $50'T when the hydrotest inspection is performed. The reactor is critical during the test to maintain pressure. Steam leaks are de.tected by sight, hearing and dripping from wet insulation. Heat exhaustion is a key factor, so a team of inspectors stays in the drywell for a short period. The ASME Code. Section XI allows hydrostatic testing at a reduced pressure of 102% operating instead of 110%, so the reactor high pressure SCRAM and the low setpoint on the safety-relief valves is above the test pressure. A few responses indicated that nuclear operation prior to the i hydrotest might violate Section XI of the ASME Code. Specifically. Table IW5-2500-1, Category B-P. Iten B15.10 has a reference to footnote 5, which says that a leakage test must be performed before plant startup af ter each refueling ourage. The definition of plant startup is determined by the wording in individual plants' Technical-Specifiestions. Startup in this case can be interpreted to be the point of power production, which occurs after heating by nuclear means. Eevever, some utilities interpret the footnote to mean that hydrotest must precede nuclear heatup. 6-4 i ~m ,.. ~.. _,. _

i hTDC-311!.0 t i s Table 6-2

SUMMARY

OT 200'T HYDROTEST OPERATIONAL INFORMATION Not Yes No

• Sure i

Is the test possiblet 12 3 Reached with pump / decay heatt 14 1 t Time required (hours)? 12-48 Operational Problems s.

1. Higher temperature than currently required, so additional pump operation time required to reach 200'T.

2. Pump cavitation near 212*T possible problem, so pressurization might be required to reach 200'T.

3. Technical Specifications require temperature below 200'T for Cold Shutdown and also require hydrotest during Cold Shutdown.

Incrsased pump operation time may cause premature seal failure. 4 f p l

5. Higher risk of burns for inspection personnel.

6. Difficult to maintain coolant temperature below 212'T and metal temperature above 200'T.

Cost from delays compared to current? $100K - $400K more per plant l-l i

• Pensibly due to Technical Specification requirements.

1 I l 6-7

_r-. NEDC-3111.0 Table 6-3

SUMMARY

OT 220'T HYDROTEST OPERATIONAL INTORMATION Not i ,N_o, " Sure Yes o i i 1s the test possible? 1 7 7 1 Reached with pump / decay heat?- 13 4 2 Time required (hours)? 20-96 Operationel Problems: i

1. Comments (1) to ($) on operational problems from Table 6-2 apply to a greater degree.

Pump cavitation and Technical Specificatbn conflict problems for most plants.

2. Primary containment isolation required, so no work in the containment can be done in parallell hydrotest becomes critical path event in outage, se problems cause significant delays.

3. ECCS and other safety systems required operational.

l

e. Operational safety-relief valves may pop during hydrotest pressuritation.

4

5. Leak detection more dif ficult 9

and dangerous because any leakage- . vill be steas; burn risk to personnel greater and more radiation I gear may be required. Cost compared to current? $150K - $1.330K more per plant Tessibly due to Technical Specification requirements. 6-8

.-~.o-i NEDC-31140 Table 6-4 NUCLEAR HEATING OPTION INFORMATION Not Yes No

• Sure Is nuclear heating possiblet 3

10 2 Time to reach 200'T.(heurs)7 4-16 Time to reach 220'T (hours)? 4-24 operational Problemst

l. Use of nuclear heating would require that the hydrotest be scheduled late in the outage, on criticti path.

2. Contradicts some plants' interpretations of ASME Code Section XI.

Table IWS-2500 requirement that hydrotest be performed before resuming operation.

3. Nuclear operation c:ust be done with normal water level; cannot be done with solid system needed~ter hydrotest at temperatures near 200*F.

This is not a problem if the hydrotet:t is at operating temperature.

4. CRD operability must be demonstrated before use; a hydrotest is required to demonstrate CRD operability.

i j

5. Hydrotest would be in startup

sequence, so any problems identified during hydr.stest would tapact startup schedulo.

