ML20206J485

Informs That R Strome Will Be Contact for No Significant Hazards Purposes & Will Receive Copies of Plant Lers.Final No Significant Hazards Rule Encl.W/O Stated Encl.Related Documentation Encl

ML20206J485
Person / Time
Site:	Seabrook
Issue date:	10/27/1986
From:	Stello V NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
To:	Sununu J NEW HAMPSHIRE, STATE OF
Shared Package
ML20205K769	List: ML20205K763 ML20206J224 ML20206J246 ML20206J276 ML20206J291 ML20206J329 ML20206J395 ML20206J464 ML20206J485
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FOIA-87-7 NUDOCS 8704160106
Download: ML20206J485 (43)

Text

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- The Honorable John H. Suni:nu Governor of New Hampshire Concord, New Hampshire 03301 j

Dear Governor Sununu:

When nuclear reactor licensees apply to NRC for a change of license, NRC consults with the State as to our assessment that there is "no significant hazards" consideration. This consultation is pursuant to Pub. L.97-415.

With the issuance of an operating license for Seabrook 1, I am writing to inform you that Richard Strome, your designated State Liaison Officer to NRC, 1

will be our contact for no significant hazards purposes. We will also send copies of NRC licensee event reports for Seabrook 1 to h'im.

I am enclosing for your information a copy of the final No Significant Hazards rule, effective May 5, 1986. Your attention is invited to pp. 7765-67, 650.91.

ill Sincerely.

Original signed by, Victor Stelley Victor Stello, Jr.

Executive Director for Operations l

Enc'osure:

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Dear Governor nunu:

When nuclear reactor licensees apply to NRC for a change of license, NRC consults with the State as to our assessment Ahat there is "no significant I hazards" consideration. This consultation is pursuant to Pub. L.97-415.

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Ir. : .ticipati:n of the'1ssuance of an op ' ting license for Seabrook 1, I am writing to infonn you tha(Richard Str , your designated State Liaison

$14 Officer to NRC, will be our contact s for, no significant hazards purposes. We will also send copies of NRC}icensee Atvent reports for Seabrook 1 to him.

I am enclosing for your infonn'at; ion copy of the final No Significant Hazards r rule, effectiv, May 5, 1986. Your/ attention is invited to pp. 7765-67, 650.91. /'

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COMMITTEE ON ENERGY AND COMMERCF c a,37 4 WASHINGTON, DC 20515 November 14, 1986 MEMORANDUM TO: Edward J. Markey, Chairman FROM: Subcommittee Staff

SUBJECT:

Emergency Planning and NRC Licensing Issues at the Seabrook Nuclear Power Plant This memorandum responds to your request for a summary of the Subcommitte,e'4 investigation of emergency planning issues related to the Seabrook nuclear power plant. We stress that both the chronology of events and full explication of them should be regarded as incomplete, since the Subcommittee has not received all materials requested and still is continuing its investigation.

t As you know, the Nuclear Regulatory Commission (NRC) has granted a fuel-load license to Public Service of New ' Hampshire for the Seabrook plant. The licensing board currently is considering the question of issuance of a low-power operating license for the facility. Decisions regarding a full-power license will not be made for some time; however, they are complicated by the f act that the Commonwealth of Massachusetts and thirteen local communities have refused to participate in the emergency planning process in

i. light of their conviction that no practicable emergency planning i is possible which could adequately protect the health and safety

• of their citizens. Under the NRC's regulations, certain findings regarding the adequacy of emergency planning must be made before a full-power operating license can be issued (see below) .

The utility, in close cooperation with the NRC staff, has been conducting technical studies with the objective of demonstrating that the Seabrook containment is so unusually strong that a reduction in the size of the " emergency planning zone" (EPZ) around the plant is warranted.

It has become apparent from public and internal agency

' documents as well as f rom comments offered by the utility and NRC staff that e o a key objective of Public Service of New Hampshire (PSNH) technical studies is to reduce the size of the emergency planning zone for Seabrook f rom 10 miles to 2 miles, thereby excluding the Commonwealth of Massachusetts from emergency planning decisions required for a full-power operating license; pz/) 4?- 7 kW

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2 o technical work undertaken to rationalize this decision was initiated specifically to circumvent the possibility that Massachusetts might refuse to participate in emergency planning on the grounds that it was impracticable for the Seabrook site; o NRC staff appears to have become an adjunct to PSNH efforts and is providing valuable technical assistance to assist the company in seeking its objective.

Investigation by Subcommittee Staff suggests that this effort to reduce the size of the emergency planning zone is questionable for the following reasons.

(1) The NRC staff has held a long series of meetings with the

. utility and its consultants, most of them private or without meaningful public notice, that have served to guide the applicant in its attempt to bolster its technical arguments important for seeking an exemption f rom the emergency planning regulations in order to to reduce the size of the EPI. It appears that the NRC staff may have crossed the line from technical comment to advocacy in assisting this applicant in moving toward its objective.

y (2) On January,1986 NRC Staff rejected a request f rom

[ Baltimore Gas & Electric Co. for an exemption f rom the regulations to reduce the size of the EPZ at the Calvert Cliffs plant, specifically favoring a generic rulemaking over a " piecemeal, site-specific approach." NRC has provided no evidence to justify why that same policy is not being followed ten months later for Seabrook.

(3) Recent NRC statements about unresolved safety issues and containment analysis techniques, some in response to inquiries by this subcommittee, contradict the utility's claims that the Seabrook containment is so strong that it is virtually impossible that a severe accident with significant off-site consequences more

,than 2 miles from the plant could occur. A close reading of NRC documents suggests that the uncertainties surrounding containment failure estimates, accident phenomena and sequences, and source terms are so great that any efforts to use such claims to support a reduction in the EPZ appear to strain credulity.

(4) A cursory review of the technical issues suggests that the applicant's technical arguments represent only one subset of numerous possible scenarios. Many technical uncertainties suggest that other, equally tenable assumptions are equally reasonable.

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( 3 In summary, this effort to reduce the size of the emergency planning zone appears to be a bold attempt by the utility to circumvent the Commission's emergency planning regulations by creating the subterfuge of a purported invincible technological barrier to disaster, thereby justifying an exemption f rom emergency planning regulations. The objective appears to be to attempt to expedite the securing of a full-power license for the facility, at the expense of the citizens of Massachusetts and their rightful participation in the licensing process. To proceed in this f ashion would ignore existing emergency planning regulations which are not being met, and would cast aside the belief by the Commonwealth of Massachusetta and thirteen local communities that no conceivable emergency planning could adequately protect the health and safety of their citizens. The NRC staff appears to have had few misgivings about whether or not their participation in these discussions dignify this effort, and seem immune to considerations of whether their efforts are defensible in light of the serious public and legal controversies surrounding the emergency planning issues in particular and the licensing of the plant generally.

We provide belows o a brief history of the development and application of

! . existing emergency planning regulations as they pertain to the

! { reactor licensing process; o a history of PSNH's and the NRC staff's efforts to create a technical rationalization for reducing the size of the EP2;

) o NRC staff's comments rejecting an earlier application to j

reduce the size of the EPZ at the Calvert Cliffs nuclear plant; o a brief enumeration of the substantial technical-uncertainties, many admitted by the NRC, that plague the issues of severe accident analysis, containment performance, and related matters, and several illustrative comments regarding the need to

- plan for severe accidents; and o a summary of technical issues that PSNH and the NRC staff may have ignored in their analysis of the updated Seabrook probabilistic risk assessment.

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.s REGULATORY HISTORY OF EMERGENCY PLANNING REQUIREMENTS Prior to the accident at Three Mile Island (TMI), the NRC required limited emergency planning in what was then known as the

" low population zone" -- an area immediately surrounding reactor sites that extended only a f ew miles. The TMI accident, during which NRC commissioners had discussed the possibility of evacuating around the stricken TMI reactor out to a distance of 20 miles, forced a reexamination of emergency planning requirements.

In direct response to TMI, the Congress passed legislation 4

(P.L.96-295) that required specific findings on the adequacy of State, local, and utility emergency planning before the NRC could issue an operating license for a nuclear power plant. These i requirements and others were added to the NRC regulations as

Appendix E to Part 50 of the Commission's regulations.

The new NRC regulations were augmented with a new set of companion regulations for the Federal Emergency Management Agency.

FEMA's responsibilities were stipulated in 1983 in 44 CPR 350. In addition, Memoranda of Understanding between FEMA and NRC clarified the' roles of each agency in the process, and in 1980 the two agencies issued joint criteria for emergency planning (NUREG-0654). The latter document was based on an earlier joint NRC/ EPA document, NUREG-0396, which provided a rationale for the l t size of the area in which emergency planning should occur and the response times considered adequate. It is NUREG-0396 that set the rough guidelines of 10 miles as the appropriate size for the plume emergency planning zone and 50 miles for the ingestion pathway.

These guidelines were based on a combination of considerations, I including estimates of accident probabilities and severities, off-site consequences, and public concern. They drew heavily on the Reactor Safety Study (WASH-1400) and EPA estimates of the hazards of given radiation doses.

Under the new arrangements, FEMA was granted the lead responsibility for assessing the adequacy of State and local

- emergency plans and the general level of off-site emergency preparedness. NIK: retained its lead role in reviewing on-site emergency planning, and was charged with reviewing FEMA's findings on off-site planning in making its determinations regarding whether or not the requirements for issuance of an operating license have been fulfilled.

However, neither the NRC nor FEMA can compel a State or local government to participate in emergency planning if that government declines to participate. FEMA's regulations under 44 CPR 350 specify that before FEMA can approve a State plan, the State and

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affected local governments must conduct a joint exercise of emergency plans in cooperation with the NRC licensee; and, any deficiencies in the planning identified by FEMA must be corrected before FEMA will grant approval. If a State and/or local government (s) refuse to submit plans to FEMA for review and/or

]' refuse to participate in the joint exercise, it is an open question whether FEMA, and ultimately the NRC, can make the i requisite findings.

This is precisely the situation at the Shoreham plant on Long Island, where the State of New York and the County of Suffolk have refused to participate in emerge.ncy planning in light of their convictions that there is no way they can adequately protect their citizens in the event of a serious reactor accident. In July 1986, the NRC issued an interlocutory decision which presumed state and local government cooperation in the event of an emergency. Emergency planning issues are currently before the NRC

Licensing Boards in the Shoreham case. No full power license has been issued by the NRC.

