ML20004E541

Proposed Findings of Fact & Conclusions of Law on Plant Design & Mod Issues (First Set).Addl Plant Separation Actions Are Necessary to Provide Reasonable Assurance That Unit Can Be Operated Safely.Certificate of Svc Encl

ML20004E541
Person / Time
Site:	Three Mile Island
Issue date:	06/04/1981
From:	Adler R PENNSYLVANIA, COMMONWEALTH OF
To:
References
NUDOCS 8106120315
Download: ML20004E541 (53)

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'b -s; WITED STATES OF MMCA i

.e{}u.s.JUtl 1119gj ,j- NUCLFAR REGULAIORY 03EISSION 4 N THE ATOMIC SAFETl AND LICENSING BOARD I d

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MEIROPOLITAN EDISON CWPAN't, )

Docket No. 50-289 4

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('Ihree Mile Island Nuclear ) (Restart) ocf,['"23^ h~

Station, Unit No. 1) ) 6-dlfy 8198;

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00ttDNWEAIjIH OF PENNSYLVANIA'S PROPOSED FINDINGS 9 #h '.$.5,}{7 -

"-4 0F FACI AND CDNCLUSIONS OF 1AW ON PIANI

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DESIGN AND FDDIFICATION ISSUES (FIRST SET)

A. INIRODUCITON .

The Cocmomealth will file proposed findings of fact and conclusions of Ira on plant design and codification issues in two sets, according to the schedule agreed on by tM parties.

The Occmomealth's approach to proposed findings and conclusions on plant design iss s is similar to that stated in the "Ccxmonwalth of Pennsylvania's Proposed Findings of Fact and Conclusions of law on Panageant Issues." 'Ihe Cctmomealth elects to exercise its right to

" advise the Cccmission" only on the discrete technical issues set forth below. 42 U.S.C. 52021(1); 10 C.F.R. 52.715(c). The Cccrorrmalth assumes that the Staff and the adversary parties will subnit ccuprehensive proposed findings and conclusions on the reemining plant design issues to assist the Board in its decision.

The Cecromealth has revicaed the cotire record on plant design issues, and suhaits proposed findings 46 cc..clusions only on those issues where the Cctrrnwealth seeks a rwlah renedy. The Cctrorw_alth does oc: adept specific findings and cnclusu.tr prcposed by any other party.

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Moreover, the Cormorwealth reserves its right to participace as a full party on all issues on appeal. Gulf State Utilities Cc; (River Bend Station, Units 1 and 2) AIAB-317, March 4,1976, 2 IUC. IG. REP. (CG) 130,053.

B. PROPOSED FINDItES OF FACT AND 00NCIljSIOIS OF IAW I. Burden of Proof and Standards of Compliance _

.l. The principles of burden of proof in IRC licensing cases are well settled. "thless otherwise ordered by the presiding officer, the applicant or the proponent of an order has the burden of proof." 10 C.F.R. $2.732.

In this proceeding, the Licensee seeks an order by tM Cocmission lifting the irmaiiate effectiveness of the August 9,1979 Order and Notice of Hearing. Slip op. at 15. Be burden of proof is thus clearly on the Licensee to demonstrate that such an order will be consistent with the public health, safety and interest.

2. De burden of proof is on Licensee on all issues in the pro ~aNg, regardless of the source of the issue in controversy. Tennessee Valley Authority (Hartsville Nuclear Plant) AIAB-463, 7 tEC 341, 356, 360 (1978);

thion Electric Co. (Callaway Plant) AIAB-348, 4 tEC 225, 227-31, 233.

For issues raised by intervenor contentions, the burden is shifted to Licensee by a showing sufficient to require reasonable minds to inquire further. Vermont Nuclear Power Corp. v. tEDC, 435 U.S. 519 (1978).

l l- 3. The nat.re of this proceeding is different from a normal construction permit or operating license hearing, as explained in Part A of this opinion. The Board does not believe, however, that Licensee bears any less stringent a burden of proof in this case than in any other rmmission proceeding. De nature of the proceeding does not affect the ftnvb=,mtal l

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principle that the utility is required to deconstrate that operation of the plant is consistent with public health and safety. Consmers Power Co. (Midland Plant, Lhits 1 and 2) AIAB-3'", February 27,1076, 210C.

REG. REP. (C G) 130,050.

4. De nature of this proceeding does, however, raise certain issues that are not emmnn to all NRC prorM4ngs. In particular, Licensee cepliance with long-term order items is judged by a " reasonable progress" standard. D e Board has had difficulty throughout the proceeding in det e ning what standards the Staff uses to judge reasonable progress.

Tr. 21, 059 (cermant by Chairman Smith).

5. The Board wants to make clear that a mere "comitment" by the Licensee to take an action at a future date is not sufficient to support a finding of reasonable progress. As explained by the 'IME-1 project manager, a conmi.tment by a Licensee is not enforceable per se. Tr. 21, 282 (Silver).

More frportant'y, a comitment without airy evidence of actual progress simply does not provide the Board with reasonable assurance that the long-term item will be couplied with.

6. ne Staff cinim to have reviewed Licensee "submittals" for es.ch l its en which a finding of reasonable progress was made. Tr. 21, 062 (Jacobs). The Board accepts this approach where it is evident that l

Licensee's subittals contain a description of substantive progress on the issue rather than a mere cocmitment to take action at a later date.

Reasonable progress must be evidenced by some concrete ca. tion, such as t

I the development of an acceptable design or the placement of the necessary purchase orders to cocply with the particular its. D e Board notes in this regard that complete ' delegation of issues to the Staff for post-baaring resolution is a practice frowned upon by the rermi ssion. Public

Service Co. of Indiana (Marbile Hill Nuclear Generating Station, Units 1 and 2) A1AB-461, March 1,1978, 2 NCC. REG. REP. (CCH) 130,274. The Board cust have scrae evidence on the record on which to reach a f4 Ming of reasonable progress.

7. The very nature of the long-tem order requiranents, of course, dic: :es that sac:e actions will occur after restart. This poses the problem of post-restart enforceaent. 2.e Staff's position is that, for significant itens, a license condition should be icposed prior to restart.

For other items, the Staff believes that otbar enforcanent actions, such as a show cause order, would be sufficient. Tr. 21, 260-63 (Silver).

Unfortunately, the Staff had not, as of May,1981, made arry decisions regarding what items would be considered for license conditions. Ibr did the Staff know when these decisions would be made. M. In fact, the Staff stated that these decisions would probable come after the Board's decision in this' case. Id. at 21, 263.

8. The Board is vested with specific authority in this area. The Order and Notice of Hearing states:

Satisfactory completion of the required actions will be detemined by the Director of Nuclear .

Reactor Regulation. However, prior to issuing its decision the Board shall have authority to require i staff to inform it of the detailed steps staff l believes necessary to implaaent actions the Board may require and to approve or disapprove of the adequacy of such measures. With respect to any uncocpleted items the Board shall have authority l

sinilar to that provided in 10 C.F.R. 50.57(b) to

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take such actions or to inpose such limitations or conditions as it believes necessary to prctect the public health and safety."

Slip on. at 13 (emphasis added).

9. In light of this authority, the Board is sacekat puzzled by the Staff's position. The Board, and not the Staff, may inpose license l

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conditions related to 'IMI-1 restart. 'Ihis position was accepted by NRC Staff counsel. Tr. 21,'352 (Cutchin). Consequently, it is not clear why the Staff elected to wait until after the Board's decision to decemine what items were ayyivy1. late for license conditions. In any event, the Board has weighed the issues that have been litigated in this i proceeding and directed that license conditions be inposed where ayy1.vyriate. The Board has insufficient basis, however,.to reach such decisions on uncontested issues that may be equally inportant and where license conditions may be ayyrvy1.iate. 'Ihe Board directs, therefore, that the Staff review all long-term requirements and NUREG-0737 items and make r m .. 4 ations regarding license conditions necessary to ensure the safe operatioa of 'IMI-l in the long-term. Since the Board will no longer have jurisdiction over this issue, these recocmendations should be certified to the Cocmission for daci< ion along with the certification of coupletion of short-term items.*

10. One further issue to be disposed of is the standard to be applied to coupliance with NUREG-0737 items.*

11. The Staff's method of judging reasonable progress toward coupliance with NUREG-0737 items was to cocpare the status of Licensee's compliance with other reactors. Tr. 21, 043-44 (Silver) . See Staff Ex. 14, at 2-3.

'Ibe Staff further judged whether there was reasonable assurance that Licensee would cmplete the action according to the IGEG-0737 schedule, although noting that additional schedule relief may be granted on certain

• All parties, of course, had the right and opportunity to propose license conditions in proposed findings of fact and conclusions of law.

'Ihis action is necessary only because the Staff stated that it would not make these deraminations in time to be includd in proposed findings and conclusions. For uncontested issues, the Staff is the only viaale source for these ran==mdations.

• • NURED-0737 contains the clarification of and current deadlines for 'IMI Action Plan Requirements.

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. . l itens. Id. at 21, 044. See also Staff Ex. 14, at 12, n.1.

12.In reaching these conclusions, the 1MI-l project managers assuned that daad14nas would be waived or amended wherever "a significant group

' l of utilities or plants could not meet [the schedule] for good cause shown." Tr. 21, 045 (Silver) . The 'IMI-1 project managers, however, are l

not responsible for deciding whether to extend NUREG-0737 dand14nas, nor do they know what daad14nes might be extended. Tr. 21 1 136 (Silver).  !

Moreover, Mr. Silver believes that all huEG-0737 deadlines may be amended '

without cwwdsing safety. Tr. 21, 045-46 (Silver).

13; However, Mr. Silver did not understand the basis for frposing dated requirements in NUREG-0578 or NUREG-0600, i.e. whether they result from a consideration of how long it was safe to operate the plant in its existing condition. Tr. 21, 051 (Silver). It is not clear, therefore, how the Staff's approach provides reasonable assurance that HE-1 can be operated in the long-term without endangering the health and safety of the public.

14. The Staff's position follows frce the Ca: mission's March 23, 1981 Order (CLI-81-3), which states:

The Coccission believes that Unit One should be grouped with reactors which have received operating licenses, rather than with the units with pending operating license applications. It arphasizes though that it expects the Board to find to the contrary when the record so dictates.

Slip co. at 7.

15. Due to this position, Licensee has delayed ccepliance for a nunber of items now governed by NUREG-0737, abiding by the later NUREG-0737 deadlines rather earlier Licensee ccr:r:d.tments cited in the Staff's SER.