a Possibly due to Technical Specification requirements. l 6-9/6-10

hTDC-31140 APPENDIX A NTDROTEST OPERATIONAL SURVEY RESPONSES The survey questionnaire shown in Table 6-1 of this report was sent to the participating utilities on the Committee. Plant-specific responses are summarised in this appendix, using the generic labels " Plant A" etc. 6 The questions posed on operational difficulties associated with high hydrotest temperatures were f airly broad. As a result, each plant's operators tended to concentrate their responses on the probles or problems they thought would be most significant. In most cases, the different te it s raised by each plant apply to almost all plants durni; e test. A few problems identified may or may not apply to all plau. ,,epending on the plant design and utility procedures. The responses common to most plants are listed in Tables 6-2 and 6-3 of the report. ll l l A-1

l NEDC41140 l l P! ANT A (SWR /2) A hydrotest at 200*T is possible, with pumps capable of heating tne j vessel to temperatura in 18 hours. This is about 10 hours longer than currently required for a 160'T hydrotest. This assumes tt.at all five recirculation pumps are run at samisus permissible speed which could cause premature pump seal failure. The higher vessel temperature would present a greater risk of burse to inspectors. The effect of a 200'T hydrotest temperature would be to add about one day to the outage. A hydrotest ac 220'T could be performed using pump heat to reach test temperature. Pump operation of about 22 hours would be required. The Technical specifications impose additional requirements above 215'T on l i reactor coolant Isakage. ECCS systems, primary containment and CRD coolant injection. Specifically. ECCS and other safety systems must be operable and primary containment-must be established. which means that isolation valves, containment spray and instrumentation for leakage rate, torus level and. torus temperature must be operational. In addition, access to the containment is controlled. These factors. along with the extra pump heating time.'would add lis to 2 days.to the outage. inspections would become significantly more difficult with the containment isolated. The safety risk of burns to personnel would increase with the higher vessel temperature. Any leaks would be steam, which are harder to detect. The possibility of airborne radiation with the steam sight necessitate extra protective clothing for the inspectors. Given the restricted access to the vessel flange region, inspection vould requ. re more people and shorter stay times to-avoid heat exhaustion. Nuclear heating prior to hydrotest is not considered feasible. l A-2 w--- - - -. ~ -- ~_...- - - -...,,,. -.-,,n, ~ ,...,.,,..u,,.,-

KEDC-311I.0 Pl. ANT B (SWR /2) A hydrotest at 200'T is possible using pump heat to reach test temperature. In fact, the current practice is to perform the hydrotest with the coolant temperature at 200'T (the minimun required metal temperature is about 187'T). The pump operation time required is 14-18 hours from a 100'T starting point. It is difficult to maintain coolant temperature below 212*T during the hydrotest, as required by the Technical Specifications. Reaching a test temperature of 220*T is possible with pump heat. but pump cavitation would be a problea. The vessel would probably be j pressurized to 200 psig to avoid -cavitation. Fatended operation of the pumps at atsospheric reactor pressure is undesirable because the recirculation pump seals are operating of f-design and because flow { causes some vibration of instrumentation in the core. The primary containment would be established and surveillances for operability of core spray, containment spray and liquid poison would have to be complete before the hydrotest. This would cause a scheduling tapact in the startup sequence, but would not necessarily cause a delay. Nuclear heating prior to a leak test has been done in cases where there has been no welding on the pressure boundary. The inspection in those cases was done at o,,erating pressure and temperature. However. I if any repairs are made to the pressure boundary, nuclear heating vill not be used prior to a hydrotest. Depending on the interpretation of-startup, it may not be permissible by the ASME Code to use nuclear heating prior to a hydrotest, i A-3 m

a NEDC-31140 PLANT C (SWR /3) I } Heating the vessel to 200'F for a hydrotest could be done with auxiliary boilers, but it is unknown if pump heat is adequate to reach 200'T. Temperatures of 160't have been attained on pump heat but not with the systen solid for hydrotest. Control of pressure / temperature using the recirculation pumps in a solid systen is not advisable. However, if pump heat would heat the vessel to 200'T it would take an f additional 24 hours. 1 Heeting the vessel to 220*F for hydrotest would not be possible because 1. Auxiliary heating boilers have limited capacity. Winter operation requires that the plant heating boilers operate near rated espacity for space hasting. If no nuclear steam is available from the other plant, the plant heating boilers are required to operate the radweste reboilers. These loads would not leave enough capacity to heat the vessel to 220'F. o 2. The radvaste system has' limited capacity to handle condensate return from the shutdown cooling heat exchanger assuming that heating boilers are used to heat the vessel. Overflow of the reactor building equipment drain tank would eventually result ~ in collection of radweste water on the torus basement floor, which is not considered good operating practice, since it increases the L of. contaminated floor space. The probles is worsened by amount i l the fact that radwaste water inventories are usually the highest 1 during an outage when hydrotests are performed. I l F [ A-4 l l L i ? i 1