, The situation is muddied somewhat by the fact that in P.L.96-295, Congress directed that the NRC could issue an operating license only if there were adequate State and local emergency response plans approved by FEMA, or in the absence of such plans "there exists a State, local, 21 utility ElAD (emphasis added)

I which provides reasonable assurance that public health and safety is not endangered" by operation of the plant. However, given the complexity of emergency planning, the requirement for coordinated actions, and the delicate questions of usurpation of government i authority, to permit a utility plan to replace the required State i and local plans in jurisdictions where State and local governments have refused to participate would raise many serious legal and practical questions. These issues doubtless will be litigated in the courts.

. PSNH EFFORTS TO CIRCUMVENT THE TEN MILE EMERGENCY PLANNING

. REGULATION AND NRC'S ROLE IN THAT PROCESS l!

The Subcommittee investigation has established that PSNH initiated its " update" of the Seabrook Probabilistic Safety Assessment in order to determine if there was a defensible l technical argument for seeking to reduce the emergency planning I

zone around the Seabrook nuclear power plant. If the NRC should concur and take formal action permitting PSNH to reduce the size

of the zone in which emergency planning is required f rom 10 miles to 2 miles or less, the effect will be to exclude Massachusetts from the emergency planning process.

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, The utility and its consultant have held a series of discussions with the staff of the Nuclear Regulatory Commission

) regarding the scope, substance, and direction of these studies.

i Legitimate questions may arise in the minds of the public -

about whether it serves the public interest to have the NRC staff /

deeply involved in precisely that work product upon which it must

• provide recommendations to the Commission. When cooperation becomes too close, the interests of the regulated and the regulators blur, and the public understandably may question whether or not the NRC staff (and thus the Commission) is rendering an impartial judgement' on matters brought before the

Commission for a regulatory ruling -- even when the f acts

, underlying such suspicions may be innocent. This issue of

! credibility is particularly important in contested or -

controversial regulatory decisions, where the integrity of the regulatory process must be maintained without even a hint of a lack of impartiality if that process is to continue to benefit from the public trust.

i In the cise of Seabrook, it becomes disturbingly difficult to discern the licensee's objectives and work f rom those of the NRC i staff, given the close cooperation of the two groups over the l . . entire period of these studies. We have uncovered troubling i

( evidence of NRC staf f's guidance of the licensee's work -- indeed, cooperation in setting its objectives and stipulating its intent.

The staff's disregard for appearances may even have engendered suspicion, at times, regarding regulatory practices which may be reasonable and customary. This is particularly unfortunate, especially in light of the controversy surrounding Seabrook and the Seabrook licensing process.

! We provide below a chronology and accompanying account of the history of the relevant events. We again caution you that there may be omissions, insof ar as we have yet to receive full responses to our inquiries.

SIGNIFICANT ISEE ACTIVITIES In December,1983 Public Service of New Hampshire (PSNH) completed its Probabilistic Safety Assessment (PSA) for the Seabrook plant. PSAs focus on events and analyses that present major sources of risk to the public. They are analytical tools used to increase overall understanding of risks, failure probabilities, and possible weak links in reactor systems and components. In 1984, a review of the Seabrook PSA by the Lawrence s

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. i Livermore Laboratory was initiated, but was halted before completion because the utility was experiencing financial difficulties. According to a PSNH chronology supplied in response to a Subcommittee data request, internal discussions of a new update to the PSA began in March 1985, and contractor work to

" update" the PSA was initiated shortly thereaf ter.

Public Service of New Hampshire commissioned the consulting firm of Pickard Lowe and Garrick (PL&G) -- who had done the work on the original PSA -- to do the work on the update. Comments by PSNH officials have indicated that the utility initiated the update so that additional options would be available to the company in the event problems developed with Massachusetts over emergency planning. William Derrickson, Senior Vice-President of

! New Hampshire Yankee (a division of PSNH) , stated before a j' Subcommittee of the the Advisory Committee on Reactor Safeguards (ACRS) on September 26, 1986:

"We started this effort some time ago because we had been j getting indications from the State of Massachusetts that in fact j what happened last Saturday might happen [i.e., the decision by Governor Dukak'is to refuse to submit emergency plans]....I am not

surc whether or not we would have done it anyway."

In September of 1985 PL&G finished a draf t of the "Seabrook l { Station Risk Management and Emergency Planning Study," the i document that constitutes the technical basis of the PSA update.

The document was circulated for peer review in September, and the first peer review meeting took place in October. The final version of the study was completed in December,1985.

In April,1986 PSNH released a companion study entitled "Seabrook Station Emergency Planning Sensitivity Study." This document sought to demonstrate that the Seabrook containment was so strong that the objectives of federal emergency planning regulations, as embodied in NUREG-0396, could be met even if the emergency planning zone around Seabrook were reduced to a size of

.2 miles or less.

PSNH submitted the Seabrook Station Risk Management and Emergency Planning Study and the Emergency Planning Sensitivity Study to the NRC for review on July 21, 1986. The NRC refers to the two documents together as the "Seabrook Station Probabilistic Safety Assessment Update" -- the "SSPSA Update." In these documents taken together, PSNH contends that an emergency planning zone of a 2 mile or even 1 mile radius may be justified for Seabrook. Since current regulations require a 10-mile emergency planning zone around nuclear power plants, NRC would have to grant i

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,, 8 PSNH an exemption f rom the regulations in order to permit a reduction of the Seabrook EPz to less than 10 miles (absent some other avenue of regulatory relief).

REC STAFF RESPONSE The Subcommittee investigation has uncovered troubling evidence that the NRC staff was involved in the process early on, '

that their contributions have been substantial, and that these early contributions may have helped shape the very documents which were then submitted to them for formal review. Moreover, comments made as part of the review process subsequent to submittal to the NRC appear to indicate a lack of impartiality -- in essence, a tendency to appear as though the NRC staff was working in tandem with the utility to accomplish the utility's objective.

In order to understand the flow of events, it will be helpful to review these events in the context of a rough chronology of NRC participation. A fuller chronology, though perhaps still incomplete, is, appended to this memorandum. The chronology has been reconstructed by Subcommittee staff from documents provided in response to subcommittee requests and f rom interviews with the NRC staff and others. The documents from which the chronology is

, derived are also attached.

(

CHRONOLOGY .QE EVENTS Although PSNH formally submitted the SSPSA Update to the NRC in July, 1986, the Subcommittee investigation has revealed evidence that strongly suggests that the NRC was involved in the development of the SSPSA Update -- more specifically, the Seabrook Station Risk Management and Emergency Planning Study (RMEPS) -- at least one year prior to the actual submittal by the utility. As noted above, PSNH has informed the subcommittee staff that it held its first internal discussions regarding an update to the Seabrook '

.PSA in March, 1985. However, NRC and utility documents indicate v]$)c.

f that in July,1985 at least one member of the NRC met with the utility to discuss the SSPSA submittal and "the regulatory process including the reassessment of emergency preparedness r equi rement s. " This NRC staff member has been deeply involved in proposals to reduce the size of the EPZ generically. To date, neither the NRC staff nor the utility has provided to the Subcommittee full details of this meeting.

In September and October 1985, the NRC and PSNH legal counsel met and held telephone conversations pertaining to a possible

9  ;

i rulemaking petition and the consideration of reducing or eliminating the 10-mile emergency planning zone at seabrook. The rationale used in these early discussions for justifying changes to the emergency planning process was the application of new source term research -- Dal the allegedly unusual strength of the containment. (In later discussions, NRC staff sugggat,ed to PSNH 7 that they not use new source term assumptions in arguing for an EP3 reduction.)

Further NRC-PSNH discussions in an October,1985 conference conference call indicate that NRC staff had become involved in discussing with the utility the appropriate stategy for developing the RMEPS. The NRC staff was concerned that:

"the approach Seabrook was taking on demonstrating the ability to reduce the EPE appeared to be different f rom what [NRC staff]

j understood in an earlier discussion."

Further still, NRC staff noted that:

"...'What he [NRC staff] explained to PSNH in previous discussion was that the approach Seabrook will need to take is to compare the risk

.. of Seabrook to the risk of typical reactor

(, (WASH-1400) used as a basis for the regulation.

[He] suggested Seabrook review NUREG-0396 --- in 1 the comparison it would be good to compare

feature by feature."

[See attached NRC staff notes]

l

In these discussions, it appears that NRC staff was actually i rejecting one approach and suggesting another that the utility y 4

should take to, in effect, justify circumventing a regulation. .

This series of comments, made in October, Ayu5, strongly suggest that the NRC staff to some degree was directly involved in the

, actual development of the basic thrust of the RMEPS, which was not published in final form until December, 1985.

According to PSNH and the NRC, there was little contact between the two regarding the subject of the SSPSA Update between i October,1985 and June,1986, save for one telephone call in February between William Derrickson and Harold Denton, Director of NRC's division of Nuclear Reactor Regulation.

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"* 10 t' The head of the NRC division charged with reviewing the SSPSA Update did inform the Subcommittee staff that he recalled Executive Director for Operations, Victor Stello, advising the i utility not to use new source term assumptions in its submittal.

This suggestion was offered prior to the completion and submittal

, of the documents to NRC. Moreover, between June 20, 1986 and July 15, 1986 there were seven meetings and telephone calls between Derrickson and several high NRC staff officials.

PSNH submitted the Seabrook Station Risk Management and

, Emergency Planning Study and the Emergency Planning Sensitivity i Study to the NRC for review on July 21, 1986. In these documents, PSNH contends that an emergency planning zone of a 2-mile or even 1-mile radius may be justified for Seabrook. Since current regulations provide for a 10-mile emergency planning zone around nuclear power plants, NRC would have to grant l'SNU an exemption f rom the regulations in order to reduce the EPZ to less than 10 miles.