1 Strictly interpreting the Ccx: mission's language quoted above, these delays I are pen:rissible, absent a showing that ng-l~shoG1Tbe creat5d'ilffferentif,J

from other operating reactors due to some unique circumstances. h difficulty with this approach is that the Staff's 'IMI-1 project managers testified that no 'IMI-1-specific review of NUREG-0737 items was underta.m to determine whether 'IMI-l warrants special treatment. Tr. 21, 118 (Silver /Jacobs).

16. h Staff also argues that Licensee was not permitted to rescind an earlier cocmitment in favor of a later NUREG-0737 daad14ne without a showing that the extension was necessary, for exacple due to procurement delays. h record does not support the Staff's position in all cases.

For exacple, de Licensee originally ccx:mitted to install reactor coolant systen high point vents prior to restart. Staff Ex. 14, at 53. When NURED-0737 relaxed the implanentation date for this item to July 1, 1982, the Licensee changes its coccit:nent accordingly, bb explanation was given in SER Supplement 3 for why this extension was necessary. Id.

Nor could the Staff's witnesses give any reason for this delay. Tr. 21, 313 (Jacobs).

17. h Cocmission directed the Board not to " reopen testimony or otherwise delay the proceeding" in order to determine when to treat 'IMI-l 1 differently from other operating reactors. Order, March 23, 1981 (CLI-81-3), Slip co. at 7. 'Iherefore, a cocprchensive analysis of NUREG-0737 items cannot be undertaken. However, the Board believes that the Cocmission did not intend to delay actions that will increase prctection of the public health and safety without good cause. Consequently, a strict standard is appropriate in det.Wning whether delays in Licensee cocmitments due to altered NUREG-0737 daad14nes is justified.

l In each case, the Board will weigh the reason stated by the Licensee for the delay, whether there is any valid reason for applying core stringent l

standards to M-1 than to other operating reactors, and the inportance of the action to the safe operatica of the plant.

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II. Plant Seoaration

18. Potential interactions between M-1 and M-2 were the subject of '

two short-term items in the August 9,1979 Order and Notice of Hearing:

4. 'Ihe Licensee shall demonstrate that Aa-aneandnation and/or restoration operations at M-2 will not affect safe operations at M -1. 'Ibe Licensee shall provide separation and/or isolation of M 1/2 r=Ainactive liquid transfer lines, fuel handling areas, ventilation systems, and sanpling lines _._

Effluent monitoring instrtrnents shall have the capability of discriminating between effluents resulting from thit 1 or thit 2 operations.

5. The Licensee shall deonstrate that the waste managenent capability, b1nAMg storage and processing, for solid, liquid, and gaseous wastes is adequate to assure safe operation of M-1, and that M-1 waste handling capability is not relied on by operations at M -2.

Slip op. at 6-7.

19. Although all intervenor contentions on plant separation issues were dropped, the Board directed that testimony be presented to da:enstrate ccxnpliance with these order items. Tr. 2397 (Board Questier. 8). The lack of specific contentions does not affect the independent responsibility of this Board to determine whether the required actions are necessary and sufficient to provide reasonable assurance that M -1 can be operated without endangering the health and safety of the public. Nor does it eliminate the. responsibility of the Staff to certify that the actions l

required by the Board and the en=iasion have been eccplied with prior to resucption of operation.

20. The Board finds that a ntraber of additional plant separation actions are necessary to provide reasonable assurance that M -1 can be operated with:ut endangering the health and safety of the public.

a. Fadinactive Waste Storage and Disposal

21. Licensee ships a large variety of radinactive wastes offsite for permanent disposal. See, e'.g., Fuhrer & McGoey, ff. Tr. 10, 020, at 9,

20. Licensee's testimony state's that "All radinactive solid waste from the operation ofLbit1, whether solidified or not, will be packaged and transported to a licensed burial facility in accordance with Department of Transportation and NRC regulations." Id. at 10. The Staff apparently accepted this cocmitment without analyzing the availability of offsite disposal facilities. Staff Ex. 1, at C4-9.

22. The Staff's witnesses were not able to add to the SER analysis.

Mr. Stoddard was not f='n414ar with the offsite waste disposal facilities available to the Licensee for the disposal of wastes from 1MI-1 and IMI-

2. Mr. Bellamy was only " peripherally f='414ar" with this issue. Tce Staff has conducted no independent analysis of Licensee's options should

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offsite burial become unavailable. Tr.10,195-97 (Stoddard/Bellamy).

j 23. There are three licensed burial facilities potentially available for the waste generated by M-1 and IMI-2. These are Barnwell, South Carolina; Richl=rvi, Washington; and, possibly in the future, Beatty, Nevada. Tr. 10, 025 (Fuhrer).

24. The Richland facility will only be avafinhle until mid-1981. Tr.

10, 025 (Fuhrer).

25. The Barnwell facility places restrictions on the amount of waste l

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that can be accepted frcm each operating facility. thtil mtd-1981,1MI-

! l's allotment is about 325 cubic feet per month. With this facility alone, the capacity of available disposal for M-1 may be in doubt.

Tr.10, (16 (Fuhrer).

26. The volume of waste that needs to be transported from 1MI-2 is l

4 different both in form and quantity as a result of the accident than uculd be true for a normal operating reactor. 'Ihis affects the ability of 'IMI-2 to attain adequate offsite burial facilities. For exanple, the disposal of zeolite wastes from the SDS system is restricted due to its high activity. In fact, the Barnwell facility has prohibited any, M-2 wastes from being shipped to that facility. Tr.10,152-54 (McGoey).

27. 'Ihe Beatty facility currently has no license for_ burial from the State of Nevada. Use of this facility is therefore in doubt at this time. Tr.10, 026-27 (Fuhrer).

28. Licensee has developed a draft contingency plan for the development of facilities for storage of radioactive material until appropriate burial sites are available. An interim wacte staging facility is expected to be constructed by the middle of 1981. However, it is not clear when the engineering aspects of Licensee's contingency plan will be completed.

Tr. 10, 027-28 (Fuhrer /htGoey).

29. Licensee's witness agreed that it is not desirable to store radioactive wastes on 'Ihree Mile Island for any long-range period. 'Ihe witness gave l a rough estimate, however, that the interim facility could be used for five to ten years without any serious concerts. Tr. 10, 031 (McGoey).

30. 'Ihe Staff's review of Licensee's prorosed interim waste storage facility stated that "it meets the intent of order item 5 and is acceptable I

j for radwaste storage up to two years." S*aff Ex. 14, at 20. 'Ihis finding was apparently based on the projected production of solid waste frcm a typical PWR for a period of two years. 'Ihe Staff has not reviewed the specific projected waste volume from M-1 and 2. Tr. 21, 403-04 (Stoddard) .

31. Licensee's witnesses testified that the interim waste staging

facility will have storage capacity for ayp w imately six months of waste generation for M -1 and 2. Normal operations are presumed for M -1. With nav4m a success of Licensee's waste generation reduction program, the facility will have storage capacity sufficient for one to one and a half years. Tr. 10, 029-30 (Fuhrer /McGoey).

32. 'Ihe Staff found coupliance with short-term order its 5 with respect to solid waste storage and disposal capability _on the basis of the planned interim storage facility. Staff Ex. 14, at 20. Based on the facts cited above, this determination was based on an incocplete analysis of the long-range solid waste disposal problem at M-1 and 2.

33. 'Ihe Board concludes that Licensee's plans for interim waste storage are adequate to meet the intent of short-term item 5. Ux:n I certification by the Staff that construction of 64 interim waste staging facility is couplete and that adequate measures have been taken to ensure that the M-1 portion of the facility will me be relied on by M-2, there will be reasonable assurance that M-1 can be operated safely in the short-term.

'Ihe Board also concludes, however, that additional requa.rements are necessary to provide reasonable assuraxe that M-1 can be operated without endangering the health and safety of the public in the long-term. Accordingly, the Board directs dat the following actions be taken:

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1. 'Ihe Staff is directed to review on a site-specific basis the capacity of the M interim waste storage facility, assuming that no offsite capacity is available. 'Ihis analysis should be based on f

conservative assucptions regarding Licensee's waste reduction cacpaign during normal operation of M-1. M-2 wastes cannot be excluded frcxn l (

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this analysis, parHenlarly if the M-2 portion of the facility may be expxted to be full befere the M-1 portion of the facility. In that event, it nust be denonstrated that no reliance will be placed on the M-1 portion to store wastes from M-2. Based on this analysis, the Staff should detamina a min 4==1 and mav4== anount of time that M-1 can be operated in the absence of available offsite disposal facilities.

2. 'Ihe Licensee is directed to develop a long-range contingency plan for the ultimate disposition of solid waste from M-1, given the current uncertainty regarding offsite disposal availability. This analysir ;uld take into account limitations on disposal of waste from M-1 that may result # rem the need tc dispose of waste from M-2 (for exanple, if an offsite disposal facil!.ty places limitations on the overall quantity of waste accepted from the M site). This plan should be reviewed by the Staff.

3. Based on the analysis conducted above, the Staff is directed to race ==and, if necessary, limiting conditions on operation of M-1.

Although such conditions should provide Licensee reasonable interim flexibility to resolve the problens of radioactive waste disposal, the Board wants to make it clear tnac it is not awcepciate to operate M -1 indefinitely until an acceptable ultimate waste disposal plan is developed.

This problem is core critical for M-1 than for other operating reactors due to the large volumes of radioactive waste already present on M as a result of the M-2 accident.

'Ihe Board erphasizes that final solutions to these problens are not l

a prerequisite to restart. In order to detennine reasonable progress, the Staff is directed to certify to the Conrnission prior to resunption of cperation that reasonable progress has been made with respect to items 1 and 2, above.

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b. Fuel Handling Building Seoaration

35. Short-term order item 4 requires separation and/or isolation M -1 and 2 fuel handling areas.

36. Licensee will not achieve emplete separation and/or isolation' of the fuel hand 14v, areas prior to restart. As explained in the Staff's Safety Evaluation Report:

The sapply and exhaust ventilation systems for the M-1 fuel handling building are . separate frczn those of M-2, but the TMI-l fuel handling floor ce=micates directly with that of M-2.

Because of the cc von fuel pool and other shared ccruponents, physical separation of the M-1 and M-2 fuel band 14ng area (is not feasible). To provide equivalent protection and meet the intent of item 4 of the Order, the licensee has cemitted in the Restart Report that (1) an independent engineered safety features (ESF) filter system will be installed at the M-1 fuel handling floor area for postulated fuel handling accidents, (2) the proposed ESF filter system will be in accordance with Regulae Guide 1.52. " Design, <

Testing, and Maintenance Lriteria for Post-accident Engineered-Safety Feature Atnespbare Cleanup System Air Filtration and Absorption Units of Light-Water-Cooled Nuclear Power Plants," (3) the M-1 fuel handling buildin6 layout will be modified to isolate the M -1 fuel handling floor from the M-1 auxiliary bi41dily and control access building to eliminate potential uncontrolled leakage paths, (4) the M-1 auxiliary btilding I

cnd fuel handling biilding ventilation system will be modified to prevent potential leakage paths between two buildings by addine two leaktight I dacpers in the discharge of the fuel handling I building air supply and in the fuel handling building ventilation exhaust main duct. The preposed modifications (items 3 ard 4) will be empleted prior to restart. The ESF filtration systen (ites 1 and 2) will be installed prior l to comencing the first refueling operation subsequent to restart. Since tbare will be no fuel movement in the M-1 fuel handlig area until that time, we find this acceptable. Iba l licensee has not yet provided detailed design infor=ation for the ESF filter systen.