1 NEDC-31110 ) 3. Coolant toeparatures in excess of 212'T require that primary containment be established. ECCS systems be operable and surveillances on numerous safety systems be complete. This would require that the dryvell head be in place during the hydrotaat. ) then removed af terward and replaced before Startup. This could add as much as $ days to a refueling outage. l Costs resulting from the extended hydrotest are estimated at $190.000 for a one-day entension to reach 200'T and at $1.330.000 for a seven-day extension related to a 220'T test. The use of nuclest heating to reach hydrotest temperature is not recommended because of water level considerations. Core critical operation at low temperatures can only be performed with normal water level, but the hydrotest must be done with the system solid. ( A-$

NEDC-311f.0 P1 ANT D (SWIL/3) A hydrotest at 200'T can be performed using pump heat to reach test i temperature, although pump cavitation becomes a concern the closer the I temperature is to 212'T. Assuming 702 pump speed currently used to schieve a 160*T temperature, it would take an additional 12 hours to reach 200'T. If pump cavitation were a concern, pump speed would be reduced increasing the' time required to reach 200'T. The cost of an additional 12 hours to perform the hydrotest is about $360.000 in lost l generation per cycle. i A 220'T hydrotest temperature could be reached on pump heat, but pump l cavitation concerns would probably require some pressurisation of the ^ i vessel. The extra time required, assuming 701 speed, would be 16 hours. The hydrotest would have to be scheduled after primary containment is established. This would restrict activities during the hydrotest. but should not add time to the outage beyond the 16 hours mentioned. The cost of the additional hours is $430.000 per cycle. Nuclear heating prior to hydrotest could be done, but there would be no time saved in heating the vessel, and the hydrotest would have to. be sche?aled later in the startup sequence, e A-6 .. ~.-.-.. - -.... - - - -

. _ _. _ _ _ _. _ _. _ _. _ _ _ _ _. _ _ _ _. _ ~. KEDC-31140 l FLANT D (SWR /3) { A hydrotest at 200'T can be performed using pump heat to reach test temperature, although pump cavitation becomes a concern the closer the temperature is to 212'F. Assuming 701 pump speed currently used to achieve a 160'F temperature it would take an additional 12 hours to reach 200'F. If pump cavitation were a concern, pump speed would be reduced, increasing the time required to reach 200'F. The cost of an additional 12 hours to perfors the hydrotest is about $360.000 in lost generation per cycle. A 220'T hydrotest temperature could be reached on pump heat, but pump cavitation concerns would probably require some pressurisation of the i vessel. The extra time required, assuntag 701 speed, would be 16 hours. ThS hydrotest would have to be scheduled after primary containment is established. This would restrict activities during the hydretest, but should not add time to the notage beyond the 16 hours mentioned. The cost of the additional hours is $480.000 per cycle. Nuclear heating prior to hydrotest could be done, but there would be no time saved in heating the vessel, and the hydrotest would have to be scheduled later in the startup sequence. 1 f 1 / 'h' .I A-6

i a l NEDC-31110 i F1. ANT E (SWR /3) i' I A hydrocest at 200'T is possible, using pump heat to achieve the test temperature. Compared to current procedures, reaching 200'T would require 2 hours more. Aneuning that metal temperatures reach about ths same value as the coolant at steady state. Cold Shutdown could be maintained. If metal temperatures (especially in the flause region) vere significantly lwer than the coolant, the conditions discussed i for the 220'T case would be required. i Pump heat could be used to achieve a 220'T hy'drotest' temperature. The heating time would be 4 hours longer than for the current procedure. l Since the coolant would be above 212*T. the primary containment would be required. This would require the safety system work to be completed earlier in the outage, prior to the hydrotest. The greatest impact would be on cutage scheduling, but there would be 18-24 hours 1 added to the outage duration. This would cost $135.000 to $180.000 in lost generacion per cycle. ) Inspection for leaks with the reactor coolant above 212*T would be acre difficult and dangerous because leaks would be steam, not water. 1 Tiange inspection would be particularly difficult because of the small space available in the flange region when the drywell head is in place. New. safe leak detection me 9 da would have to be developed. l Nuclear heating prior to hydrotest would not be feasible because core critical operation at low pressures requires normal water level. In addition, control red drives would have to be used before their hydrotest, which is part of the procedure to show their operability. Even if nuclear heat could be used. the plant has a policy of no access to the dryvell during any control' rod movements. These would presumably be required to maintain pressure and temperature during the hydrotest, so inspectors would have to periodically evacuate the dryvell. A-7 n,,, +, .--w-,- - -,,., - ~

l J I i i l NEDC-31140 j i i

i F1. ANT E (SWR /3)