On July 29, 1986, PSNH requested that the NRC expedite the review as fo11,ows:

l "A future submittal, deoendina en tha results

' gi ih2 technical review. may request a change 7 to the emergency response plan process for ,

1 ( Seabrook Station. We cannot, at this time, i'

specify what action such a future request might seek, but it is important that we

! address as soon as possible what ootions are

! available to us relative to full power i licensing. This is important in light of the

! apparent strategy of the State of l Massachusetts to delay the process."

i (emphasis added)

Subcommittee interviews with NRC staf f indicate NRC's complete understanding that at the outset, PSNH solicited and

. secured the benefit of the technical expertise of the NRC to

! ensure that if and when the company seeks an exemption, the l Commission will be predisposed to grant it, because the NRC staff

[ already will have approved the technical basis for that exemption.

l A July 29, 1986 staff note indicates, "Derrickson wants to know if this could serve as a technical argument. If not, he won't file.

Pointout which technical arguments are good vs. no." NRC staff also acknowledged that they were and continue to be well aware that if the study's conclusion is upheld by the Commission, the utility probably will be able to remove itself from its apparent stalemate with Massachusetts over emergency planning and will be in a position to obtain seabrook's operating license.

I

. i 11 NRC mobilized very quickly to coordinate this review. Within 15 days of receiving the study, the Commission had contracted with the Brookhaven National Laboratories to assist in the review. By August 6, Brookhaven had prepared a detailed project description for the review. Brookhaven's list of review milestones outlined a three month undertaking at a cost to the NRC of $245,000. NRC staff concedes that this short time frame is "out of the ordinary."

On September 26, 1986, PSNH and the NRC staff presented i reports on the study and its review to a subcommittee of the Advisory Committee on Reactor Safeguards (ACRS), and on October 10, 1986, reported to the full ACRS Committee. The ACRS Committee is composed of scientists who advise the NRC on potential hazards i

' of proposed or existing reactor f acilities and on the adequacy of proposed safety standards. ACRS has been reviewing the study separately from the NRC review, and will provide to NRC in early

1987 an independent recommendation on the merits of the study, i

At the . September meeting, the utility reiterated its desire to increase their range of options relative to emergency planning.

Mr. Derrickson related the utility's "need to know the conclusion of the NRC as to our results so that we can move forward. We I

/ 'q really can't move forward until we know we have some level of j agreement."

The Subcommittee staff has serious concerns about the NRC staff's approach to the review, and specifically how they perceive j their role in this process. Although NRC contends that what they 1 have undertaken is strictly a review of the technical merits of

PSNH's study, NRC internal documents and staff notes f rom meetings and discussions with the utility and its consultants indicate that members of the NRC staff f requently speak as advocates rather than objective regulators. NRC staff of ten sound as if they are trying i to help the utility put the best f ace on the conclusion of the

l. study.

i For example, in a July 25, 1986 staff meeting, almost immediately af ter the NRC had received the study, one NRC staff person noted, "what do g3 (emphasis added) have to justify to change EPZ? ... need to consider technical and legal" (emphasis in original). The tenor of these comments seems to contradict what the NRC has told the Subcommittee about the review. It is important to recognize here that the NRC has not been asked yet to i

address the question of a change in the emergency planning zone,

and has told us specifically that it is not considering it.

i Moreover, even if PSNH had requested an exemption f rom the l

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! emergency planning regulations, NRC staff, as employees of the

! pertinent reaulatory agency, would not be considering what the HBg i

had to justify to change the EPI. In such a case, the burden of proof would be on the utility.

NRC's regulatory mandate to oversee the nuclear industry runs contrary to this approach. NRC staff has told the Subcommittee that NRC made the determination that this review was appropriate, 1 in the context they outlined for us, because it is an " objective" review of the studies' technical merits only. By setting out to support the studies, NRC has abdicated its regulatory responsibilities. An NRC staff note f rom an August 27, 1986

!. meeting with PSNH staff further illustrates the cooperative nature l of the NRC/PSNH approach to this review: " Review aroup fillegible] Novak, coordinated glih th3 utility, with a list of i goals... We need to think about what this group can do 'in three months" (emphasis added). [Novak is the acting Director of the j Division with primary responsibility for the review.]

Moreover, NRC staff meeting notes and the attached chronology of meetings between NRC and the utility appear to suggest NRC staff accept'ance of the political and legal objective of the study to "get rid of Massachusetts so they don't have to submit" emergency plans for the affected communities. In another meeting, i

/l "'3 an NRC staff representative wrote: " Uni Seabrook containment - Let's [NRC staff]que try features of the to make it more l unique - show it's better than average." Through these comments and many others in which the staff seems to be trying to help the utility to put the conclusions of the study in the best possible j light, it appears that the NRC staff well may have surrendered the

" impartial" professional attitude befitting the staff of a

regulatory agency.

. The NRC expects a preliminary Brookhaven report on the review 1

by the last week in November. Presumably, if Brookhaven

indentifies inadequacies in the study, the utility may correct

! those inadequacies and resubmit the study for.a renewed

! NRC/Brookhaven review. NRC staff has informed the Subcommittee

! staff that as long as the utility wants to pursue NRC approval of l the study, this process of evaluation and reevaluation could

( continue indefinitely. We believe that should such a process evolve, the NRC essentially will be a full partner in the l development of the study -- if it is not already, i

The conduct of the NRC staff with respect to the Seabrook studies pertinent to emergency planning raises numerous troubling l

questions for consideration by the Subcommittee.

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• I o Is it proper for NRC staff to conduct a review of this i

study in the absence of a formal petition or exemption request by PSNH7 o When is early involvement by NRC staff in a licensee study associated with a clear regulatory intent inappropriate? How much 1 involvement is appropriate? What guidelines should the NRC staff use in making these decisions?

o Has the NRC blurred, if not erased, the line between advocacy and technical comment in the licensing process? When the Subcommittee staff asked the NRC. staff to draw a distinction between the goals of the utility and the NRC in this review, i especially as supplemental information is exchanged during the review, the NRC was hard pressed to draw such a distinction.

l i We believe that the Subcommittee might usefully question the

! propriety of an NRC review of a licensee study which the staff l understands may be used by the utility as the technical basis for

, a future request for an exemption f rom NRC regulations. PSNH has i

{

solicited and, secured the benefit of the technical expertise of the NRC to ensure that if and when the company seeks an exemption,

, the Commission staff will support granting it, because that same I NRC staff already will have approved the technical arguments

~- supplied as the basis for that exemption. If PSNH pursues such an

I exemption, the same NRC staff will be called upon by the Atomic Safety and Licensing Board and/or the Commissioners to provide l their views.

The NRC is charged not only with the licensing of nuclear power plants. More importantly, it bears the responsibility of protecting the public f rom the health and safety hazards of nuclear power. The public has every right to expect that the Commission will discharge its duties impartially. If the integrity of the regulatory process is to be maintained, the Commission and its staff must take particular care, especially in contested licensing proceedings, to avoid even the appearance of

. favoratism, promotion, or partiality. The staff must abide by a standard of conduct which avoids planting seeds of doubt in the public's mind that the regulatory process has been biased. In effect, the public regards the NRC as judge and jury. When the staff acts in such a way as to engender doubts akut the Commission's objectivity, then the public may perceive that the Commission is stacking the jury in favor of the nuclear industry.

To permit even such an appearance undermines the credibility of ,

the regulatory process.

s

, 14 Especially in the case of Seabrook, where the public confidence in both the NRC and PSNH is strained at best, the public interest might have been better served by a more open and formal review process. How and where to draw the line is another question the Subcommittee may wish to address.

In short, the NRC appears not to have been sufficiently i

sensitive to the public's concerns about the potential safety implications of reducing the emergency planning zone at Seabrook.

The staff's actions to date threaten to inflame public sentiment rather than to provide the inarguably impartial discourse necessary to calm it.

PAST NRC STAFF COMMENTS ON DEFERRING REQUESTS FOR REDUCING i

THE SIZE OF THE EPE EXCEPT IN GENERIC PROCEEDINGS As noted above, the question of reducing the size of the EPZ had arisen previously, in a request by Baltimore Gas & Electric to seek a reduction in the size of the EPZ for Calvert Cliffs. 'In

that instanc.e,. Robert Minogue, the Director of the Office of Nuclear Regulatory Research, wrote the following in a January 27, t

1986 memorandum to Harold Denton.

l "At the last RES quarterly meeting it was agreed

( that a decision on this matter would be premature l

at this time and that no decision in this area would have a firm foundation until af ter NUREGs

-0956 and -1150 are published. Nevertheless, in reponse to your request we recommend that the requested exemption be denied at this time or that a decision be postponed until a generic rulemaking on the subject is completed in FY 1987 or FY 1988.

Our reasons for this position are multiple."

A few of those reasons are illustrative.

. o "The orderly progression of generic rulemaking on the emergency planning issue will serve the public better than a piecemeal, site-specific approach."

o "The public must be involved in major rule exemption or rulemaking decisions. However, because of the broad policy implications of this action, the public would be better served with a generic rulemaking rather than a site-specific one. "

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i o "RES would prefer an' integrated approach looking at the source terms and risk from all of the six SARRP/NUREG-1150 reference plans to determine what the generic NRC positions should be and what, if any, plant-specific alternatives might exist. To proceed first on a plant-specific basis could preclude a more rational, policy-level look at the problems, which could i result in inadvertent, unfortunate decisions." i The memorandum concluded by saying that "We understand that you have sent a letter to BG&E to the effect that it is premature 4

to consider reducing the EPZ, and we strongly support that action."

A January 6,1986 memorandum to Denton f rom Thomas Murley, NRC's Regional Administrator for Region I, also offered pertinent comments on the BG&E proposal. In that memorandum, Murley wrote:

...The consideration of this exemption request should be approached cautiously. Two impo,rtant questions must be answered before

. this request, and others of this type which may follow, can be considered solely for their logistical and technical merits. The first w question is ... Who determines the size of the l I EPI? ...One may argue that the state and local i governments should play a major role in

. determining who is to be protected and i subsequently in setting the size of the EPE.

The State of California, for example, requires planning for an Expanded Planning Zone

. (20-mile radius)...."

"The most conservative approach would be to consider and resolve the generic questions prior to considering exemptions on a plant-specific basis. This is our [the Region

. I staff's] recommendation."