Staff Ex. 1, at C4-8. See also Fuhrer & McGoey, ff. Tr. 10, 020, at 7.

37. %e Board finds acceptable this explanation for not providing couplete separation and/or isolation of the fuel handling areas. %is finding is based on practical considerations rather than a literal reading of short-term order iteci 4. However, in order to justify the proposed alternative to cocplete separation, the Staff is required to inpose rigorous standards to ensure that decontamination efforts on the Unit 2 side of the fuel handling building do not adversely affect operations either on the thit 1 side of the fuel bnd14ng buiHing or in the Unit 1 W 147 bi41 Mng. %e Staff has not conducted the rigorous review racessary to ensure that these standards are met.

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38. Se Staff accepted Licensee's two-phase approach based on the fact that no fuel handling will occur prior to,cocpletion of phase 2. Staff Ex. 1, at C4-8. B is rationale ignores other activities in the H -2 portion of the fuel ha%. building, as explained below. 2e Staff's evaluation should encompass both phase 1 and phase 2 of the proposed isolation progract.

l 39. ne Staff presented W. Phillip G. Stoddard to tescify to the adequacy of the proposed separation systecs. W. Stoddard did not personally review the proposed design. Nor was the design review conducted by the reviewing engineer, & . Boegli, formally documented by the Staff.

In fact, the only documentation of this review is apparently contained in an informal staff mano. Tr. 21, 387-88 (Stoddard)_. Bis mano has not been presented to the Board or the parties in this proceeding.

40. &. Stoddard was not aware whether W. Boegli conducted a separate review cf the adequacy of phase 1 of the isolation progract; nor did he make a personal determination of the adequacy of phase 1. Tr. 21, 391 (Stoddard) .

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41. We Staff did not review the EEF system to decemine whether it is adequate to isolate the Unit 1 auxiliary bi41 ding from the joint fuel handling bi41 ding. Rather, the Staff's review was limited to a determination that the system will meet the requirments of Reg. Guide 1.52. Reg.

Guide 1.52 does not address the adequacy of b'41 ding isolation; it addresses the criteria for engineered safety feature and filtration systems. W. 21, 388-89 (Stoddard). Ibreover, only phase 2 ites (the ESF filtratica syste) are covered by Reg. Guide 1.52. Staff Ex. 14, at 19.

42. Licensee's witness agreed that there will be scoe intWdng of air between the Unit 1 and Unit 2 sides of the joint, fuel handling building. Tr.10, 056 (Fuhrer). Berefore, any potential source of f

contamination in the Unit 2 portion of the fuel handling building is of potential concern to Unit 1 operations.

43. Fuel handling is not the only potential wce of contamination in the fuel handling building. Prior to Unit 2 fuel reoval, there will be operations involving the sihged dominaralizer syste used by License as part of the Unit 2 decontamination process. 'Ihase activities may involve the release of radioactive materials. Tr. 21, 391-92 (Stoddard).

44. We s11%ed demineralizer system (SDS) decontaminates bigh activity water (greater than 100 uCi/ml), principally from the 'LMI-2 reactor b d1 ding sucp. D e SDS is installed in the Unit 2 spent fuel pool in  !

the joint handling bt41 ding. Fuhrer & McGoey, ff. Tr. 10, 020, at 14-15.

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45. The SDS system produces a high activity zeolite resin material waste. The ultimate disposition of this material has not been resolved.

Meanhile, the materials will reain in a spent fuel pool in the joint fuel handling building. Tr.10, 061 (McGoey); Fuhrer & ltGoey, ff. Tr.

10, 020, at 20. l l

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46. Analysis of potential accidents related to the operation of th2 snhwrged d=4navalizer system indicates that there is a potential for airborne particulate contamination. Licensee's witness characterized 2

this potential is "very slight." He could not attach any precise probabilities or concentrations to this analysis. Tr. 10, 059-60 (McGoey).

47. 'Ihe Staff's witness Mr. Stoddard, testified that in the event of either an SDS or fuel drop accident at 'IMI-2, the M-1 fuel handling area would have to be sealed off. Decontaminaticn followir4!; such an accident could take up to 3 to 10 days. Tr.10,161-62 (Stoddard) .

48. 'Iherefore, analysis of potential SDS accidents, or accidents involving any other decontamination activities in the M-2 fuel handling area is necessary to determine whether phase 1 of Licensee's isolation program is adequate to safeguard activities in Unit 1. 'Ihis aralysis must be performed prior to restart.

49. 'Ihe Board also lacks assurance that Licensee's final isolation sf3 tem will be adequate to safeguard 'IMI-l activities during Unit 2 fuel handling operations. 'Ihis problem, however, need not be resolved prior to restarr.

50. 'Ihus far, no comprehensive analysis has been performed of the poter.tial for fuel handling accidants in the 'IMI-2 side of the fuel bnndling b d1 ding. 'Ihis analysis cannot be performed until the configuration of the 'IMI-2 core is understood. Tr. 10, 058-59 (McGoey). *

51. Mr. Stoddard's written testimony stated that "there is a potential that an accident, such as the dropping of a fuel transfer cask containing fuel removed from 'IMI-2, could result in radinactive contamination of the shared area, which could potentially spread to the portion dedicated to 'IMI-1. . . suspension of work in the M-1 area during 'IME-2 fuel reveJents

will be a procedural requirement." Stoddard, ff. R. 10, 159, at 22-23.

52. Accident analyses contained in Section 15 of the M-2 ESAR showed potential doses from fuel drop accidents in the hundreds of mi.llirens, using conservative assucptions. Fuhrer & McGoey, ff. Tr. 10, 020, at

42. It cust also be borne in mind that the M-2 defueling operation will involve both intact and ruptured fuel. Id. at 43.

53. For some unusual events, a full core offload of M-1 fuel might be required. In fact, such an offload as necessary early in the operating history of M -1. Tr. 10, 067 (McGoey). Licensee's witnesses agreed that an analysis is necessary of the ability to conduct such an offload durirg all possible necMents involving M-2 fuel handling. Tr. 10, 068-69 (Fuhrer /FcGoey).

54. Given the existing design, administrative procedures will be required to ensure cocplete isolation of the Unit 1 auxiliary building frm the fuel handling building during fuel handling operations from Unit 2. Tr.10, 055-56 (Fuhrer).

55. In Mr. Stoddard's judgment, a license condition should be inposed on M-1 prohibiting the hardling of fuel in the fuel ha%. building prior to cocpletion of phase 2 of the isolation syste. Tr. 21, 394 (Stoddard) .

56. Firally, the Board finds that the Staff has not imposed sufficient testing requirements of the proposed isolation system to ensure that M -1 opeations can proceed either in the sFort-term or in the long-term.

57. Licensee's written testfrony indicates that ventilation systen changes and the installation of separate filtration units will eliminate l

the ca:runication of air frcra unit to unit. Fuhrer & McGoey, ff. Tr.

l l

10, 020, at 7. 'Ibe Staff reached similar conclusions based only on descriptions of the proposed designs. See Staff Ex.1, at C4-2, C4-8.

No evidence was presented either by the Licensee or the Staff ro support these claima.

58. Licensee's witness was not f=414= with the testing requirments to determine whether this barrier and the ventilation system will be cot:pletely airtight. Tr. 10, 055 (Fuhrer). _.

59. Mr. Stoddard also was not aware of the requirements for testing the adequacy of the separation that will be constructed between the thit 1 auxiling building and the fuel handling area. Although s e e tests need to be performed to " balance" the air flows through the system, Mr.

Stoddard testified that this testing "does not in ard of itself assure that there would be transfer of air frcra one unit to the other." Tr.

10, 203-04 (Stoddard).

60. Counsel for the Staff indicated that a couplete test program would be required for restart, and that a request for a description was being prepared. 'Ihis was in January:

MR. 'IOUR'1TJIDI'IE: I wanted to make an announcement.

Questions were asked by the state relative to the testing program for restart, and I wanted the record to reflect that the. staff has discussed with the L9ensee the need for a eccplete test program for restart, and we are now preparing a request for a description of such a program and we will' review and approve that program. Awwsiate testing in regard to the Ormwealth's concerns will be l included.

Tr. 10, 215.

61. Yet by the middle of May, the Staff's 'IMI-l Project Managers had not reviewed Licensee's proposed pre-restart testing program. Tr. 21, 347 (Silver). 'Ihere is no evidence that leak testing of the fue'. handli:q area isolation systs will be conducted prior to restart or prior to fuel handlirg,.

i I , _ . - . . , - __ ,. . - . _ _ _ . _ . , _ . _ _ _ _ . _ . . _ _ _ ~ , . _ , . _ _ , _., _ ,___, ,,_,..__ . . ,,,_. ..,_ , _ . _ _ , , . _ _ , _ , , . _ _ _ .

l

62. 'Ibe Board finds that there is an inadequate basis to detamina that the requirments of short-term 1te 4 have been met with respect to separation and/or isolation of the M-1 and M-2 fuel handling areas and the 'IMI-l auxiliary building. 'Ihe following sbort-term and long-term actions are necessary to provide ram enable assurance that M -1 can be operated without endangering the health and safety of the public:

1. 'Ihe design of phase 1 of Licensm's separation system nust be analyzed by the Staff in terms of its capability to prevent contamiration of the Unit 1 auxiHary building in the event of an accident involving the SDS or other decontamination activities that will occur prior to the completion of phase 2. A testing program to ensure that the aypwp1.iate degree of isolation of the auxilian building can, in fact, be accomplished shall be designed and conducted prior to restart.

( Any leaks or other deficiencies in the syste shall also be cmpleted l

l prior to restart. 'Ibe Board wishes to emphasize that no delays in the Unit 2 decontamination process should occur as a resul'c of thit 1 restart.

'Ihe Staff is directed to certify to the Coccission that these actions

! have been cocpleted prior to restart.