A hydrotest at 200'T is possible, using pump heat to achieve the test temperature. Cor. pared to current procedures, reaching 200'T would y require 2 hours more. Assuming that metal temperatures reach about J. - the same value as the coolant at steady state. Cold Shutdown could be i asintained. If metal temperatures (especially in the flange region) j i were significantly lower than the coolant, the conditions discussed for the 220'T case would be required. i pump heat could be used to achieve a 220'T hy'drotest tosperature. The l 1 heating time would be 4 hours longer than for the current procedure. ) Since the coolant would be above 212*T. the primary containment would be required. This would require the safety system work to be t 1 l completed earlier in the outage, prior to the hydrotest. The greatest ispect would be on outage scheduling, but there would be 18-24 hours added to the outage duration: This would cost $135.000 to $180.000 in lost generation per cycle. j inspection for leaks with the reactor coolant above 212*T would be more difficult and dangerous because leaks would be steam, not water. Tiange inspection would be particularly difficult because of the small l space available in the flange region when the drywell head ie in place. New, safe leak detection methods would have to be developed. Nuclear heating prior to hydrotest would not be fossible because core critical operation at low pressures requires normal water level. In addition, control red drives would have to be used before their 2 hydrotest, which is part of the procedure to show their operability. t. t Even if nuclear heat could be used, t.ie plant has a policy of no access to the drywell during any control rod sovements. These would ~, presumably be required to maintain pressure and temperature durir.g the .tg hydrotest, so inspectors would have to periodically evacuate the dryvell. g' A-7

1 .\\ NEDC-31140 Another problem with hydrotesting after using nuclear operations is that many instrument cables would be attached in the vicinity of the CRD flantes. These would normally be' disconnected during the hydrotest, avoiding the possibility of damaging the cables' thile eliminating leaks at the CRD flanges. j l Regarding beating options. Plant E has ausiliary. steam supply to the RMR heat exchanger, but there is no provision for condensate return. Use of this availiary steam we 'd' first requits hardware changes to the plant piping. r I s ) 1 6 l i A-8

.. ~ - -.. .NEDC 31140 l l l . PLANT ~F'(BWR/3) l A hydrotest temperature ' of 200'T iould be achieved with pump heat. Assuming little decay heat, the f pumps would probably require 12-24 hours to recah temperatt:re. 'It would be difficult to maintain Cold t. l Shutdown cond.4.uten and keep the vessel metal'above 200'F,fso'it might be necessary ?.chedule the hydrotest af ter leaving' Cold Shutdown. The impact to the outage schedule would be - about-1 days,. costing $300,000 in lost generation.. ~ Pump heat could be used to reach a hydrotest temperature of' 220'T in: 24-36 hours. However, the test could not be done in Cold Shutdown..so ) the Technical Specifications would have to be revised. The safety

I systems required would not be available until later in tLe outage, so I

the hydrotest would have to be'reacheduled. The impact would be-about 1 2 days delay, with a lost generation cost of $400,000. Nuclear 1. eating prior to the hydrotest would mean that the test would be the last thing done before going ~ into operation. Any problems identified from the hydrotest would add. delay to the outage. There would be no opportunity to correct problems in parallei with other ' activities. l l l 1 j A-9

i -NEDC-31140 Pl. ANT C (BWR/3) ? ?; ant G normally employs' the use of a cold hydrotest. where the pressure vessel is.at approximately 1000 psig and is less than 212'T.

n situations where required (alternations, attache ** '

or repairs to the primary pressure boundary) an operational t hydrotest is employed, where the reactor is essentially at normal' operational eenditions_(1000 psig and up to 500'T). The cold hydroteet is performed at 1018 pois 25 pois and - 180*T to 212'T utilizingz the CRD pumps. Adequate margine above MPT are initially maintained' using RHR pumps (85 peig) with a subsequent.

heckpoint established at 500 pois.. Upon reaching all three indicated -

pressures, a rapid visual inspection' of system boundaries is conducted' and docuse.nted on procedural checklists. -i Upon conclusion of the cold hydrotest, reactor water level is returned to +157 to +167 inthes,'sain steam lines and HPCI and RCICI steam lines are drair '.

• reactor vessel is vented to atmosphere The RHR system is ti-e.ed (optional) to the shutdown cooling mode.