Murley's comments on the appropriate role of state and local governments in determining the size of the EPS sound somewhat incongruous in light of the staff's present involvement in an attempt to craft a technical argument designed precisely to circumvent state and local government concerns about emergency planning by summarily excluding those governments f rom the EPZ.

The NRC staff position in January 1986 clearly opposed dealing with such requests on a plant-by-plant basis. We see no evidence that the situation 11 months later is substantially different.

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16 PAST NRC STATEMENTS REGARDING REACTOR SAFETY, CONTAINMENT FAILURE, AND ACCIDENT PROBABILITIES CURRENT REQ HQBE SE ELSK REBASELINING AHQ SEVERE ACCIDENTS Af ter the TMI accident the NRC began a comprehensive review

of the risks of a severe accident. According to discussions 1

between NRC staff and Subcommittee staff, this review has taken more than 5 years and has cost more than $250 million. The objective of the review is to attempt to establish with greater precision appropriate source term assumptions for a severe accident and to refine risk assessments so as to improve upon the work of the 1975 Reactor Safety Study (WASH-1400).

According to staff, the draft document resulting from this review, NUREG-1150, is currently scheduled to be released for review and comment early in 1987. Following a 120-180 day review period, staff expects to take an additional 120-180 days to digest comments and make changes to the draft document. While the staff has informed us that the objective includes various rulemaking

/

proceedings, they do not expect those proposed rules to be issued until the final NUREG-1150 is released late in 1987 or early in 1988.

! Many of the issues under review in the NURBG-1150 process are directly relevant to the issues under discussion with regard to l

Seabrook. It is extraordinarily difficult to understand why the

~

NRC staff seems intent upon fine-tuning a study for one plant when a sweeping generic review is drawing to a close. The NRC bears an extremely heavy burden to justify its departure f rom the rationale offered in the Minogue memorandum (above) . The NRC has provided -

no reason why it could not await the promulgation of rules pursuant to the publication of NUREG-1150 before taking any action

• on the size of the Seabrook EPZ that depends upon containment and -

risk analyses for the Seabrook plant. To proceed otherwise directly contradicts positions taken by staff only months earlier.

LARGE UNCERTAINTIES IE SAFETY-RELATED ISSUES In an April 30, 1986 memorandum from Victor Stello, Jr.,

Executive Director for Operations to Commission Secretary Samuel Chilk, Stello identified a long list of safety-related issues that

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

1 17 require further research. Many of them bear directly on issues related to containment performance under severe accident conditions, and therefore on off-site consequences resulting from a severe accident. Although the purpose of Ste11o's memorandum was to argue for additions to the NRC budget necessary to pursue

, research programs in these areas, the issues themselves paint an l

extraordinary picture of the gravity of outstanding safety-related issues and the extreme lack of knowledge on the part of the NRC about severe accidents. Given the number of items enumerated and the breadth of their implications, it is difficult to understand why the Commission staff is intent on expending its time and public funds on the Seabrook technical studies in light of these many extant technical uncertainties. Indeed, the breadth of the safety issues appears so great that we recommend the Subcommittee consider an independent investigation into their implications for operating reactors.

i Stello wrote in his introductory narratives i

"In evaluating the various technical issues that must be resolved to provide a basis for (regulatory] actions, it is becoming apparent that ASEA 91 tha uncertainties maz ha an larae, even with ths knowledae cained from ihm

,' four y3Agg 91 intense focused research in l

date, thal EBfs EAX D91 h3 able in provide 3 l satisfactory reduction 21 these uncertainties mith existina resources.... Some of these issue may be the chemical and physical form of radioactive iodine and cesium and their mutual interaction, the degree of direct containment heating from the expulsion of molten reactor core materials, hydrogen generation and loads and containment performance in resisting these and other loads placed upon it f rom a severe accident." (Emphasis added]

. We provide below excerpts f rom the Stello memorandum.

Safety Issue: Large uncertainties exist in analyzing in-vessel accident behavior at core-melt conditions, including clad oxidations heating, hydrogen generation and l

transport. This in turn leads to large uncertainties in containment heating and penetration calculations which in turn are used in health effects consequence determinations.

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.. 18 Safety Issue Large uncertainties exist in the in-vessel fission-product release rates currently used in Source Term analysis with the Source Term Code Package (STCP) . These rates are taken f rom early, old, relatively coarse, out-of-pile data at atmospheric pressure. The relatively volatile fission products not released f rom the fuel in-vessel and then removed by plate out in the reactor vessel are immediately released ex-vessel by melt-concrete interactions. In cases of early containment failure these fission products can become the largest part of the release to the containment. ,

Safety Issue One of the safety issues is whether or not, in the absence of a hydrogen control system, given hydrogen-steam mixtures in large dry containments, would such mixtures autoignite and burn slowly destroying containment seals or, alternatively, burn rapidly and damage containment equipment and structures. The information would allow NRC r' to evaluate the need for changes in equipment

, l qualification and hydrogen regulations.

Another issue not resolved involves the 4 quantitative loads to containment walls and electrical and mechanical safety equipment from local hydrogen detonations and the probability of wall and equipment damage."

1 Safety Issues Significant uncertainties exist in the safety analysis codes ability to calculate reactor response to feedwater-line/steamline breaks. These j uncertainties affect NRC's ability to evaluate i

. such effects as: 1) thermal shock to the reactor vessel under pressure in the event of a steamline break; and 2) coolant system overpressurization during feedwater line breaks.

Safety Issue: To confirm the adequacy of data obtained f rom plant instrumentation and the availability of electrical equipment under severe accident states which provide the basis for operational and emergency preparedness s

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19 actions; and to provide more accurate data for deterministic and probabilistic calculations as they pertain to severe accident states. i Safety Issue: The requirement of separation of the redundant trains of safe shutdown

equipment prescribed by 10CFR50, App. R cannot be met in many control rooms. The rule requires in such cases that an alternative or dedicated shutdown capability be provided.

Whether in the event of a credible control room fire the operator will have time enough and the physical ability to transfer control of the reactor to the alternative or dedicated

shutdown panel is at issue. There is also the question of how much of the control room i equipment will survive and whether and when control room operation can be resumed. ,

Safety Issues The currently available 1 re.al-time atmospheric dispersion, precipitation washout and plume rise models for dose projections in emergency response

, situations have not been adequately validated, i

Accordingly, there is the potential for an I inappropriate, unreliable or inaccurate model being used in an emergency response situation i and : sic] which would produce incorrect dose

! proj ections. . .. Risk assessments and other

licensing evaluations in which the models are

used have large uncertainties.

Clearly, there are an extraordinary number of open

! safety-related issues, some bearing directly on the Seabrook j analysis. We f ail to understand how the utility and the NRC staff can provide sufficient assurances that the Seabrook containment is

virtually " fail-safe" in light of these considerable uncertainties

! ' remaining in understanding and modeling basic phenomena.

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OTHER COMMENTS DH POSSIBILITIES DE & SEVERE REACTOR ACCIDENT In light of the efforts by PSNH and NRC staff to demonstrate j that the Seabrook containment is virtually "f ail-safe" and that a i miniscule probability exists of an accident with significant I

off-site consequences beyond 2 miles, subcommittee staff felt it useful to review several knowledgeable sources' comments regarding l

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4

20 severe reactor accidents. We found that even a cursory review t suggests that the path being pursued by PSNH and the NRC staff is ill-considered, and that many acknowledged experts -- including some within the commission itself -- have offered comments that reflect poorly on the effort to reduce the size of the EPI. We provide a f ew examples below.

On the need for planning for serious reactor accidents, from the report of the Kemeny Commission: " ...We must not assume that an accident of this or greater seriousness cannot happen again, even if the changes we recommend are made. Therefore, in addition to doing everything to prevent such accidents, we must be fully prepared to minimize the potential impact of such an accident on should one occur in the future." --

public health and safety, Report of the President a Commission on the Accident at Three Mile _

Island, p.15.

On the question of the accuracy of probabilistic risk assessments estimations of core meltdown probabilities: "I believe that the recent Davis-Besse event illustrates that, in the real world, system and component reliabilities can degrade below _

~

those we and the industry routinely assume in estimating core melt f requencies. Our regulatory process should require margins against such degradations and also to reflect the uncertainties in our PRA estimates." -- Harold Denton, NRC Director of Nuclear I Reactor Regulation, in memorandum dated June 27, 1985, to William J. Dircks.

On the possibility of severe reactor accidents: "There is a distinct possibility of one or more additional severe reactor accidents, beyond the one at Three Mile Island, in the remaining life of the plants now in operation or under construction, unless the estimated accident f requency declines sharply with modifications, or has been significantly overestimated in current PRAs and actuarial inf erences." -- NURBG-1070, 'NRC Policy on Future Reactor Designs Decisions on Severe Accident Issues in Nuclear Power Plant Regulations," August 1984, p.108.

On the possibility of an accident in the U.S. as severe as the Chernobyl accident: " ...given the present level of safety being achieved by the operating nuclear power plants in this country, we can expect to see a core meltdown accident within the next 20 years, and it is possible that such an accident could result in of f-site releases of radiation which are as large as, or larger than, the releases estisated to have occurred at Chernobyl.' -- NRC Commissioner James Asselstine, in testimony before this subcommittee, May 22, 1986.

. , e 21 Even these few examples suggest that both the NRC staff and PSNH are proceeding under a dubious assumption when they attempt to demonstrate an invincible technical basis premised on containment strength and severely limited accident scenarios in arguing that the Seabrook plant should be considered essentially immune to severe accidents.

ERLQB EEG COMMENTS 21 CONTAINMENT PERFORMANCE In recent times the Subcommittee has had numerous exchanges with the NRC regarding emergency planning, containment analysis, and source terms. Several statements by the NRC have either directly contradicted the positions now being taken by NRC staff and the PL&G consultants or have emphasized that the present state of the art is characterized by substantial uncertainties in analytical techniques. It appears evident the NRC staff has ignored such qualifications in its attempt to assist the utility to justify a reduction in the size of the EPI.

The utility, its consultants, and PL&G have argued that the containment failure pressure is so high as to make a severe accident implausible. However, in a May,1986 exchange between the Subcommittee and the NRC, the NRC clearly stated that i

uncertainties in containment analysis were too great to permit a t

precise estimate of containment failures. The exchange reads as

follows

QUESTION 5. What degree of confidence does the NRC have in the ability of different I containment buildings to prevent a major release of radiation during a core meltdown?