2. A license condition shall be inposed on M-1 operation to the effect that M-1 may not operate during M-2 defueling operations prior to the cocpletion of phase 2 of the separation program. Moreover, l

once the configuration of the M -2 core is understood in sufficient l

l detail, a cceprehensive study shall be conducted of potential accidents involving M-2 fuel handling operations.* Phase 2 of Licensee's separation program shall then be analyzed to determira its ability to

• 'Ibe Board assumes that this analysis will be corducted in any event in conjunction with NRC regulation of the M-2 decantamination l

l effort.

l

safeguard the Unit 1 auxiliary building and the thit 1 fuel handling building against potential M-2 fuel hanA1hu accidents. A test program to deconstrate the adequacy of phase 2 shall be designed and conducted, and. any deficiencies in the design and construction of phase 2 shall be corrected prior to thit 2 defueling, or thit 1 shall cease operation. Based on these tests and analyses, the Staff shall recocmend additional license conditions, if appropriate, regarding-M-1 operation during M-2 defueling. Particular attention shall be given to the potential need to offload fuel fract M-1 during M-2 fuel handling operations.

3. Also prior to M -2 fuel removal, a study shall be conducted by Licensee identifying potential accidents at M -1 that may require offloading of M -1 fuel. 'Ihis study shall focus on the acount of time necessary to offload M-1 fuel, the acount of time after the event within dich the fuel nust be offloaded, and the effect of accidents in the Unit 2 fuel handling building on M-1 offloading procedures.

'Ihe Staff shall review this analysis and reW apprvpriate limiting conditions on operation for M -1, if necessary.

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III. UCS 4

63. UCS Contention 4 states that:

Rather than classifying the pressurizer heaters as safety-grade, the staff has pro psed simply to add the pressurizer heaters to tae on-site emergency power supplies. It has not been denonstrated that this will not degrade the capacity, capability and reliability of these power supplies in violation of GDC 17.

Such a daronstration is required to assure protection of public health and safety. _

64. Licensee presented testimony by Joseph A. Torcivia, Senior Project Engineer, GPU, and Paul J. Shipper, Jr. , Project Electrical Frginaar, Gilbert Associates, Inc., to address this contention.

65. The pressurizer heaters are normally powered frcm tm non-essential electrical buses. Power to these buses would be lost as a result of a loss of off-site power. In this event, the pressurizer heaters can be powered from the emergency diesel generators. Torcivia & Shipper, ff.

Tr. 9098, at 2; UCS Ex. 19, at 12.0.

66. Connection of the pressurizer heaters to the energency diesel generators was required as a short-term restart requirenent by Fer',andne4cn 2.1.1 of NUREG-0578:

Provide redundant arcgency power fe: the minbrn number of pressurizer heate;s required to mainea4n natural circulation ccrditions in the event of loss of offsite power. Also provide emergency power to the control and motive power systems for the power-oriented relief valves and associated block valves and to the pressurizer level indicadon instrument cbannah.

The Board ruled, therefore, that the effect of loading pressurizer heaters on the diesel power supply was within the scope of the Proceeding independent of UCS ContentLon 4. Tr. 9549 (Cba4mn SmLth). The folles issues, s therefore, were incorporated uto the hearing:

Careful attention should be given to assure that the capacity, capability, and reliability

l of the emergency power a rce (diesel generators) is not degraded as a result of inplementing the capability to supply selected pressurizer heaters from either the offsite power source or the energency power source when offsite power is not

, available. Furthermore, appropriate procedures )

and training will be needed to make the operator aware of when and how the pressurizer heaters should be connected to the emergency buses. The procedures should identify Ed~coiiditions iEid ~._

which selected emergency loads can be shed from  ;

the emergency xser source to provide sufficient 1 capacity for t'ie comection of preselected pressurizer heaters. Information required by the operator should be specified to determine ,

what loads can be shed under that conditions as well as the time required to cocplete load shedding and connection of the heaters to the emergency buses.

NCREG-0578, at A-3. Tr. 9550-51 (G airman Smith).

67. UCS argues that connection of the pressurizer heaters to the diesel generators may degrade th capacity, capability and reliability of the onsite emergency power supplies at 'IMI-1. Pollard, ff. Tr. 9607, at 4-1.

UCS also argues that this constitues a violation of General Design Criterion 17. Id.

68. GDC 17 states, in relevant part, that:

Criterion 17--Electric power syst es. .

An onsite electric power system and an offsite electric power system shall be provided to permit fmetioning of structures, systas, ard cetponents important to safety. The safety function for each i

system (assuming the other systen is not functioning) shall be to provide sufficient capacity ard l capability to assure that (1) specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded as a result of anticipated operational occurrences and (2) the core is cooled and containment integrity anl othu vital functions are maintained in the /

event of postulated accidents.

The onsite electric power supplies, including the batteries, and the onsite electric distribution systs, shall have sufficient independence,

redundancy, and testability to perform their l safety functions assuming a single failure.

10 C.F.R. Part 50, app. A.

t . - - - . . . - . - - . .- _.- . . , . _ . - - - - . - . . _ . , - . - . . . - . . ~

. . l

69. The 'IMI-1 sergency procedures require that " sufficient capacity be available for tha emergency diesel generator to handle the pressurizer heater loads as verified by wattmeter indication. Further, the rated capacity of the diesel has been verified as being capable of handling the heaters in addition to the safety-related loads required during loss of offsite power events." Torcivia & Shipper, ff. Tr. 9098, at 4.

70. Thus, Licensee's reliance on the transfer of the pressurizer heaters

-~

to the energency onsite powers supplies relies on the shMAbg of non-safety related loads.

71. Licensee's Emergency Procedure 1202-29 (EP 1202-29) ' covers pressurizer system failure. UCS Ex. 19. EP 1202-29 lists loads that may be removed from the diesel in order to ensure sufficient capacity to load the pressurizer heaters. UCS Ex. 19, at 12.0-13.0.

72. The continuous rating of the 'DII-1 diesels, as indicated in the

'DtI-l FSAR, is 2600 W. The 2000-hour rating is 3000 m, and the half-hour rating is 3300 m. Tr. 9412 (Torcivia).

73. Licensee revised EP 1202-29 to include instructions that the diesel load nust be kept at or below 3000 W. UCS Ex. 28.

74. If all of the loads that could bq added to the diesel were, in fact, loade.d, the capacity of the diesel would be exceeded. Tr. 9563-64 (Torcivia).

75. The equipment listed on page 13 of UCS Exhibit 19 was characterized l

by Licensee's witness as " nonessential." By nonessential, the witness l explained that the equipment was necessary for the safety of the equipment or for the ccxnfort of individuals in the plant, rather than for the safe operation of the plant  ? . 9507-09 (Torcivia/ Shipper).

.;24= -

l L .

i

76. At least two nonessential loads are autmatically loaded onto the diesels. h rest are apparently loaded manually. Tr. 9591 (Torcivia). 'Ihese include such items as air empressors, ~.he spent fuel ptmp, and the control building emergency supply fan, as indicated in Table 3-1 of Licensee's Restart Report. M.at9592. Licensee's witness did not know whether there are procedures covering tb4 loading of these itens on the diesel in order to ensure that the capacity of _

the diesel is not exceeded. h witness also agreed that these procedures should be in place and evaluated to ensure their adequacy. M.at9593.

77. Licensee's witness indicated that he would "probabl'/' go back to ensure that these procedures are in place and adequate. Tr. 9593-94.

'Ihere is no current Staff requirenent, however, directing that this review be conducted; nor is their any provision for NRC Staff review or analysis.

78. h Board finds that prior to restart, the Licensee should detemine that the procedures are adequate to ensure that all noressential loads being added to the diesel are adequately covered by procedures. 'Ibe Staff should certify that this review has been conducted and that necessary changes to energency procedures have been made prior to res~mt.

'Ibese procedures are necessary to ensure that the capacity of the l

i emergency diesel power supply will not be exceeded. 'Iberefore, this I review is necessary to provide reasonable assurance that 'IM -1 can be operated without endangering the health and safety of the public.

l l

L _

l m 25-i

IV. Instrunentation for Detection of Inadequate Core Cooling

79. h Staff considers cecpliance with item 2.1.3.b a requirement for restart. Staff Ex. 14, at 27-30. Included in this itan is the requirement to develop a design for instrunentation to provide an unanbiguous indication of core water level. Id. 'Ihe Staff's ultimate position was stated by Staff Counsel: "a showing of reasonable progress toward coupliance with the requirement for core water level instrunentation should be made a condition of restart, and that a showing of reasonable progress would be satisfaction of the five criteria set forth at tF4 bottom of page 29 of Supplement 3 to NUREG-0680 (Staff Ex.14]. Tr.

21, 207-08 (.Cutchin).

80. h Licensee asserts that core water level inst:unentation is not required to provide unambiguous indication of inadequate core cooling, and that existing instrunentation is adequate to assess core cooling conditions. Keaten, et al., ff. Tr. 10, 619.

81. Inadequate core cooling instrunentation was also the subject of three intervenor contentions (UCS 7, MEE 5(B), Sholly 6(B)). 'Ihese contentions were not pursued by the intervenors. '&e issue was litigated by the Licensee and the Staff.

82. Based on the evidence produced on this issue, as outlined below, the Board finds that the Staff's requirements for restart should be modified. .

83.' Licensee's testimccy indicates that there are several exacples of where the availability of core water level as a supplementary indication could tend to confuse the operator. For certain SBIDCA's, the water level in the core will decrease and expose the top of the core, with

the core still being adequately cooled. Tr.10, 661-66 (Jones, Keaten).

h operator should.take no action on decreasing core water level until in-cor,e therroccuples are higher than would be expected. Tr. 10, 682-83 (Keaten) . h Licensee gives exmples of the consequences of tahing inproper action before inadequate core cooling is present. Tr. 10, 667-76 (Jones).

84. h Staff gives several exanples of various conditions where the availability of the core water level insertment wuld be helpftil in properly diagnosing and following the course of a transient. Only in the case of using the reactor vessel venting system did they indicate that this instrumentation could be essential. Phillips, ff. Tr. 10, 807, at 2-5. 'Ibe Licensee rebuttal testfrony refutes these specific exmples by indicating that existing instrumentation would be adequate.

Tr.10, 632-37 (Ross, Jones) . Tr. 10, 772-73 (Jones). In the case of the reactor vessel venting systen, it was indicated that BW has developed operator guidelines which show that pressurizer level is adequate for conitoring operation of the reactor heat vent. Tr. 10, 692; 10, 777 (Jones).

85. 'Ibe Licensee testified that inadequate cxe cooling does not occur until a significant taperature rise has occurred in the core (to about a clad tmperature of 1400 F), at which point energency procedures call for action. 'Ihe in-core thermocouples and the expanded range hot leg RIDS are sufficient to measure this parameter and will respond adequately to this event. Tr.10, 620-25 (Jones) .