~he presad"ec for hydrotesting and leakate testing includes provisic<

• r

.,v.tatures of - 200'T. 300*T, 400*T and 500*T and' establi > J *r r soute range for each temperature. The hot s hydrote + o'* .leving repairs by velding on any pressure retaining Ad at c, near the -end of each ISI interval. The het hydrotest is done followi's the cold test and in accordance with routine station startup procedu esi l A -.-

o. h*EDC-31140 A normal reactor startup is conducted, the mode switch is not placed - in runLand pressure is controlled - using the mechanical pressure o regulator (MPR) with the electrical pressure regulator (EPR) in CUTOUT. The MR is used to increase system pressure to the appropr64t! ied ifenure. Following repairs once the test pressure is resaksi, n God time is established. as appropriate, dependent on whethat tie associated piping u iseulated' (4 hours) or uninsulated (15 minutes per inch). For hydretacing ~ at the end of an ISI' L interval, the established soak time is four hours. At the conclusion of this interval, a complete visualt inspection.of the system within the boundaries identified is performed..- [,,

• When completed, pressure is reduced with MPR, the EPR is energized and the appropriate action taken (continue startup.or. commence plant t

shutdown), depending on the results of-the test. f-Acceptance criteria established by the procedure are no mechanical

leakage, only slight leakage on the recire pump - seals and e

gi 20 drops / min, on the CRD housings. u I i 1 l A-11 j-.

~ . hTDC-31140 PLANT H (BWR/4) More restrictive P-T curves would require that primary containment be established prior co performing the hydrotest. Inspection for' leakage ~ in-the vessel-flange region would be very difficult because the drywell come would ' have to be in place. The vessel head compartment would have no. light, no ventilation and ' high temperatures. Since. leakage would.be steam, the possibility of high humidity and airborne radioactive elements 1 would exist. Access to the flange region when the dryvell does is installed 'is very limitedt thus, heat exhaustion would become a major safety : concern for inspectors, since removing them from the head compartment would be extremely difficult, if not impossible. As currently' scheduled, post-hydrotest work can be performed in parallel with pre-startup work. Assusing that primary containment is required auch of the pre-startup work would have to: be -done before hydrotest. This, combined with the added restrictions on work in the containment during and af ter hydrotest, would add nine daye s to = a typical outage schedule. If the outage schedule called for insulation to be removed for a hydrotest and replaced afterward, the ' primary containment requirement would add another six days to the outage.' If ( leakage were detected in the vessel head compartment. repair might require that the drywell head be removed, or that repair too11n6 be brought up from the botton drywell elevations. 'This would result in an additional delay of about one day, i l A-12

NEDC-31140 'L - l P1. ANT I (BWR/4) The following reasons are given for avoiding a hydrostatic test at or above the boiling points 1. Excessive time, and therefore cost, would be required.to achieve the hydrotest temperature. 2. Installation of a separate water heating system to heat the vessel prior to hydrotest would. be costly and add Lto,the complexity of the plant. 3. Above 200'F, the ability to d'etect leeks is impaired. 4 Many changen in the Technical Specifications and procedures would. be required because the hydrotest.. which is currently dene in Cold Shutdown, would have to be perforined in ' Hot Shutdown condition above 212'F. 5. No other work is permitted inside.the containment during: Hot Shutdown, so the outage would be-extended by tha hydrotest, i 55 F I 4: s$ ~ 5 . 13

l STDC-31140 1 l Pl. ANT J'(3WR/4) j l A hydrotest.at 200'T is possible assuming that there is adequate decay 1 heat to heat the vessel metal. Reaching 200*T with pump heat is I o questionable. With. cecay heat, the vessel temperature is simply } allowed to increase prior to hydrotest. The estimated time to. reach j 200'T is 4-6 hours. Hydrotest at a temperature of 200'T would be very l difficult to do in the Cold Shutdown condition.- Metal: temperatures of 200'T would probably require coolant temperatures above 212'T. ; As long as cold Shutdown can-be maintained, there is little impact to a higher hydrotest temperature. A.hydrotest at 220'T would require the~ plant to be in the Hot Shutdown condition. The following are' required by the Technical: Specifications: HPCI and RCIC systems tested and operable Safety relief valves (SRV) tested and operable Other safety systems operable Primary containment established. i Some of these requirements are in direct conflict'with the hydrotest conditions. For example, the SRV setpoint is lower than the pressure i required to meet the Section XI hydrostatic test ' conditions, making a valid hydrostatic test impossible at 220'T. Inspection for leak testing would.be ' more ' difficult because leaks ( vould be steam. Inspections. would be conducted in respirators or i forced air equipment, which would make access to some tight places difficult. The difficulties related to inspection could cause delays of 8-20+ hours. j A-14 s

l NEDC-31140 t L'se of nuclear heating would result in a severe impact on other outage 1 activities and scheduling. The option would be extremely complex atid probably would not save a significant amount of heating time prior to hydrotest. One absolute limitation would be the high pressure SCRAM at 1055 psig. In addition, core critical operation prior to hydrotest-contradicts the requirements of ASME Code Section XI, i Table IVB-2500-1. Category B-P. Iten B15.10. which says (Note 5) that systes leakage testing shall be conducted prior to plant startup following each reactor refueling outage. + t th e. l l l I l-A-15 /