For each type of containment building, what is the estimated probability of containment failure given a meltdown and state precisely

! what uncertainty bounds are assigned to this l

estimate and how it was calculated?

i ANSWER. At present, the NRC staff cannot

specify with a high degree of preciseness, the conditional probability (and uncertainty bounds to be assigned) of containment f ailure with a major release of radiation....There are very low probability severe accident conditions under which a containment may be j unable to prevent a major release. Although containment structures are conservatively

j. designed to withstand the substantial

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22 temperatures and pressures associated with a

, major pipe rupture...they are not designed to withstand the additional challenges that might be associated with a complete core melt. Such

, challenges include phenomena such as increased

pressures from an uncontrolled hydrogen combustion or release of large quantities of noncondensible gases f rom core-concrete interactions....

The exchange in Question 6 reiterates these inherent uncertainties.

QUESTION 6: In a supplement to the record of the subcommittee's April 17, 1985 hearing, the NRC staff wrote: " Analysis shows that there are some kinds of accident sequences that could cause failure of any containment

design...." For each type of containment please enumerate each mode of potential containment f ailure and the conditions that

ca'n lead to each mode of failure. What is the

, relative likelihood of different modes of containment failure?

ANSWER. The potential containment f ailure modes for severe accidents in all U.S. LWR

, designs can be generally classified into six groups. The definition of these groups depends on the timing of the f ailure (relative to core melting and major releases of radioactive material into the containment) and

the failure location. These groups ares l 1. Early failures directly to the atmospheres

2. Early f ailures into other plant buildings; i

'

• 3. Late f ailures directly to the atmosphere;

4. Late failures into other plant buildings;

5. Late failures into the ground; i 6. No containment failure.

The exact failure modes and causes, and their relative likelihoods, vary considerably among plant types and even among plants of similar containment design....The NRC staff and supporting contractors are presently engaged in a major reassessment of the risks of

( current commercial reactors. More i

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t 7- 23

+ quantitative information on the types of containment f ailures and their relative likelihood and importance will be provided in the staf f repo rt NURBG-1150. . . .

t j For PNR's with "large, dry" and i "subatmospheric" containment designs (like i Seabrook), early failures to the atmosphere have been predicted to result f rom accident phenomena such as direct heating, steam l explosione, and coincidence of hydrogen burns

! and rapid steam generation. For most plants i

with a large dry containment, the likelihood of these events including an early containment f ailure is _quite small.

l The above comments demonstrate that there are no ready

! answers to questions regarding containment f ailure. If the NRC is unable to supply any hard numbers and is reduced to citing analytical uncertainties and studies in progess, it is difficult I to understand how in three short months PL&G could accomplish an analysis sufficiently well-grounded to warrant a change in emergency planning regulations for the Seabrook plant.

POTENTIAL TECHNICAL PROBLEMS WITH THE UTILITY'S CLAIMS I

The utility's consultant has asserted that the pressure at which the containment would f ail is significantly higher than for l

most other similar large dry containment pressurized water reactors. They attribute this greater strength to several j

l factors, including a larger than average containment volume and more steel used in the construction of the containment. The Subcommittee staff questions whether PL&G has f ailed to consider

, adequately a number of significant phenomena and possible reactor accident scenarios, several of which could possibly result in i outcomes very different from those cited by PLEG. A few examples

! of these technical questions may be summarized briefly as follows.

i I (a). The conclusions of the PL&G report depend crucially on the assumption that the containment f ailure pressure is roughly

! 225 psig. This value is much higher than the estimated

! containment f ailure pressures for other plants, and may not be j

adequately supported. In addition, containment f ailures generally j may be assumed to occur at points of structural discontinuity I

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i 24 (e.g., where piping enters or exits the containment) . Even if the shall of the containment is stronger than average, one does not expect a shell rupture to be the first f ailure point of the

containment.

(b). PL&G examined only one containment bypass scenario; however, other bypass scenarios may well be relevant and could produce releases greater than those estimated in the PLEG study.

(c). PL&G has argued that if noble gases are released f rom '

the containment af ter an accident, the containment will hold them long enough to permit decay to levelr low enough to warrant a reduction of the EPI. However, emergency planning as currently

' conceived assumes that off-site dose-response relationships must be examined without any protective actions. It appears that if noble gases are released without a significant delay, under various realistic worst-case weather scenarios, the acceptable

off-site dose levels might well be exceeded at 10 miles.

(d). Assumptions about the radioactive inventory released under high pressure melt ejection scenarios may not be consistent ,

with recent analyses conducted by Sandia National Laboratory on these phenomena.

< '. (e). The report may not adequately consider the prospects I

that accident recovery actions could cause steam de-inerting by condensation, leading to hydrogen burns or detonation within the containment.

(f). In examining earthquakes as a contributor to radioactive release scenarios, the report may not adequately consider af ter-shocks while the containment is pressurized, which adds an impulse pressure load.

(g). The report's conclusions on radioactive releases as a consequence of steam generator tube ruptures may not be consistent

• with recent analyses conducted by Battelle Columbus Laboratory on steam generator tube rupture scenarios.

In summary, it appears that PL&G may have been selective about their assumptions; if one made other, equally reasonable assumptions, the outcome might be much different.

4 1

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- ATTACHMENTS -

NOVEMBER 17, 1986 SUBCOMMITTEE MEMORANDUM RE:

SEABROOK NUCLEAR POWERPLANT

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EMERGENCY PLANNING AND SAFETY ISSUES e

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y ,,

x CHRONOLOGY OF KEY MEETINGS / DISCUSSIONS INVOLVING NRC AND PSNH AND FEMA STAFF RELATIVE TO REDDCED EPZ July 30, 1985 NRC/PSNH meeting -- discussion of SSPSA submittal -

according to NRC this included "the regulatory process including the reassessment of emergency preparedness requirements" September 9,1985 NRC/PSNH staff and counsel conference call --

Discussion of a possible rulemaking petition and idea that a 10-mile EPZ is not necessary at Seabrook -- based on source term research October 10, 1985 PSNH counsel, and NRC legal staff discussion --

" reduce ~or eliminate the entire EPE - Taking Seabrook PRA applying new source term data Discussion that "once technical analysis is done subject to peer review group ... if the peer review group raves, then sometime in Nov.1985 will come in with a package" October 11, 1985 NRC/PSNH staff conference call -- NRC staff notes that feedback about the approach Seabrook was 4 l, taking on demonstrating the ability to reduce the EPZ appeared to be diff erent f rom what he understood in an earlier discussion (with PSNH staff)

" (NRC staf f) noted that what he explained to (PSNH staff) in previous discussion was that the approach Seabrook will need to take is to compare the risk of Seabrook to the risk of typical reactor (WASH 1400) used as a basis for the regulation. (NRC l

staff) suggested Seabrook review NUREG-0396 -- in

.. the comparison it would be good to compare feature

. by feature"

"(NRC staff) said Seabrook should not be developing the case that Seabrook . . . " (incomplete note)

October 22, 1985 Seabrook coordination meeting with FEMA, NHY, NH and MA state civil def ense staffs - Includes discussion of impact of the probabilisitic risk assessment on emergency planning issues FEMA staff notes that Brookhaven is reviewing the containment and time-to-f ailure aspects of the PRA Discussion of the size of the EPZ and legal analysis work Nov ember 26, 1985 Seabrook coordination meeting with FEMA, NHY, NH b

CHRONOLOGY - 2 i

and MA state civil defense staffs - Further l discussion of the PRA and impact on emergency planning issues Dacember 1985 "Seabrook Static 7 Risk Management and Emergency l Planning Study," Pickard, Lowe, and Garrick February 1986 NUREG/CR-4540, "A Review of the Seabrook Station Probabilistic Safety Assessments Containment Failure Modes and Radiological Source Terms,"

Brookhaven National Laboratories April 1986 "Seabrook Station Emergency Planning Sensitivity

Study," Pickard, Lowe, and Garrick July 9, 1986 NRC, PSNH, PL&G meeting to discuss PSA program July 21, 1986 PSNH submits the Risk Management and Emergency Planning Study (RMEPS) and the Seabrook Station Emergency Planning Sensitivity Study to NRC for review July 25, 1986 NRC staff meeting - "(PSNH staff) says - before

,'- they go for it - give them a reading by October --

(- what are the merits of this piece of work -- does it provide some basis to go forth with exemption" NRC staff notes "What do M3 (Emphasis added] have to justify to change EPE? ... need to consider technical and leaal" (emphasis in original)

July 29, 1986 PSNH requests that NRC expedite the technical review of RMEPS as follows:

"A future submittal, dependina 2D tha results 2f thg technical review, may request a change to the -

emergency response plan process for Seabrook

. Station. We cannot, at this time, specify what action such a future request may seek, but it is important that we address as soon as possible Mhat options 313 available in MA relative to full power licensing. This is important in light of the apparent strategy of the State of Massachusetts to delay the process. " (emphasis added)

July 18 or 29, 1986 NRC staf f meeting - " (NHY) wants to know if this could serve as a technical argument. If not, he won't file. Point out which technical arguments are good vs. no"

( "Seabrook has thrown away containment f ailure" "Did they include any real data vs. generic -- May have to go back and tidy up PRA, but probably not"

CHRONOLOGY - 3 August 5, 1986 NRC Staff and PSNH meeting - "NRC is beginning an expedited review of the study to assess the technical adequacy of PSNH's analysis to support the study's conclusions."