86. 'Ihe adequacy of the instruments existing at restart is explained in Licensee testimony as follows: "the Tsat meter as the indication that when the subcooling margin is lost, that inventory needs to be added, T 27-

, . 1 l

l l

and the core ther m couple as an indication that for an unknown, unidentified reason the ECCS is not operating as designed and that "here,is t starting to be an approach in the direction of inadequate core cooling." Tr.10, 730 (Keaten).

87. In the opinion of the Licensee, the core water level instrument is

" superfluous"; the in-core thermocouples are superior and less ambiguous for detecting inadequate core cooling. Tr.10, 655-57 (Jones, Keaten) .

88. In the opinion of the Licensee, during the DfI-2 accident'a core water level meter, if available, would not Fave indicated an inventory deficiency until the reactor coolant ptz:ps were tripped at about 100 minutes. The Tsat meter would have indicated an inventory problem at about 5 minutes. Tr. 10, 731 (Jones).

89. The Staff agrees that saturation conditions will occur before the core is uncovered with the possible exception of scxne large LOCA's.

The Tsat meter will be irstalled prior to restart and will indicate this condition. Tr.10, 829 (Phillips) . The Staff also agrees that the core water level meter " supplements and cottplements" other indications. It is.not adequate t i itself. Tr.10, 812 (Phillips) . In additional testimony by Stda senior managenent, the need for a core water level instrunent was characterized as a " core diverse instnrnent that would l

let operators cope with anomolous transients." The Staff witness agreed that the cw water level instnraent is strictly an additional and diverse metnod of determining inadequate core cooling and provides 1

I defense in depth for that event. Tr.15, 994-95 (Ross) . He also agreed l

that there is reasonable assurance that PWRs now operating without the core water level instrument do not endanger the baalth and safety of the public. Tr.15, 956 (Ross).

i l l

90. The only additional justification that the Staff made for finding that the core wate:: level instrunent is necessary is that it would have been helpful in following the course of two recent incidents where it was suspected that a steam bubble was formed in the reactor vessel head.

Ross, ff. Tr. 15, 915, at 3-5.

91. The Staff's current position was stated to be that licensees are required to install and test a proposed core water leve1 instrunent prior to the Staff deciding whether in fact the iratrunent is acceptable. Tr.

10, 824-25 (Phillips). This is in spite of the fact that the MIC is currently performing a generic evaluation of the differential pressure and heated junction thermocouple devices. Tr. 10, 834 (Phillips). This generic evaluation "is intended to assure us and others that the system will work and will not give false or misleading information." Tr. 15, 937 (Ross).

92. Whether or not the generic evaluation is coupleted, the Staff position is currently that licensees will be required to install the proposea devices by the first refueling after January 1,1982. Tr. 10, 834 (Phillips). A senior management Staff witness stated that "there is nothing sacred about the January 1, 1982 date." A firm date just provides incentive for licensees to find a solution. Installation, engineering and procurement by licensees will be proceeding hand in hani with Staff review and approval. Tr.15, 944 (Ross).

93. Currently the two most advanced methods for determining cere water level are the differential pressure (which is a Westinghouse design) -

and the heated junction thermocouple (which is a Combustion Engineering design) . Licensee testimony indicates that there are potential unresolved problens with both. Tr. 10, 760-61 and Tr. 10, 709-10 (Keaten, Jones). .

!29 .

94. h Licensee has done some additional studies which indicate that there are potential problems concerning how the differential pressure and heated junction cham ~ cale devices would interface with the specific B6N design. Because '. tne bottcm entry in-core thermocouple installation on B6N reactors, the heated junction tFarmoccuple design would require additional instnment tubing, since the Ccxtbustion Engineering d_ ?.gn (for which this device was intended) uses top entry thernoccuples. Tr. 10, 765-66 (Jones). The Staff witness agreed that this would be in fact a potential problem. Tr. 10, 869 (Phillips). In addition, Licensee witness stated that the differential pressure deiice would require an additional pene ration in the bottom of the reactor vessel. Tr. 10, 764 (Jones). The Staff stated that this would certainly affect their review in that a system which would require additioral penetrations would in wive the potential for an unreviewed safety question. Tr. 10, 870 (Phillips).

95. A Staff senior management witness stated that if it is determined that either device is not feasible for installation at B6M reac. tors, they 1

will permit a delay. "Any such delay would also require Licensee to l

ccomit to acceptable additional alternate raasures to be utilized in the interim that would cocpensate for the lack of required instnnentation, such as further operator training." Ross, ff. Tr. 15, 915, at 10.

96. h Licensee testified that they had been following what has been .

l l going on in indus* y with potential devices for measuring core water level. They stated that they "have nct seen aqr covergence of opinion I toward what is a good or acceptable device for measvring water level.

l l I would have to characterize all these efforts as still being in the development stage." Tr. 10, 707-09 (Keaten). The Licensee furtbar 30~

I i

stated chat: "we intended to pursue this through the"B E owners group in conjunction with the odier owners. We would take additional action -

i toward, development if those proved reasonable. So we have not closed the door on doing any further work." Tr. '10, 919 (Kenten). In ,

addition, the Licensee has requested a proposal to do an independent

~

evaluation of whather there are other promising ways of decamking l water leval other than those that aire being developed. ~ Tr. 16, 521-23  ;

(Keaten) . 'Ihe current BE position is "that the in-core therecouples cover the runge from normal operation to accident conditions, provide i direct, unambiguous indication of inadequate core coo?ing and also provide J

4 the approach to it." Tr. 10, 720-21 (Jones). 'Iberefore, until the BW

position changes, the Licensee can do very little through the B W owners ,

5 l group. ,

t  !

97. 'Ibe Board finds that the Staff has not made a sufficient showing that the core water level meter is necessary to provide reasonable  !

assurance that 'DE-1 can be operated without endangering public health and safety, at leas t on the schedule required for this Licensee. 'Ibe 1

i Staff's justification appears to be based prdmily on the premise that

! this device would add to the defense in depth of the plant safety i

features, that it would provide an additional arki desirable, but not a

. necessary, method of determining inadequate core cooling. It is therefore ordered that the Licensee does not have to satisfy the five (5) criteria of NUREG-0680, Supplement 3, Itan 2.1.3.b in order to show reasonable progress toward satisfying this item. Staff Ex. 14, at 29.

98. Recognizing that such a device would be desirable for the long l term, the Board directs that the following reasonable solution be inplemented. 'Ihe Licensee will be required to submit prior to res*st

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a detailed program for keeping up with the progress within the industrf for finding an acceptable device. Completion of this its will satisfy the reasonable progress criteria prior to restart.

99. h Licensee will be required, prior to January 1,1982, to sutxnit 4

for NRC review an evaluation of the feasibility and potential problans with installation of a differential pressure or heated junction type of core water level device. --

100. h Board finds that a generic NRC safety evaluation is necessary

! which shows that the installation of a core water level instrunent is I feasible and will provide accurate and useful information, prior to the Licensee having to cocmit to install such a device. Depending upon the results of the NRC genene safety evaluation of the differential pressure I and heated jun~ tion thermocouple dedces and depending upon the results of the Licensee's evaluation of the feasibility of installing these devices, the Licensee will be required to sdxnit a specific safety evaluation for installatian of a core water level instnunent at 'DfI-1.

Upon subnittal of this evaluation, they will be required to meet the six (6) criteria of NUREG-0680, Supplement 3.

i 101. If reasonable, based upon the above ccx:mitments, the Staff should require the Licensee to install a core water level insertroent at the i

first refueling after January 1,1982, which is the same as the current i

schedule for other licersees.

i I

7Afd'_/

v. Plant Cocputer Syste 102. 'Ihe intervenors pursued two contentions related to the adequacy of the M-1 cmputer syste:

Sholly Contention No. 13 It is contended that the thit 1 cocputer syste does not meet the requireents for instr mentation and control specified in GDC 13, and is inadequate to insure proper operation of the thit i reactor under all conditions of normal operation, including anticipated operational occurrences and postulated accident conditions.

It is further contended that the lack of real-time p & tout capability during accident conditions and the lack of sufficient redundancy in the cocputer syste place the public health and safety at significant risk during accident conditions, especially if cccputer function is lost and no back-up unit is available. It is contended that until the thit 1 cocputer system is upgraded to t meet the standards of GDC 13 ard until suitable l redundancy is provided within the emputer system to assure real-time printout capability at all times, permission for restart ::ust be denied on the basis of risk to public health and safety due to inadequate availability of operational infomation to Unit 1 operators.

ECNP Contention No.1(a)

The plant cocputer for M-1 is old, obsolete, and inadequate to respond appropriately in mergency situations. During the accident at the adjacent 'IMI-2, the alarm printer on the similm cocputer at lhit 2 had a delay time of over two and one-half hours at one point, and ran more than an hour behind events for over seven hours. 'Ihis delay cannot be viewed as having adequately served the needs of tb4 operators of

'IMI-2, and there is no reason to believe that a similar accident sit =Hnn, with as severe or worse consequences, cannot occur at 'IMI-l and be severely aggravated by slow and ambiguous cocputer alarm printer readings.

103. Licensee responded to these contentions with testimony by Willie P. Hamilton, GPU Manager of Process Cocputer Section, and Robert W.

Keaten, GPU Manager of Systecs Engineering. Hamilton & Keaten, ff. Tr.

7395, at 1.

104. Licensee's primary response to these contentions is that the M-1 plant computer system "is not required for safe start-up operation and shut-down of the plant. Licensee has designed M -1 to provide dedicated, hard wired instrumentation and control systecs on the control board."

Hamilton & Keaten, ff. Tr. 7395, at 2. See also Tr. 7422-23, 31, 44 (Hamilton /Keaten) .

105. Based on this testimony, Licen'see's witnesses testified that M-1 l is in conpliance with General Design Criterion (GDC) 13, which states:

Criterion 13--Instrtraentation and control.

Instrtraentation shall be provided to monitor variables and systens over their anticipacM operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systans that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate

. controls shall be provided to raintain these variables and systen within prescribed operating ranges.

J.0 C.F.R. Part 50, app. A.*

106. Licensee's witnesses also testified that: "In all cases where there is any effect on the safe operation of the plant, alternate insertraentation or canual procedures are provided to ensure safe operation I

should the cocputer system not be available." HactLiton & Keaten, ff.

Tr. 7395, at 3 (emphasis added).

107.The Staff's position is that the operators must rely on safety grade instrumentation at all Hmas during a plant transient. Tr. 7474-75;.

7494 (Joyce). The Staff's witness testified chat safety grade instrumentation

~

is obviously a more reliable source of information than the plant d&puter

• M-1 was designed to meet the 1967 version of 10 C.F.R. 50, i Appendix A. However, Licensee's witnesses testified that the requirements l of the 1967 version of GDC 13 are similar to the current (1971) version.