NEDC-31140 PLANT K (B),1t/4). A hydrotest at.200'F.is possible, but relying. on pump heat and decay heat only for heating the vessel. If there is 2.ittle decay heat, i reaching 200'T by pump heat could take'as long as a week. The impact would be substantial. Inspection time would be increased, as would the number of inspectors required because of the higher drywell temperatures. The hydrotest would take 3-4 time as long as is currently required. I A hydrotest at 220'T could 'be -performed, ' but.the time to heat the vessel would be greater than for the 200'F case. Since the coolant y would have to be above. 212*F. ' the Technica1' Specifications require that primary containment be established. Leakage detection ~and operator / inspector safety could be a problem. The ' problem of increased inspection time and increased numbers of itepectors would be more severe. Nuclear heating to reach hydrotest temperature is possible. Temperatures could be reached of 200*F in - 2 honra 'and 220'T in 15 hours. However, use of' nuclear heating would-require that primary containment be established, all ECCS. systems, be operable : and - all relief valves be operable before ~pu111ng1 rods, which. causes serious difficulty in reaching temperature. Scheduling a hydrotest would be nearly impossible, since it would have to be done af ter all critical path work was completed. Subsequent leakage related repairs would .j require that the vessel be cooled and then reheated after repairs. I A-16 i

.} NEDC-31140 PLANT L (BWR/4) Plant' L has a Technical Specification requirement to have coolant temperature below 200'F to be in-Cold Shutdown. Therefore, a hydrotest with metal temperatures: of 200'T is not' possible without Technical Specification revisions. However, the ' pumps would be able to heat the vessel for a 200'F test. probably in about 10-48 hours. The main problem with a hydrotest in Hot Shutdown or Startup is that the ECCS systems suet be operable, but hydrotest conditions require i that systes _ pressure interlocke be defeated. This-could be resolved by Technical Specification revisions. Of course, the inspection during hydrotest would be. acre dif ficult ' becaust of the higher .*A temperatures. ~ The impact on outage lenkth would be an increase of about 10 hours. As above. 'a 220*F hydrotest could 'not be performed in Cold Shutdown. J The same restrictions apply as for 200'F and the same revisions would be needed to the Technical Specifications. Pump heat would be. 'i j adequate to reach 220'F. It would take 16-72 hours. The total I increase to an outage would be 12 hours. Nuclear. heating could not be used without receiving extensive licensing changes. The times required to reach hydrotest temperature are r.ot significantly-shorter than those-using pump heat. Aside from. licensing changes, there should be - no problems - with using. nuclear 5 heat. I h; I i -mm ammmei -am

NEDC-31140 //gY*b-PLANT H (BWR/4)' ' J Plant M employs the use of an operational hydrotest, meaning that'the t. reactor is essentially ready for' operation (1000 psig. 550'T) when the hydrotest is performed. The.hydrotest is nearly the last activity in an outage before resuming power production. A response to the questionnaire vas not received describing.the procedures used, but the following was determined by conversations with plant personnel. The hydrotest -is scheduled as,the_ last task prior to closing and' inerting t he_c on t aina,en_t.. The reactor has been heated in a normal Startup sequence to operating conditions. A team'of inspectors enters the dryvell to inspect piping.. weld overlays, look for packing leaks. Leakage is detected by the sound of leaking steam, by dripping etc. from insulation saturated by the leak,'or by observation of a' visible steaa jet. The vessel head compartment, to date, has not been entered to inspect the vessel flange for leaks. During the-inspection, control rod movements may be-made ~to maintain pressure and temperature. Once ~ the hydrotest inspection is completed, the final steps are taken to produce power, which is considered "startup" as noted in Table IWB-2500-1 of Section'XI'of the ASME Code. \\ 1 1-A-18 1 l + ~