August 6, 1986 Brookhaven National Laboratories (BNL) compiles project description for Review of the Emergency Planning Sensitivity Study for Seabrook August 6, 1986 NRC Staff, BNL, and PSNH onsite walk through --

" Objective to reexamine emergency planning basis --

wanted enhanced mathodology for 3113 soecific olannina -- determine risk impact of different options

• August 11, 1986 Internal NRC staff memorandum - "It is important to decide what direction NRC is going to take on this issue before a detailed technical review can start. A decision chart set up in the form of

. . three questions is attached for your consideration" August 13, 1986 NRC staff notes -- PSNH rquest "NRC will give them top priority ... preliminary review in 3

{ , months - normally takes - out of ordinary"

" Sense of urgency - no discussion of it holding up licensing" "There are so many options - ggt Ild 91 Mass so they don't have to submit" (emph. added)

August 14, 1986 Meeting at BNL with NRC staff and PSNH -- NRC staff notes "Do review in short period of time to get positive response or questions needed to get there" -

NRC staff " Unique features of Seabrook containment - Let's try to make it more unique -

. show it's better than average" (emphasis added)

August 27, 1986 NRC staff and PSNH meeting - " Review group to (NRC staff) , coordinated ylih tha utility, with a list of goals ... We need to think about what this group can do in 3 months" "What are possibilities -- 3 mile EPZ with plume?

-- 1 mile EPZ"

" Agenda (What they could do to show they're different to get credit)" [ original ' () ']

" Shrinking of plannina zone vs. evacuation zone ...

may be able to reduce evacuation zone but not plannina zone" (emphasis in original)

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CHRONOLOGY - 4 August 28, 1986 NRC staff memo outlining staff review plan for Seabrook EPZ Sensitivity Study:

" Goals of review:

1. To provide a technical assessment of the adequacy of the Seabrook Station Emergency Planning Sensitivity Study to support its conclusion that the degree of public protection afforded by a 1 mile emergency planning radius around the Seabrook Station is equivalent to the degree of protection that was perceived for a 10 mile emergency planning radius at the time the 10 mile generic planning radius was established in NUREG-0396.

2. In the event it is concluded that the Study does not adequately support its conclusion at the 1 mile radius, to determine the radius at which the study can support a conclusion of equivalent protection."

September 26, 1986 Advisory Committee on Reactor Safeguards (ACRS)

Subcommittee meeting -- PSNH staff comments: " Step

, one was for us to write and submit (the saf ety

" ( assessment). We really need to know the conclusion of the NRC as to our results so that we can move forward. We really can't move forward until we know we have some level of agreement."

PSNH "We started this effort some time ago because we had been getting indications f rom the State of Massachusetts that in f act what happened last Saturday might happen ... I am not sure whether we would have done it anyway. But give7 the f act that we have had some difficulty in Massachusetts prompted us to do this in 1985. If nothing else, it certainly lended a sense of urgency to the situation."

ACRS Member: "Can you go ahead by just agreeing with NRC and FEMA that your plans are okay and Massachusetts can sit up there and suck its thumb all it wants?"

PSNH: ... the exercise in New Hampshire will come possibly af ter the decision in Long Island. So we

< are going to know before we finish with New Hampshire whether or not that path is really i viable.

"We are trying to learn f rom everybody. We want to look at every possible alternative and everything l

we can do so that we increase our range of options in case one option doesn't work."

l

7 .

4 CHRONOLOGY - 5 ACRS Member: "How far is the nearest point in Massachusetts from the plant?"

PSNH: "Two miles 690 feet. "

ACRS Member "If you had an emergency zone of 2 miles 690 feet you would be in the clear?"

PSNH "I think the answer to that question is yes."

October 10, 1986 ACRS full committee meeting ACRS: "The two reports that we have looked at have both been reviewed by an outside, independent group of experts in each case, and these experts concurred with the results of the findings.

However, when we asked in terms of the containment, I questioned why there wasn't a containment expert on the independent review groups, and they pointed out that the staff, NRC staff, through a contract with Brookhaven National Laboratory, is reviewing that aspect. So it will be reviewed."

i ACRS: "I would hope the (NRC) staff develops some kind of safety philosophy, if it hasn't, that it uses to guide itself in decisionmaking concerning these matters. ... I'm speaking for myself but let me strongly suggest that you try to develop some, what I'll call general philosophy in this regard, and then some, what you might call generic quantitative guidelines, before trying to make an ad hoc decision on a specific case."

ACRS: "But let's face it. They wouldn't be .

.. reviewing this PRA at this time if they didn't

. think there was something like a change in the prior guidelines for emergency planning in the wind. Otherwise, I would like them to tell me why they are taking time away f rom other tasks like finding out how good a mark 2 containment is, etc.?"

l l

1

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Ng.c. 9 96 p.espc a se Tb S uac.cHH ITf66 -

QUESTION 8. d) Has the Commission staff ever advised PSNN concerning the studies it would deem appropriate for justifying a reduction of the EPZ? If so, please provide all detafis of such dis-Cussions.

ANSWER.

In the recent past there have been some discussions of risk and emergency prepared-NRC. between management of PSNH and Robert Bernero Director, of 8WR Licensing a ness Prior to taking his present posttion in November 1985, Mr. Bernero was in HRC's Office of Research and at NRR in positions involved in risk analysis, source term research, and regulatory utilization of such information. In those positions Mr. Bernero gave many public statements about the regulatory process including the reassessment of emergency preparedness requirements. Mr. ero did discuss such issues with Mr. William Derrickson and others of PSNH on occasions. These were informal meetings which took place in Mr. Bernero's offic One took place on July 30, 1985, and Mr. Bernero recalls the other one as being a few months late P -

(

Markey/NRR 9/24/86 O

9 l

6 .#w

, F$ Nl+ l<.6tPc wed Tb Ad%ssi Ke cuEST Fcs.eHr m s.co.y er

• C t uutJi c hTi t.Ni wirH Mi2.d.

[ -

08/27/86 Letter from B. J. Huselton (UE&C-Phil) to C. Hoffsayer transmitting containment-model data requested previously.

08/18/86 Notes of 08/14/86 trip to BNL from R. E. Sweeney to Distribution #SBB-86-126.

07/09/86 Meeting to discuss PSA program at Bethesda among NRC, PLG, and PSNH.

07/21/86 SBN-1167, submits SSPSA Update.

07/29/86 SBN-1173, submits S$PSA Update / Review - discussion.

06/86 SMA 1589.01, " Seismic Fragilities of Structures and Components at the Seabrook Generating Station issued.

. 05/17/86 SBN-1053, responds to NRC/LLNL draf t review of SSPSA.

02/04/86 Meeting between Derrickson and Denton to discuss Seabrook.

05/86 "Seabrook Station Emergency Planning Sensitivity Study" issued (PLG-0465).

P14:-0432, "Seabrook Station Risk Management and

[ 12/85 Emergency Planning Study issued.

07/30/ Meeting - NHY Management /NRC Management , Discussion s of SSPSA submittal.

P 06/12/85 SBN-842: SSPSA review by NRC.

In the time period following January 1,1985, there were management meetings, formal meetings, and telecons with the NRC staf f with respect to the Emergency Planning Zone for Seabrook Station. However, because formal or. personal notes are not available for all meetings or telecons,

, to the best of our knowledge, we cannot provide you with a detailed chronology, or identify of all meeting participants or cover with any l

specificity all topics of discussion. However, based on available l notes and recollections of these meetings available at this time, we believe the above chronology addresses your question.

( R_esponse to Part (b)

See Attachments.

l l

l l

l l Revised 11/06/86 i

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0 7-FGH& Srwe nere FGHk ID tc mpg AGENDA FOR SEA 8R00K COORDINATION MEE

. November 26, 1985 -

ency I. FEMA ITEMS impact on emerg .

< # \/

1) Probabilistic risk assessment p1anning issues.

Future FEMA sponsored training. -IT)ce.ES M*M 5(ou

2) and f ull-scale exercise. .

Status of training drill 0.8

3) .

- Impact of plan submissions -

l

. Remarks by Spence H. Joseph Perry, Flynn, Acting Attorney General Counse Advisor 4) l, What is the status of the ETE up-date?

5)

UTILITY ITEMS ing*A*6. 4 4

!!. M ghcAa -

" Construction schedule up-date - ASLB and pl ann M M6

' I 1) 1/r submission considerations. *bsdoKpo W

2) Other issues. SmtWM GC III. STATES SW%h g j 1)

Report f rom drill / exercise group. c. w%

2)

Status of ETE study.

3) Other issues. W

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, PSbill It b-?(, R.csPcusg 08/27/86 Letter f rom B. J. Huselton (UE&C-Phil) to C. Hof fmayer transmitting containment-model data requested previously.

08/18/86 Notes of 08/14/86 trip to BNL from R. E. Sweeney to Distribution fSBB-86-126.

07/09/86 Meeting to discuss PSA program at Bethesda among NRC, PLC, and PSNH.

07/21/86 SBN-1167, submits SSPSA Update.

07/29/86 SBN-1173, submits SSPSA Update / Review - discussion.

06/86 SMA 1589.01, " Seismic Frag 111 ties of Structures and Components at the Seabrook Generating Station issued.

05/17/86 SBN-1053, responds to NRC/LLNL draf t review of SSPSA.

02/04/86 Meeting between Derrickson and Denton to discuss Seabrook.

05/86 "Seabrook Station Emergency Planning Sensitivity Study" issued (PLG-0465).

/ 12/85 PLG-0432, "Seabrook Station Risk Management and I

Emergency Planning Study issued.

07/30/85 Meeting - NHY Management /NRC Management - Discussion of SSPSA submittal.

06/12/85 SEN-842: SSPSA review by NRC.

l In the time period following January 1,1985, there were management meetings, formal meetings, and telecons with the NRC staf f with respect to the Emergency Planning Zone for Seabrook Station. However, because formal or personal notes are not available for all meetings or telecons,

. to the be'st of our knowledge, we cannot provide you with a detailed chronology, or identify of all meeting participants or cover with any specificity all topics of discussion. However, based on available notes and recollections of these meetings available at this time, we believe the above chronology addresses your question.

Response to Part (b)

See Attachments.

Revised 11/06/86

- ~ , . - . . - . . - - . . . - - _ . . . . . . . . , . . . . ~ _ _ _ , - - _ , , - - _ - - - - . . . , . , - , - _ - , _ . _ , , . , . , _ _ , - - - - - - - - c,-- , n ., ,

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George S.Thomcm W l July 29,1986 vice hedders-Nuclear Producson SBN-1173 -

MNdMM T.P. B7.1.99, 37.1.8 .

New Hampshire Yankee Olvision '. ~

I.

United States Nuclear Regulatory Commission Washington, DC 20555 ,1-f.