Tr. 7482-83 (Joyce):

Safety grade instrumentation as a general rule will be reviewed by the staff. Safety grade instrumentation has to satisfy certain criteria, standards and other regulatory requirements, GDCs, branch positions, et cetera. Therefore, the instrunent Athe sensor, the whole sensing line, et

- ~ ~ -

cetera is" called a higher pedigree, both frcxn enviwms:ntal standpoint, seismic et cetera; whereas a nonsafety piece of equipment, to illustrate, may very well be a Sears & Roebuck, local hardware store tegierature indicator.- It probably is not, but, you know, I am exaggerating '

the point, okay?

108. In fact, the Staff's witness testified that if a plant computer was used to perform a safety function, it would have to undergo tFa sane safety review process and criteria as other pieces of safety equipnent. Tr. 7503 (Joyce) .

i 109. The Staff conducts no review of nuclear power plant computer systems because they are not considered safety-related pieces of equipment. Tr.

7483-84 (Joyce) .

110. tb quantitative reliability studies have been performed on the W.-1 c m puter system. Tr. 7435 (Keaten). tb studies have been performed on the ca:puter's reliability in terms of either the amount of time it is l functional or the probability that it will given erroneous readings.

l l Id. (Hamilton).

l 111. Licensee's witness acknowledged that the computer might produce an incorrect reading. He clahi, however, that the error would not be undetectable since the operator has other means of checking the data coming off the ccx:puter. The witness acknowledged,'however,~that the operator would probably not check the cocputer information. The reading

" looked correct," or was in "the range of where he thinks it should be."

Tr. 7435-36 (Hamilton).

. _ . - ~ _ _ . _. .,

112. h M-1 computer is not powered from rdmdant sources of power.

Tr. 7433 (Hamilton).

113. h Board does not believe that reliance on mannal procedures is sufficient to ccruply with GDC 13 for critical safety-related functions during an emergency. Nor does the Board believe that Licensee's operators should rely entirely on the plant emputer to attain key safety information during the course of an emergency, even when the cocputer is "available."

Dese issues are considered sequentially below,

a. Incore Parmoccuoles 114. h incore ther occuples are necessary and essential for dete=dning when fradequate core cooling is befag approached. D e B6N position is that "the incere themocouples cover the range from normal operation to accident conditions provide direct, unambiguous indication of inadequate core cooling and also provides the approach to it." Tr. 10, 720-21 (Jones). It is also the Licensee's position that incore thermocouples are the " principle" method for detecting inadequate core cooling. Tr.

10, 750-51 (Keaten). Re Staff also agrees that incere &arroccuples are the " ultimate" method for dete- ining whether inadequate core cooling is in fact occurrin6, even if a core water icvel instrument were avail able. Tr.10, 893-94 (Fhillips).

l 115. Licensee utilizes a system of 52 incore dermoccuples with a readout l

range to 2300 F. Rese thermocouples are connected to the M-1 process caputer, and the therroccuple readings will be displayed on the Bailey l cocputer console by use of a GT, with hard copy also available on i

demand. Staff Ex. 15, at 10-11.

i i

116. he only backup systen for conitoring incore thermocouples that the Licensee has comnitted to prior to restart is taking local readings at the instrument cabinet with a portable voltmeter which then can be converted to ta:perature. It was indicated that this is the same backup system that was utilized during the TE-2 accident. Tr. 10, 748-49 (Keaten) . Bus, the only means of obtaining incore thermcouple readings in the control room relies on the plant corrputer systen. Tr. 7410 (Keaten/Fanilton) .

117. %e plant cor:puter was, in fact, used during the M-2 accident to obtain incore thergccouple readings. Hamilton & Keaten, ff. Tr. 7395, at 5. Rese events demonstrate the potential unreliability of relying on canual readings for important information such as thernoccuple readings. Incore thernoccuple readings were taken by an instn: ment technician and reported back to the emergency director "in a fashion that led him to conclude that they were unreliable." Tr. 7411-12 (Keaten).

Licensee's witnesses testified that thit problem would be remedied by

! the increased ta:perature range of the ca:puter printout (700 F at the time of the accident cor: pared to 23000F now). Tr. 7411-12 (Hamilton /

l Keaten) . Bis method, of course, still relies en the cocputer systen 1

for this critical information.

118. He Staff later found unacceptable Licensee's reliance on manual readings as a backup method of obtaining incere thermocouple readings:

We conclude that a better backup system is required because (1) of reliance on in-core l ta:perature infonnation in the inadequate core cooling procedures, (2) a system similar to the proposed backup systen was showc to be inadequate during the IE-2 accident, (3) the vintage of the present plant cocputer raises

! questions about the reliability of the information displays, and (4) the poor human factors interface of the proposed backup systen during stressful

~57-

operating conditions. We require as a minfara, data logging or recording equipment displays capable of displaying the temperature information for a minimun of 16 operable thermocouples, 4 )

from each quadrant, on demand, in the control room. A power source independent of the CRT power supplies shall be used to power this equipment to assure independence and reliability of displays. This system is required to be operational prior to escalation beyond 5 percent of rated power. This requirenent is revised and the licensee has not yet coumitted to its implenentation.

Staff Ex.15, at 11-12. Without this modification, the Staff could not find that the potential for operator error leading to serious consequences is sufficiently low to pen:d.t full power operation of IMI-1. Id.'at

12. The project manager for 1MI-l stated that this requirement is firm. Tr. 21, 362 (Silver) .

119. Currently, Licensee cocmits only to provide the required backup display systen according to the deadlines set in IUREG-0737, which is currently January 1, 1982. Staff Ex. 14, at 29. There is no assurance by the Staff that the January 1,1982 deadline will not be extended further. Tr. 21,136-37 (Silver).

120. The incore thermocouples are also important for other accident situations besides inadequate core cooling. The emergency procedures direct the operator to use the incore thermocouples to determine downcomer 1

ts:perature for the criteria which allow throttling of high pressure injection so as not to exceed the pressure temerature limits of the reactor vessel. Clark, Ross and Patterson, ff. Tr. 6225, at 6-8. The other instruaent that is called out in the procedure for determination criteria is the Tsat meter, and this meter does have backup safety grade te@erature and pressure sensors that can be used to determine cargin to i

saturation. Tr.10, 740-46 (Keaten).

738-

121. A Staff witness representing senior management has stated that if it is not feasible to install the differential pressure or heated thermcouple core water level device at B& W reactors, then a delay in inplementation will be permitted. However, alternative measures need to be comnitted to in the interim that wuld cor:pensate for the lack of this required ins,trunentation. Ross, ff. Tr. 15, 915 at 10. A backup display for the incore thermocouple could qualify as_an additional measure for this purpose. h same Staff witness also testified that it would be important to make sure that the bac16pTincere thermocouple was

a part of the training program on inadequate core cooling. Tr. 15, 701 (Boss).

-122. The Board finds that the Licensee is required to install a backup hardwired indication for the incore therm:<:ouple as specified by the Staff position, and to provide training for its use prior to installation.

This finding is based on the importance of the incere thernoccuple in l determining inadequate core cooling and in its use as an instrument l

t called out in the emergency procedure to decide when to throttle high I

pressure injection. Based on the Board's position with respect to core water level instrunentation, the incore thermocouple becomes a cuch nore inportant instrunent.

123. The Board concludes that these requirenents are necessary to provide reasonable assurance that 'IMI-l can be operated without endangering the l health and safety of the public.

b. Operator Reliance on Computer System 124. The incere thennoccuples are the only instruments for which there is evidence on the record that Licensee places primary reliance on the 1

t

plant ecxrputer to obtain key safety information. ,However, there is also evidence %t at the operators may rely on the ccx:puter, if " is available, to obtain safety information during plant transients. In addition, the Staff apprently did not review all important safety parameters to determine whether they can be measured using safety grade instrunentation in the control room. 'Ibe Board wants assurance that these concerns are addressed prior to restart. --

125. 'Ibe Staff's witness on Sholly Contention 13 testified that Licensee cocplies with GDC 13 because all important safety parameters are measured using hard-wired instrumentation. 'Ihe witmess did not know, however, whether reading incore therroccuples we an important enough parameter to be included in the GDC 13 requirement. He was not fd14- with the necessary accident analyses; nor was it his function to decide what parameters are required to be displayed in the control room using bard-wired instrumentation. Tr. 7468-70 (Joyce).

I 126. 'Ihe basis for the Staff's conclusion that GDC 13 is cocolied with, i

l therefore, is not clear. 'Ihe Board's conclusion that incore thernoccuples are the only case of improper reliance on the cocputer can only be based on a negative inference from the state of the record. 'Ita Board directs, therefore, that the Staff cordet a cocplete review of the key safety l parameters necessary to ensure plant safety, and to determine that all such parameters can be read in the control room using reliable, hard-f wired instrunentation. Satisfactory results from this review rust be certified to the Cocc:ission prior to 'DfI-1 exceeding 57. power in order l

l to provide reasorable assurance that 'IMI-l can be operated witbout endangering the health and safety of the public.

( 127. Licensee's witness agreed that the potential for operators becocdng,

overly depeMent on the plant couputer is s valid concern, particularly if the couputer is used for basic plant safety functions. Tr. 7420-22 (Keaten) .

128.The Staff's witness agreed that "there is nothing to prevent the operator from using the conputer whenever he sees fit to do so." Tr.

+

7485-86 (Joyce).

129. There are certain situations in which the form of information 4

available from the GT is core convenient than from other instrtrnentation.

In these cases,, the operator might choose to rely on the GT. Tr. 7428 (Keaten) .

• 130. Although Licensee's witness testified that the operators use the l computer very litrie during a transient, the conputer is used to call up data during a transient. Tr. 7413 (Keaten). For example, the conputer is the most " convenient" place to read PORV tail pipe temperatures. The alternative method of obtaining this information, a nultipoint recorder, 1

may result in some delay. Tr. 7414 (Keaten).

t 131. Licensee's witness agreed that installation of the new GTs will make it more likely that the operators will use GTs to obtain information.

Tr. 7429 (Keaten).

132. 'Ihere are certain functions for which the operator normally uses the ccmputer, such as the "flur. tilt in the reactor core, the unbalance between the top and the bottom of the core, in order to assure that the plant is operating within its tech spec limits..." Tr. 7441 (Keaten)

(erphasis added).

133. The 'IMI-l procedures in general do not direct the operators to use particular instruments to obtain the required informacien. Tr. 7458 (Keaten).

134. Licensee's witness testified, however, that there are probably some procedures that instruce the operator to rely on the coguter if it is available. Tr. 7442 (Keaten). Conversely, the witness knew of no procedures that instruct the operator not to use the cot:puter, although he acknowledged that he was nec f=miliar with all plant procedtres. Tr.