L 1 L NEDC-311I.0 L 1: l '. F1. ANT H (BWR/5) ) The Technical Specifications' are written so that the transition ' f rom Cold Shutdown to Het Shutdown or Startup happens ~at a coolant temperature of.200'F. A hydrotest at 200'T and one at 220'T have the ~ 2 same requirements. Both temperatures' are ' above the limits permitted for hydrotest in the Technical Specifications. This write-up describes the differences between a 200'F or' 220'T hydrotest and one done in Cold Shutdown. The vessel could be heated with pump heat, assuming little decay heat, from 100'T to 200'T in about 10 hours, _ and in :12 hours to 220'F. As currently performed in Cold Shutdown, the hydrotest is done during the outage in parallel with - other activities. A higher' test temperature for hydrotest would -have no ' s chedule, as long as the Cold Shutdown condition is maintained. There would be a major administrative burden to move from Cold Shutdown to Hot Shutdown, as' the opposite normally occurs. -Instead, the plant would ' schedule the hydrotest - during the Ste.-tup - sequence. The primary containment would be required, mga ECCS instruments not needed in Cold Shutdown would have to be operable. all ECCS divisions would have to be operable, etc. In effect, the riant-~would have to be 100% operable in Startup 1nstead of 30% operable in Cold Shutdown. l { For a critical path hydrotest in ' Startup ' mode,- the de' ay = over a Cold ~ l 4 Shutdown hydrotest is about 8 hours, which represents $123,000-in lost I generation. Nuclear haating to reach hydrotest' temperature is considered possible but highly impractical. Major Technical Specification changes would be required. I A ! L u

'EC-31140 s Pl. ANT 0-(SWR /5) The Technical Specifications limit coolant temperature to 200'T in [ Cold Shutdown' and require that hydrotest be : performed in Cold i Shutdown. Therefore. a test at 200'F or 220*T cannot be performed j without a change to the Technical Specifications. Reaching such test i temperatures with pump heat is: possible.

• Assuming that little_ decay i

heat were available, it would-taka about two days to reach 200'F and four days to reach 220'F. Pugeavitation would be a probles as 220'F is approached, so the vessel would have to be pressurised partially to s. provide adequate NPSH for the pumps to operate in fast speed, I i i At coolant temperatures above 200'F. 'the primary containment is i isolated. sogone areas have.liaited access, especially in _the vessel flange region. Since leakage ~ vill _be steas, the possibility of. airborne radiation-is greater. Inspection would be done in respirators and possibly with ' forced air systems. The forced air systems are bulky enough that some locations could not be reached to be inspected. The_ extra gear required _ vould cause deleys in inspection. i Assuming that the Technical Specifications allowed these hydrotest temperatures.-it would require about-one day additional time compared to the present procedure to reach 200'F. which is-about 5270.000 in lost generation. The-additional time required to reach 220*F represents $540.000 in lost generation. J 7 Nuclear heating is a feasible option, but the Technical Specifications would have to be changed. The estimate to heat to 200*F is 4 hours. { and.4.5 hours to reach.20'F. i a l' u o A-20

i NEDC-31140 .l l 4 APPENDIX B PARTICIPATINC UTILITIES This report applies to the following plants,-whose owners participated [ in the report's development:- BWit OWN!R. PLANT Boston Edison Company -Pilgria. 4 Carolina Power & Light Company ~ Brunswick.1 & 2' Commonwealth Edison Company Dresden 1. 2 6.3 ~ Quad Cities 1 6-2 La Salle-1 &.2 Detroit Edison Company Fermi 2 Georgia' Power Company Hatch 1 & 2 Cenetal Public Utilities Nuclear Oyster Creek Gulf States Utilities River Bend ~1 & 2 Illinois Power Company Clinton-1 Iowa Electric Light & Power Company, , Duane Arnold Long Island Lighting Company Shoreham Mississippi Power & Light Company ~Crand Gulf'l &'2 Nebraska Public Power District Cooper; New York Power Authority- ' Fit: Patrick l Niagara Mohawk Power Corporation Nine Mile PointL1 & 2 Northeast Utilities Company Millstone-Northern States Power Company Monticello Pennsylvania Power & Light Company Susquehanna 1 h 2 ~ Philadelphia Electric Company Peach Bottom 2 & 3~ Public Service Electric & Cas Company Hope Creek'l & 2. Tennessee Valley Authority Browna Ferry 1, 2 & 3 Vermont Yankee Nuclear Power Corporation . Vermont Yankee Washington Public Power Supply System-Hanford 2 t B-1/B-2 l h IE.

} CHRONOLOGY ) i Apr 29,86: Bernero replies to Gibson(IV) L on interpretation of ASME XI regarding system leakage test. Position is that reactor should not be critical. Follow the appropriate P-T l j limits; liquid water only. Basis: prudence. -Defined Startup:. mode. switch to startup/ hot stby. i Stated this was.the intention of ASME Code. j j - May. 2,1986 Letter, Muller to GPC: o [ Noted:1) staff:had permitted such nuclear heat in past, l t

2) staff position:was late l8 3)would cause startup. delay,4)need to review proc.