Attention: Mr. Vincent S. Noonan, Project Director -

PWR Project Directorate No. 5 taferences: (a) Construction Permits CPPR-135 and CPPR-136, Docket Nos. 50-443 and 50-444 (b) PSNE latter (SBN-1167), dated July 21,1986, "Seabrook Station Probabilistic Safety Assessment Update",

J. Devincentis to V. S. Noonan (c) PSNE Letter (SBN-1152), dated July 3,1986, "Reques t for Issuance of an Operating License; Seabrook Station Unit 1", W. B. Derrickson to H. R. Denton Subj ect s Seabrook Station Probabilistic Safety Assessment Update

Dear Sir:

t Reference (b) submitted, for your review, new information we intend to include in a future update of the Seabrook Station Probabilistic Safety Assessment (SSPSA). The importance of the Staf f's technical -

, review of these documents has been heightened by articles appearing -

recently in local newspapers ( Attachment 1). These articles discuss the

' ~

apparent strategy of the Commonwealth of Massachusetts to intentionally -

, delay the Seabrook licensing process.

Based on these developments, it ie clear we must request the NRC to .

expedite the technical review of the Reference (b) attachments so that

, we can determine whether there is an adequate basis for future submittals/ .

requests. A future submittal, depending on the results of the technical review, may request a change to the emergency response plan process for

• Seabrook Station. We cannot, at this time, specify what action such ,

, a future request may seek, but it is faportant that we address as soon as '

possible what options are available to us relative to full pover licensing.

This is important in light of the apparent strategy of the State of Massachusetts to delay the process.

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BROOKHAVEN NATIONAL LABORATORY ASSOCIATED UNIVERSlilES. INC.

Uptort Long161055.dewYork 11973 (616) 282s 333, FTS 666' Cmice of the Director ,

August 14,1986 Mr. David Schweiler, Manager .

. Brookhaven Area office .

D. 8. Department of Energy Upton, new York 11973 .

- Dear Mr. Schwe11ert Enclosed is one copy of a proposal to the Nuclear Regulatory Countasion entitled, "Rev.iew of the Emergency Planning sensitivity study for seabrook "

FIN A-3852,'being submitted for your review and approval. The propeaal is being submitted in response to a verbal request from D. Floravante of the NRC. Two copies have been sent to Mr. R. W. Barber. Department of Energy, one copy has been sent to Mr. M. Raltaan, Nuclear Reguistory Coun!ssion, and three copies have been sent to:

Ms. Diane 5. Fioravante Program Assistant Division of FWR Licensing-A Offlee of Nuclear Reactor Regulation

. -. D. 8. Ikielear Regulatory Cmunission Washington, D. C. 20535 The total cost of this progras is $245,000. No funds have been ob!! gated to date.

l

- This proposal has been prepared in accordance with discussions held be-tween W. T. Pratt (BNL), C. Hofmayer (SNL) and NRC ataf f. If there are any l

questions regarding the document, please es11 the principal investigator or Mr. A. J. Romano, FTs 466-4024, Department Administrator for the Department of helear Energy.

sincerely'yours,

.- (

8. Baron 6 ---.

Associate Director '

for Applied Programs i Enclosures cc: R. W. Barber (2) N D. Floravante (5) s '

M. Kattaan N.

l l __ _._____.____ _ _ __ _ ___

y ,

uA huc6aAm AtQvtAfont ocuwissicM eATo er P mAi, nac Pon w see 88 **8 Aurust 11. 1986 PROJECT AND BUDGET PROPOSAL FOR NRC WORK m us.

O navisio= i.e enesis, fir 6:

Pi==wwasa saview of the Emergency Planning Sensitivity A-3852 Study for Seabrook I insseaavusen

""C C"'C8 Nuclear Reactor Regulation l 20-19-4041-4 ,

con co vnActon Assoctated Univers1 ties. Inc. lpyyd',5rSa^5'o**'

Brookhaven National Laboratory- l N*w sifs

• IoosseAawuesa Upton, New York 11971  : l PTEPHopetNWhettR ytmiocoP PSRPoAWJ

~

C0CNIZANT PE Rao488 Ek ORGANIEAftoM NRC PAOJECT M ANAQ(A [ 4TARTiho OAfa C. Baschi NRR/DPIA/tB 492-7070 I 0s/15/s6 oTM84 Nmc T8CMNICAL STAPP , [COMPLmoN oAM i 09/30/87 Xs_e.

" " d'I"""*"'"

David Schwellar N$o)$a 666-3424 mg,;-grey coNTaacToa-Pmostcf uAnAQsm y' ' BNL/DNE lll M PR4NCIP AL INV E4760 Afonist W. T. Pratt ~ C. Hofmayer BNL/DNE 666-2317 - - - eTAPP YE ARI oP SPPOA7(Aemnd as meser more efe pearf PY 1986 W 1987 PT PT ET o.o senaune/Teeaiset o.3 o,3 ( ow ce.n sommel o o TOTAL DIRICT STAPP Ys Ans 0.8 0.8

                                 ~

COST P AopORAL I one serm. 56 54 u , i w su.o. is.wg Aop 1 1 AoP se 3 3 l s w w.as . 0 0 TresetEspeaws - Demostw & A sae.ma c " 18 16 on omneir rs . n n o.aomi and Amni.irati.et 43.5si 3g 3y l l TOT AL OPE R ATIwo ons? 125 120 CAe:TAL eouiPwsNT 0 0 . TOTAL PfloJECT cott 125 120

                                     ,, = w as                                                                        * ****                        ***"                   d'**^*'       '***'

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                                <.srr.ea= = iss 44 441
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              *                                                                         ~

wac roa+  ; es ~ WA NvCMAR NEsWLA.hY GGWWilslO% FM NWWsIR PROJSCT AND SUDGET PROPOSAL POR NRC WORK .

                                                                                                                                                          .Av.

August 11, 1984 rnew 1.m l1 . I Review of the tunergency Planning Sensitivity Study for Seabrook oos rpososa s saeamsarious Asoociated Universities, Inc. , Brockhaven National 1.aboratory POR EC AST Wiklif 0NE CH ART: Behedwies to Start = = Cemswtes (Stews M gharter Farf PROV10g estiWATED DCLLAR CCST POR S ACH TAsit ma H pisCAL YEAR

                                  ,                                            pylyoo           PY 1957                         PY              FY                fy TA8" ra idelwl4m is lmelm i sm inldelvsima valdelwlan                                    ta l pelwlan sensouts            *g4                           ;

Task 1 cost 25K 13K-

, eensouts &4 Task 2 -

cost ' 25K 13K scusputs g.g - i' Task 3 cost 50K 47K I sannouLs yg 4 Task 4 .

                                                                       "          13K l

scusouts gg Task 5

• i **** 12K 13K i

j sensoute 7,,,

                                      .                                                                                     ,,                                                   j Task 6                                    gegy                              34g                                     ,-

scHsouts cost . I scuseuLa , a cost sensouts con , scusouts cost scusovos cost . Terrat estiu Atso enoasce cost 125K 120E l ar, . 46avea 1.-..e ar as l "

j'

                                                                                       . . . . .                                                                                 \

Review of the Energency pisaning Sensitivity Project

Title:

Study for Seabrook _ _ _

o. 3 PROJECT M SCRIFTION ,,

i

1. OsJtcTIVE OF FROPOSED WORE .

l

a. Backaround On July 21, 1986 Public Service of.New Hampshire (PSNE) submitted a s neitivity

study on the eastgency planning sone (EP2). The study provides a comparison of dose versus distance curves for the Seabrook plant and site with similar generic curves from NURIG-0396 which were used in developing the BPE regulation in 10 CFR 50.47.

The study concludes that a 1-mila evacuation radius at seabrook provides for a stai-i lar or greater degree of public protection than was shown by NUREC-0396 for a 10- i

               .        mila evacuation radius around the plants considered,by WA85-1400.

The study is largely based on the Seabrook Probabilistie Safety Assessaant that F5NR submitted about 3 years ago. The source terms used in the Emergency Planning Sensi-

                     . tivity study were drawn from the source teras used in the WA55-1400 calculations, with some modifications under specific scenarios. Also, sons of the probabilistic models have been changed from the Safety Assessment. Thus, the report is intended to azanine differences made by the Seabrook design and site, plus the improvements la accident seguence modeling espabilities, without credit for source tars redue-tions that may result from recent studies. The RF8 study attributes reductions in
 '                       their offsite dose predictions to the higher strength of the Saabrook containment, a more refined failure modes analysis for the containannt, and a more realistic treat-meat of the initiation and progression of interfacing systems 1,0CA sequences. Along with the Eastgency Planning Sensitivity Study. PSNE has also submitted a report titled "8eabrook Station Risk Management and Isergency Planning Study " which pro -

vides results of Seabrook specific calculations with saw source terms based upon tha  ; i recent IDC01 work.

• i The applicant has requested that the technical serits of the BFE study be reviewed with respect to its adeguacy to support a change to the emergency response process.

The exact nature of the change has not yet been specified. PSNR has further re-guested that the review be completed on an expedited basis. .

• The concluetons of the EFE Sensitivity Study are based upon comparison of the re-euits of the study to three acceptance criteria that were drawn fras Ntc documents.

One of the criteria is a comparison of the individual risk of early fatality in the - population within 1-s11e of the plant, assuming so immediate protective action, to the Mtc proposed safety goal. A second criterion ip the comparison of early fatali-

                        ' ties at the Seabrook site, assuming a 1-mile evacuation, to the early fatalities re-sults of VASE-1400. Wich assumed a 25 mile evacuation. The third criterion is the comparison of the risks of exposure to 1. 5. 50, and 200 ran sole body doses at various distances from the seabrook site to the corresponding NUREG-0396 results at 10 miles, assuming no imediate protective actions.

To properly review the EFZ 5ensitivity Study, it will be necessary to identify the baseline against which comparisons are made, to identify the appropri' ate criteria for asking the comparisons, and to review the basic assumptions and the more signif-icant aspects of the probabilistic calculatione. . .. (See Continuation sheet) ,

Review of the Emergency Planning Senaitivity l i

s. 4 Project

Title:

Study for Seabrook

n. Obioetive .-

The objective of this effort is to assist the NRC in evaluating the technical valid-ity of the applicant's conclusions regarding the Energency pisaning Sensitivity study for Seabrook.