7443 (Keaten).

135. Based on the above, the Board does not find that adequate measures P2ve been taken to prevent unwarranted operator reliance on the cocputer during plant transients. The Board directs that Licensee revise its cmergency procedures prior to restart to reaedy this problem. Each emergency procedure that requires the operator to obtain key safety parameters should clearly instruct the operator to use only reliable, hard-wired instrumentation. The cot:puter may be used to verify the information received from the prd=ary source. Moreover, the Licensee shall ensure that its operator training programt quately e:phasize l

l the need to use hard-wired instrumentation as the primary sources of

! information under emergency circunstances. The Staff shall review this program and certify its cocpletion to the Catraission prior to 21I-1 exceeding 57. power.

136. These actions are necessary to provide reasonable assurance that 1

i T4I-l can be operated without endangering the health and safety of the public.

1

VI. Board Question 11 137. Board Question No.11 inquired into Licensee's coupliance with and NRC Staff review of cocpliance with Racermandation 2.1.9.C of NGm-0578:

'Ihe board is not satisfied with the staff findings in the SER with respect to Pae% tion 2.1.9.C (transients and accidents) of NUREG-0578. The staff concludes that satisfactory progress has been made and the item i is couplete. SER, pp. B-10, C3-49. Accordfig to Table B-2, the analyses and procedures were scheduled for coupletion by early 1980. We observe that in May of this year, it was reported that "the Staff is perfor=ing a generic review of transients and other accidents in accxdance with Re ~+ Antion 2.1.9 of IUREG-0578" (NUFM-0667,

p. 5-26) .

Recocoendation 2.1.9.C requires Licensee to:

Provide the analysis, a::ergerry procedures, and training to substantially irprove operator performance during transients and accidents, including events that are caused or worsened by insygwriate operator actions.

138. Licensee addresses Facermandation 2.1.9.C by participating in the l

Abnormal Transient Operating G34dalines (ATOG) program conducted by Babcock and Wilcox (35). Broughton, ff. Tr. 10, 941, at 2; Jensen, ff.

Tr. 11, 005, at 2-3. Although ATOG appears in concept te meet the requirec:ents of Recocoendation 2.1.9.C, the Board is not satisfied that Staff review of ATOG has been adequate to fird reasonable cocpliance

- with this long-term order iten. Such review is necessary to provide .

reasonable assurance that 'IMI-1 can be operated without erdmgering public health and safety.*

139. The goal of ATOG is to produce a " single procedure, and therefore a single set of operator actions, based on key plant parameters, without

• Operator training in the new ATOG procedures was discussed in Findings 106 to 111, suora.

[

requiring that the specific initiating event be known." Broughton, ff.

Tr. 10, 941, at 4. Licensee's witness testified that E UI " differs considerably in approach frcm the procedures that have 'seen used traditionally." Tr.10, 948 (Broughton). The approach used in ATOG is to treat the symptoms of plant transients without necessarily knowing what tha specific initiating event is. In fact, Licensee's witness testified that an operator "may get two-thirds of the way tirough the procedure without e.rer having identified the initiating event." Tr. 10, 948-49 (Broughton). This is essentially a new, untested approach to accident response. The Board is concerned, therefore, that all possible ramifications of this new appraoch may not Fave been carefully evaluated.

140. Licensee's witness agreed that it was necessary in developing the ATOG procedures to assure that the generic steps taken by the operator following a plant transient were appropriate for all transients, regardless of the specific initiating event. Tr. 10, 950 (Broughton).

141. ATOG development also involved additional analysis of certain transients (small steam line breaks, loss of feedwater, loss of offsite power, excessive feedwater addition, and steam generator tube rupture).

Broughton, ff. Tr. 10, 941, at 2-3. These analyses should be reviewed by the Staff as they would during an FSAR review. Board Question 11 l

l focused not only on Licensee's compliance with Re htion 2.1.9.C, but also on the information and analysis necessary for the Staff .to find I

that reasonable progress has been made on this item.

142.One of the primary goals of ATOG is to develop a document which can be used by control rocxn operators during a transient. Broughton, ff.

Tr. 10, 941, at 2, 4. Thus, friplementation of ATOG will require incorporation and modification of existing 1MI-l emergncy procedures.

M. at 4 ATOG implementation will involve the actual cancellation of many existing procedures, as well as alterations of other existing procedures. Tr. 10, 953 (Broughton). This makes Staff review of ATOG prior to iglementation all the more igortant.

143. Licensee does not intend to conduct coglete simulator evaluations of the TM-1 ATOG procedures prior to iglementation. Rather, "ATOG (will) be walked through, as a mininun, at the plant And there would also be scue evaluation of those on the simlator." Tr. 10, 952 (Broughton) (ecphasis added). No information was presented by the Staff regarding the adequacy of Licensee's pre-ig lementation evaluation of the workability of the ATOG procedures.

144.The Staff witness on Board Question ll, Mr. Jensen, was not the NRC Staff member responsible for conitoring Licensee's ATOG cogliance schedule. Ibr will &. Jensen continue to monitor ATOG compliance. Tr.

11, 006 (Jensen). In fact, &. Jensen did not understand why he was the ,

i witness selected to represent the Staff in this proceedure. Tr. 11, 017 ,

(Jensen). Staff counsel represented to the Board that Mr. Jensen was selected because he "was already going to be here as a witness...he has as good a view of what the status of progress is as arrjone else." Tr.

11, 018.

145. The Staff's witness did not know whether the IRC had formulated any positions on what should or shculd not be included in the procedures.

In particular, the Staff apparently had not determined whether it was appropriate to treat transients sytptomatically. Tr.11, 009 (Jensen).

To date, the Staff has reviewed only generic guidelines; review of specific procedures has not yet begun. Id.

146. The Staff's witness fail =:d to undertake an adequate review of the

45~

_. _ _ . ~ . _ _ . - - - _ - _ _ _ . _ _ . _ _ _ _ . _ _ . . _ . . _ . ~ . . - . _ _ _ . . . _ . _ _ _ -

ATOG program to find that reasonable progress had occurred with respect to Recomendation 2.1.9.C. Mr. Jensen did not look at the Arkansas draft operator guidelines in any detail, yet concluded that "it seems to i

be fairly detailed work; and I can see that a lot of effort has gone into this particular document." Tr.11, 014-15 (Jensen).

147. Other Staff members apparently have reviemd the Arkansas guidelines in some greater detail. However, even this review was. not in sufficient detail to determine whether they were satisfactory. Tr. 11, 016 (Jensen).

148. The discussion of ATOG in SER Supplement 3 evidences n__o additional analysis on the part of the Staff. Essentially, only the anticipated cocpliance dates are changed frcxn the original SER. Com are Staff Ex.

1, at C8-49, with Staff Ex.14, at 46.

149. The Staff's ATOG testimony stated that "for the interim period before emergency procedures based on ATOG are cocpleted, the IEC is revicaing the current Emergency Procedures for TMI-1." Jensen, ff. Tr.

11, 005, at 3. The Board does not understand how this review is perceived by the Staff to be relevant to ATOG. The Staff cust find reasonable progress with Re% tion 2.1.9.C indgendent of existing procedures.

150. Licensee used the draft guidelines from A10-1 (Arkansas Nuclear) in order to develop their training and changes to procedures. Licensee believed that the TMI-l guidelines would be very shWr "in terms of format" but that the content would differ to reflret the specifics of 1MI-1. Tr.10, 947 (Broughton).

151. The Staff accepted the Arkansas guidelines for all B W plants in order to meet the January 1, 1981 dead W e for generic guidelines. Tr.

11, 016-17 (Jensen). There is no firm evidence, however, of how similar the TMI-1 guidelines will be to the Arkansas guidelines. While generic gn4da m w s mla be quite ayyrvyclate, the Staff did not have adequate information to make this det e ntion.

152. One of the Staff's objectives in suggesting the develognent of generic guidelines for plant design was to " minimize the anxnt of plant-specific procedure review and the approval required." Tr. 11, 007 (Jensen) (quoting IUIED-0737, at 344).

153. Staff review of the gereic B6N progran is not sufficient to find reasonable progress toward KIOG inplementation at DfI-1. According to the Staff's witness, "bacause of plant differences each of the B6N plants will require different guidelines and emergency procedures."

Jensen, ff. Tr. 11, 005, at 3.

154. Licensee had not, as of January, 1981, received any feedback from I the NRC on whether the generic guidelines were ayytvysiate. In fact, delayed Staff review may affect the KIOG implementation schedule. Tr.

10, 955-56 (Broughton).

b. Implementation Schedule 155. Licensee's proposed KIOG implementation schedule appears to meet l the requirements of NUREG-0737. NCREG-0737 requires that procedures be l

revised by the first refueling outage after January 1,1982. Staff l

Ex. 14, at 46. Licensee's current "cccmitment" is to have tlw KIOG procedures in place by the first refueling outage after January 1,1982.

I Tr.10, 955-56 (Broughton); 15, 584 (.Capra).

i 156. 'Ibe Board does not feel, however, that Licensee's "conmitment" to early coupliance warrants any lesser degree of Staff review. 'Ibe Staff has yet to confirm Licensee's cocpliance schedule.

157. In fact, there is evidence that Licensee's coupliance has slipped

- - - - _ - - _ . . - -~ . _ . - . - . . - . - . . . . . . , - - - _ - . . - - - . - . . - _ . - - . . - . - - _ _ .

a

a nunber of times. Mdalhes for plant procedures were due January 1, 1981. Jensen, ff. Tr. 11, 005, at 2. As noted earlier, Licensee submitted the AND-1 gnida14nas to meet this daad14ne.

158. B6N experienced some slippage in the ATOG development schedule between August and October of 1980. Tr.10, 953-54 (Broughton).

Licensee's witzvess testified ti.2t the delay "in general...was due to the fact that this was a first of a kind program in which_schdules and j costs are sometimes hard to estimate accurately." Tr.10, 954 (Broughton).

Moreover, the witness testified that further delays may result after the draft guidelines are issued: ,

It is possible that when we see the draft 'IMI guidelines, that it will take us longer to review cocoents and resolve ccuments...The estimate of Feptanber to be able to implement .

these also has same uncertainties in it with respect to how close the draft guidelines are to the final guida14naa and exactly how noch work is involved in interfacing these new

, procedures with the existing procedures So we would not view those as cemitment dates, but those are rather target dates...

159. At the time that the Staff testimony on ATOG was presented (January, i

1981), the Staff was relying on out-of-date scheduling information, even though correct informatier. had been available for approximately 3 conths.

Tr. 10, 953-54 (Broughton).