~ 1 L{ 5)ininimal difference in safety,6)need to review Code i .j 71GPC promises to~ reconsider for Unit 2 l Allowed one time waiver; noted GPC had been usingffor years 1 ]

a ~ l i i April 10,87 Letter,Sniezek to O'Reilly i-Noted. Mar-13,87 incoming appeal letter and j appeal meeting of April 1,87 GPC had been using nuclear heat since plant. commissioning Letter to GPC-said no; cited as' basis can do better i i with cooler;and less hazardous environment. 1 l . April 25,1988 Letter, NUBARG to Murley ( Said.this was.a new position and:should -(but did ~not) comport with 50.109 Quoted minimal difference in safety. Cited App G IV.A.5~ provisions regarding the potential for. criticality.during test. j Claimed saftey improvementiin that would be testing i outside the brittle fracture-zone???). j Lots of legality regarding 50.109 2 1

l i i Aug 17,1988 Letter, Shao to NUBARG; Denied claim i i Said leakage test had to be done prior to startup, and mode switch is thus is shutdown, and FSAR says 3 i all rods must be inserted. l -NEDO 21778-A says all rods in; i So did staff evaluation. J Says staff not aware, prior to April 88 j ASME defines hydro test as liquid only. NEDC-31440(Jan 86), Appendix A' notes several plant views; safety problems associated with higher temp pressurizing fluids; not generally regarded as a good idea. i .a .l I 3

1 4 3 i i Sept 18,89 Memo,EDO to CRGR; directing review t Said our task was to:

1) consider the denial process (in particular j

the compliance exception of. 50.109? { l

2) evaluate possible ambiguity of regulations and the need for clarifying rulemaking.

I

3) evaluate the correctness of the staff l

interpretations (presumably including the safety question.) l r t Attached was the Mar 16,89 NUBARG' appeal to EDO.. which argued:new staff position, minimal safety affect, elevated temps avoid brittle region,more orderly startup, 1 l less wear on recirc pumps, cheaper? operation j l l 1

• +,-v+

a g arej .-y _ y s

3 i i i y i CRGR Staff Issue Sheet 1. Approved Staff Position? Five points (a-e) supporting view that it is a-backfitting, not issued pursuant to 50.109 t 2. Technical Merits? What is the safety problem? i is the problem well-posed any.where? l Is there an adequate: protection standard? 3. Argument against adequate safety standard (no case-made) and against compillance question. Thus: If needed, then do a more-complete,- rigorous analysis, pursuant to 50.1090. l y

i i CONSIDERATIONS i 1. App. A to NEDC-31140 is persuasive (by utilities) that nuclear heat is not the best way, and a 'might.be counterproductive. hy did not W .the staff make more use of this? l

2. Safety case not well argued or documented l

by any of the parties;.why not? l

3. Unclear as to whether this is a plant-specific or. generic matter. What are the utilities actually;doing?

i j 4. Regulations-do indeed seem a bit ambiguous. i .Why don't we propose a minor, clarifying change? l

5. What are the views of the ACRS?

l-l j a

8 i l . to the Minutes'of CRGR Meeting No. 172' Backfit Discussion in CRGR Review Packages October 25, 1989 TOPIC d The CRGR staff hAd prepared a draft memorandum to Office Directors concerning. i the contents of CRGR review packages. The object of the memorandum was to; clarify bacaritting considerations at CRGR meetings as well as-in published generic ;etters and bulletins. Bf,CKGROUND The draft memorandum was transmitted to CRGR members by a memorandum from E. L. Jordan dated October 20, 1989. A copy is attached to this enclosure. CONCLUSIONS / RECOMMENDATIONS i The CRGR considered the draft memorandum generally appropriate, subject to the following comments: 1. There appeared to be an. inconsistency between the compliance exception a discussion in the memorandum (no new or modified interpretations) and the enclosed examples regarding NPC Bulletin 88-01.- 2. In the memorandum, presentations should be-stated to focus on the technical subject matter as well as the basis and-justification under the backfit rule. '3. In the enclosed evample evaluation regarding NRC -Bulletin 88-01: ~ It was not clear that the cited. regulations required that welds be a. i j inspected. b. The _ statements regarding the application of "similar provisions"' and. the application "to various degrees-at various times in the past" were inappropriate. 4. In both enclosed example evaluations the wording "and they represent generic _ staff positions approved before 0ctober 21, 1985" was inappropriate. 5. In the enclosed example evaluation concerning IE Compliance Bulletin 86-01, a weakness was that the design had previously been reviewed and approved. However, this should not preclude using.the example. k p --w -,, + g --}}