160. The Staff'first learned of the changes in Licensee's inplanentation schedule by reading Licensee's ATOG testimony. Tr.11, 006 (Jensen) .

The Staff's witness based his testimony on the "only schedule that the Staff currently [had) available." M.

161. In January, Licensee dia not expect draft ATOG gaidelines to be couplete until April,1981. Broughton, ff. Tr. 10, 941, at 5. The final M-1 ATOG was not expected from B6W until July,1981. Broughton,

- ~ _ . _ _ _ . _ . . _ . _ _ _ _ . . _ ~ . _ - - _ _ . . . ~ _ _ _ _ _ . _ _ . _ . - . _ _ . _ ~ - - . _ _ _

ff. Tr. 10, 941, at 5. Coupletion of the 'IMI-1 procedures is now expected in Septecber of 1981, evidencing an apparent further delay.

Staff Ex.14, at 46.

162. Staff counsel represented to the Board that the Staff's basis for finding reasonable progress was that it had "no reason...to believe that they will not achieve their present intentions." Tr. 11, 018. Part of the rationale for the Staff's approach was that ATOG Qas not a contested i issue. M. at H , 019. 'Ibe Staff's determinations should be based on

analysis rather than assumptions.

i t

I

c. Scope of ATOG 163. The Staff intends the ATOG procedures to include an accidents and transients that might occur at the plant. This includes cultiple failure l events such as cultiple tube ruptures in a single steam generator and tube rupture in more than one steam generator. ATWS events are also j

included. Tr. H , 010- H. (Jensen).

l 164. The events and transients included in Redation 2.1.9.C

\

includes events that have been analyzed in the past, such as loss of coolant events and loss of unin feedwater events, as well as events that have not been analyzed in the past, such as inadequate core cooling and nultiple stean generator tube ruptures. Tr. 11, O n-12 (Jensen).

See also, M. at 11, 014.

165. ATOG procedures cover only transients that involve a reactor trip, "with the possible exception of a very sman steam generator tube rupture."

! Tr.10, 958 (Brcuginton). No Anticipated Transient Without Scram (ATWS) events are covered by the procedures. Id. at 10, 961; 10, 973.

l l

166. Licensee explains this limited reach of AIOG by citing analyses which show that transients and accidents that do not cause react $r trip are not severe enough to cause da-nge to the plant. Tr. 10, 960-64.

Licensee's analysis of this area, however, appears rather indefinite.

For exaraple, referring to integrated control system failures, Licensee's witness testified that "if the upset was not severe enough to cause a l

j scram, then it was probably a fairly cdnor upset.. ." -Tr.10, 963 (Broughton). Instead, Licensee relies on the " initial design in the licensing of the plant, that is, the analysis provided in the FSAR," to ensure that transients not covered by ATOG procedures will not degenerate into a dangerous condition. Tr.10, 967 (Brcughton).

167. The ATOG procedures are based on events which are likely to occur l

during the life of the plant. Tr.10, 943-44 (Broughton). This includes l

i j potential steam generator tube leaks and ruptures. M.at10,944-45.

A steam generator tube rupture is an abnormal transient which is expected to occur between once in 10 to once in 100 years, i.e. appwximately once in the lifetime of the plant. Tr.11, 075-76 (Iavy).

168. Multiple failure events are not considered in AIOG. Thus, for exa:ple, steam generator tube ruptures involving mre than one tube break were not considered. Tr. 10, 972 (Broughton). This reasoning, i

apparently, was hr,e.d on the history of steam generator tube leaks and tube ruptures, rather than an analysis of the actual probability of l

l occurrence of this event. M.at10,974-75. Licensee intends to address these events in future ATOG programs. Tr. 11, 001 (Broughton).

169. Licensee also has not made sure that there are other events outside the design basis that are not covered by ATOG. Tr.10, 970 (Broughton).

Licensee feels that Reccrmendation 2.1.9.C is bounded by the design

I basis of the plant. Id. at 10, 971. Licensee's witness testified that ,

steam generator tube rupture was considered in the FSAR. Tr. 10, 975 (Broughton) .

g 170.1he Board finds that a nore exhaustive Staff review of ATOG is i necessary to determine that Licensee has made reasonable progress toward l

cmpliance wich Reccumendation 2.1.9.C. This review should include the

following areas

--

1. Is the syg tematic ATOG approach apprcpriate and desirable l for 1MI-17 he is, has BW or Licensee adequately deuranstrated that the synptomatic approach will not result in inappropriate operator actions for specific transients?

. 2. Is reactor trip the appropriate event to use to trigger use of ATOG procedures?

3. The additional transient analysis conducted by Licensee should be reviewed in detail. Particular attention should be given to analyses of single and ::ultiple steam generator tube ruptures.

! 4. Proposed cancellations of and modifications of existing procedures should be scrutinized to determine whether inportant operator instructions involving specific events are not deleted.

5. The adequacy of Licensee's pre-inplementation " walk-through" program should be evaluated. The workability of the new procedures is critical to their effectiveness.

6. Further delays in Licensee's coupliance schedule should be reported to the en=f ssi'n, particularly if the NCREG-0737 i daadWaa may not be met.

i 7. Is the scope of ATOG adequate to meet the intent of R+1-. sgiation 2.1.9.C? In particular, the review should focus on 1 the acceptability of eliminating ATWS events and deferring cultiple failure events such as nultiple steam generator tube ruptures.

h results of the reviews of items 1, 2, and 7 should be certified to the Cctmission prior to restart, since these items are fundamental to Licensee's ccepliance with Rey- adation 2.1.9.C. Items 3-6 need only l be cocpleted prior to inplementation of ATOG.

Respectfully submitted,

& c ROBERT W. ADLER /

Attorney for the Cn r==m1th

. ,, n UNITED STATES OF AMERICA NUCEAR NEGli.AM COEISSION BEEDRE 'DE AMC SAFETY AND LICEZEI1U BOARD In the Matter of )

)

Mm0POLITAN EDISON NANY, )

) Docket No. 50-289 (Three Mile Island Nuclear ) (Restart)

Station, thit Ib.1) )

CERTIFICATE OF SERVICE I hereby certify that copies of the attached "Cccroruealth of Pennsylvania's Proposed Findings of Fact and Conclusions of law on I

Plant Design and Modification Issues (First Set)" are served on the parties on the attached service list this 4th day of June, 1981.

Parties were served either by hand or by deposit in the U.S. mail, first class postage prepaid.

. Y?. Y' -

/ n ROBERT W. ADIIR 't c.- -,m . . , . - - - . . , , - . . - - - . - ' . , - -rw . , - - - - =,. - - , , - - -

• , o UNIUD STAU: CF R E CA NIIIAR RELu"."Utf C2tcSSICti BD' CPI FI A*.t2cc SAFE"Y MD LICE:SI'C 3 CARD In de Matte of )

)

}EIRCF.:I.ITRI EDIS31 C2'JKTf, ) .

) Ibcket !b. 50-2S9

("J:ree Pile Islard !Lelear ) (Rastart)

Station, Lhic 26.1) )

SERVICE 1,IST Coorge T. Trabridge, Esq2 ire Dr. Linda W. Little Shaw, Pitt: nan, Potts & Trowtridge Ammta Safety ani Licensing 3osrd Par 41 1800 M Street, N.W. 5000 Farmitage Drive Wshirsten, D.C. 20006 Raleigh, beth Carolina 27612 Ms. l'arjerte M. Aacodt Docketing and Serrice Secticn R.D. 45 office of the Secretarf Cestesville, Perrsfivania 19320 U.S. :Lclear Pagdaterf C',-itsien Ms. Holly S. Keck,14g. Cairman Anti-:Lelear Group Pepresenting Ellyn R. Weiss York (N CRY)

Shelden, Ear:en, Poiscan & Weiss 245 W. Philadelphia Street 1725 I Seeet, N.W., Suite 506 W shingten. D.C. 200C6 Ycrk Perrs/ vania L 17404 Ms. Frieda Be= / hill, Omi - an Karin P. Shelden. Esqaire (PR I)

Coalition fer Lelear Power Shelden, Marten, Sciscan & Weiss Plant Postper. ament 1725 I Street, N.W., Suite SC6 2610 Granden Drive W shingt n, D.C. 20CC6 Wi=ingt:n, Cels a: e 19808 Jess A. Tourtellotte, Esqdre L. Rcbet Q. Pollard Office of ce Executive Legal Director 609 1cntpelic Sceet U.S. Lelear Pagdat:rf c,,-itsien Bal-J::cre, Parf lard 21218 W shingten. D.C. 20555 Alte W. C&mt, Esquire John A. IAVin Esquire C:r.sumer Ad.tx: ace Assistant Counsel Depar=snt of Justice Pennsylvania Pablic Utility CMesien Straster:y Square,14th Floor P.O. Sex 3265 Harrisburg, Perrsflvania 17127 Marrisburg, Perrs/1vania 17120 Dr. Catreey Kapferd Pobert L. Ynsp Esedre Judith H. Jch=srad Assistant Solici=r, Cctr.rf of Dauphin Ir:vir=renntal Coalitten en lLelear P.O. Ecx P 407 !Ierth Fr:nc S= eet Pcwer Farrisburg, Perrs/ vania l 171C8 433 Crlando Aver:ue John E. .* r.ich, 041=:an State College, Perrs/tvarda 16801 l Dauphin Ca.r.cy 3 card of c,-r"<sixers Mr. Steven C. Sholly Dauphin Counef C ar touse l

thicn of C=ccerrad Scientists Front and 'Arv.at Streets 1725 I S=eet, N.W. , Suite 601 Harrisburg, Perrsfivania 17101 Wahingt n, D.C. 20006 Jcrdan D. Cunnfr.ghsc:. Esgaire Ms. Iouise 3radford At=res/ for :e. terry "deship j

"MI Alert T.M.I. Stacing C mf tree l 315 Peffe Street 2320 lbeth Secord S=eet Kar-isburg, Perrsf1vania 17102 Farrisburg, Ferrsfivania 17110 Ivan W. Sc:ith, Esqdre, Cat-an Parrin I. I4wis Atele Safety ard licensing Soard Panel 6504 Bradferd Terrace U.S. bclear Reguisterf Cxr:issien Pri'aMphia, Perreylvania 19149 Washington D.C. 20555 Jara lae R.D. 3, Sex 3521 i Dr. W lter H. Jordan 1 Atxic Saferf ani Licensirg 3 car:1 Panel Ec:cs, Pe:rsf vania i 17319

$81 West Guter Drive Cak Ridge. Terrassee 37830 ~hecas J. Cc=ine. Essaire Deputy Attorney Gearal. Divisi:n of te Pocx:: 316,1100 Ra:ex:nd Deulevar:1 t

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