           ~

2. SLDO(ART OF PRIOR F,FFotT8 . A review of the Containment Failure Modes and Radiological Source Terza presented la the Seabrook Station Probabilistic Safety Assessment was performed by BNL under FIN A-3778. Bovover, this review did not evaluate in detail the containaant performance under severe accident conditions. Nevertheless, the experience gained by SNL staff

                  -       under FIN A-3778 will provide useful background for,the current review of the Emergency Planning Sensitivity Study for Seabrook.

3. ll0RK TO BE PERFORMID AND EKFECTED RESULT 8

a. Work taquirements Task 1: Systen Evaluation BEL will review those pertions of the seabrook Energency Planning Sensitivity 8tudy related to system failure to determine the appropriateness of the calculated acci-dont sequence probabilities. In particular, the probability for interfacing systes LOCA vill be carefully assessed to determine the potential for containment bypass.

[ ENL will also review the probability of equipment malfunctions, personnel errors, and design errore resulting in containaant bypass at the tian of a severe accident. Estimated Level of Ifforts (FT 86) 2 Professional Staff Months (psa)

             .      -.                                     (FT 87) 1 pse

! Estiasted Cosgletion Date: October 31, 1986 Task 2: Containment Event Tree Review BEL will review the conditional probabilities of early containannt failure given in

                    -      the seabrook'aubmittals. In particular, the vulnerability of the Seabrook contain-

• ment to uncertainties in containannt loads will be carefully assessed. This task will be highly coupled to Task 3, ubich will assess the performance of the Saabrook containment under severe accident conditions. .

Estimated Inval of Efforts (FT 86) 2 pam - (p1 87) 1 pse Estimated Completion Date: October 31, 1986 Task 3: Evaluation of Containment Sahavior , The purpose of this task is to evaluate the technical validity of the applicant's conclusions rosarding the behavior of the Seabrook containment under severe accident conditions. (See Continuation Shaet) [ s

                                                .                                                                                                                           y                            ,
                              -          s                                                                                          r.                                                                   i 1

Review of the Emergency Planning Sensitivity l Project

Title:

Study for Seabrook 3. $

3. WORK TO BE PERFORMED AND EIFECTED RESITLTS (Cont.)

a. Vork Requirements (Cont.)

Task 3: Evaluation of Containmen_t Behavior (Cont.) 3E will review and evaluate the relevant containment structural analyses performed by the applicant and its consultants. In addition, a plant site tour and engineer-ing audit at the applicant's (or consultant's) office will be conducted to better understand the containment analyses and design and to identify any unique design features and/or analytical assumptions that serit further investigation. Based on the above review, EE will develop an exisyanstric finite element model and perform analyses utilising 3E's NFAP computer code to confirm the applicant's pre-diction of the overall capacity of the containment. Special attantion will be given to the post-cracking behavior of the concrete which controls the shear failure ande of the containment. To expedite the performance of this task. EE will utilise, to the nazimum extent practical, the input paranators obtained from the applicant's analytical models. In addition, staplified hand calculations will be performed to assess the applicant's conclusions regarding the behavior of selected containasnt penetration assemblies. Finally, SE will perform a qualitative assessment of the applicant's seisate fragility analysis of the containment structures,and components. ENL will also support meetings with upper NRC annagement and the ACRS to describe 4 the interia status of this review, as well as the final results. Estimated Level of Efforts (FT 86) 4 pas (FT $7) 3.5 pan , Estimated Completion Date: October 31, 1986. i Task 4: Review of Source Tarse The appropriateness of the new source terms based on RSS esthodology used in the Saabrook submittal will be reviewed.

• Estimated Level of Efforts (FT 86) 1 paa -

Estimated Completion Date September 30s 1986. - - -- - . - ^- - Task 5: Site consecuence Modelinz The site conseguence modeling will be reviewed to determine the appropriateness of - the consequence calculations presented in the Seabrook submittal. In addition, any conseguance calculations found necessary as a result of the work to be performed under Tasks 1 - 4 will be performed. Estimated Level of Efforts (FY 86) 1 pse - (FT 87) 3 pen - Estimated Completion Date: Octo r 31, 1986 (See Continuation Sheet) l

  -,,..-y     - - - , , - ,              a-,  .   ,,,,-m,     . , . _ , , , . - - , - _ - . , - ,   .--,--,,_v._ , - . - - - - _ - - - -      .w-- . - . . , -     -

7

                                                                                                                                                   )                            ,
           ~

Review of the hargency Planning Sensitivity Project

Title:

Study for Seabrook p. 6 -

3. WORK TO BE PERFORKED AND IIFECTED RE8UI.T5 (Cont.)

Task 6: 3enort Freoaration and Follow-on Effort A final report will be prepared based on the results of Taske 1 - 3. .In addition, any follow-on effort in terms of resolution of issues will be provided under this task. The final report will be 1seved on 12/31/86 and will address B E's recommen-dations on procedures, testing or design nodifications, to redgee the probability of containment bypass in monjunction with a severs accident. Estimated Inval of Efforts (FT 87) 1.5 psa Estimated completion Date: september 30, 1987 ,

                        .                    Level of Effort and Performance The estimated total level of effort is 19 professional staff-sonths over a period of 14 months.

b. Meetinas and Travel It is expected that therr., nal be two trips, one to the seabrook plant site and one to UE and C's offices in Phladelphia, PA, for three 3E engineers each lasting 2-3

( days. In addition, three one-day meetings in Washington, DC are anticipated. One aseting may require 6 BNL engineers, the other two will require 2 BE engineers.

a. EC Furnished Waterials -

N2C will provide all documents prepared by the applicant or its consultants which a:a to be reviewed by 3NL.

4. DESCRIPTION OF ANY FOLLOW-0W EFFORTS Not applicable.

5. Ez1ATION8MIF_ To OTHIt PROJECTS SNL has ongoing efforts for MAR /NRC in the area of severa accident analysis and risk '

assesenent. These include peer review of plant-specific, industry-sponsored prob-abilistic risk assessaants. To date BNL has reviewpd various aspects of the follow-ing PRAs Big Rock Point, Indian Point, Eton, Limerick, GESSAR, shoreham, oconee-3, M111 stone-3, and Seabrook. BNL also has ongoing efforts for RES/ RC dealing with

severe core damage, fission product release sad transport to the environnant. In particular, BNL will verify the source Tara Code Fackage and the calculations to be performed as part of the updating by SARF of nuclear accident risk (NUREG-1150). In 2 addition, BNL staff are evaluating uncertainties in source ters estiastes (FIV A-3286). Finally, SNL staff are providing assistance to the NRC staff in developing a technical basis for offsita energency preparedness under FIN A-3268.

(See Continuation sheet)

                                                                                                                                         . ,- - - - ,,,   ,-y, . ,--m-,i          -, - - - - - - - -.  --
             -__-________._________.,.----,,.--.,-----,-,-.,-,,.---------,--,--r-,,-

Review of the Emergency Planning Benettivity Project Titlet Study for Seabrook 3. 7

6. REPORTIIIC REQUIllDSNTS P

Technical _ tenorts - gis copies of all technical reports which are required for the program shall be submitted to the NEC Project Manager. ' Business Letter Reports i . A monthly bustosse letter report will be submitted by the 20th of the month to Mr. Goutan Baschi. NEC Project Manager, with copies provided to the Director, Division of FWR Licensing A. Atta D. Fiorvante end M. Kaltaan. NER. The report will iden-tify the title of the project, the FIN, the Principal Investigator, the period of performance, and the reporting period and will contain 1 sections as follows:

 ~

a. Project Status Section

1) A listing of efforts completed during the period, milestones reae'hsd, or if missed, an enplanation provided.

2)' Any probless or delaya encountered or anticipated and recommendations for resolution. (NOTE: If the recommended resolution involves a coat-rect modification, i.e., ehenge in work requirements level of effort

             ;                                                      (costs), or period of performance, a separate letter will be prepared and submitted to Coutam Baschi.)

3) A summary of, progress to data.

4) A brief summary of plans for the next reporting period.

i

b. Financial Status Section BNL will provide the personnel time expenditures and the amount of funds expended (costed) by category during the period and total cumulative

         ,                                                   year-to-date se follove
                          .                                                                                                      current Month             Year-to-Date f

1. Direct Staff Effort (S&P Staff Months) .

11. Direct Salaries * ' ~

Meterial and services . . ADP support Subcontracts Travel Espenses - Indirect Expenses Canaral and Administrative , Total Costs - Percentage of available funde spent (See continuation Sheet)

Review of the Energency Planning Sensitivity Project

Title:

Study for Seabrook 9. $

7. SU5 CONTRACTOR INFORMATION P BNL does not intend to subcontract any portion of this work. In the event a subcon-

                                                                                                    ~

tract te anticipated. BNL will notify MRC before it is initiated.

4. NEW CAPITAL EQUIPMElfT REQUIRED ,-

i None. .

9. SPECIAL FACILITIES REQUIRED Eona.

10. CONFLICT OF INTEREST INFORMAT!0N ,

There are neither significant contractual nor organisations1 relationships of the

                   -   Department of Energy. BNL and employees, or expected subcontractors or consultants on this proposal, with industries regulated by the F.RC and suppliers thereof that give rise to an apparent or actual conflict of internet.

11. EXPECTED CLASSIFICATION 01 SENSITIVITY

! This is an uncloseified program. safeguards, proprietary, or other sensitive infor-nation is not involved. 4 5 6. l l . D

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SUBCOMMITTEE ON ENERGY CONSqh b** AND POWER ,,, ,,7lg'6 ,, , '

         'VARD J. MARKEY                                                                                                          ""                                    " " 

CHAIRMAN COMMITTEE ON ENERGY AND COMMERCE -- U.S. HOUSE OF REPRESENTATIVES *"'*"*"8""*"a WASHINGTON, DC 20515 a, nSIl- . NEWS RELEASE EMBARGOED FOR RELEASE 1

Contact:

Raoul Rosenberg Until 2:00 p.m., (202) 225-2836 Saturday, November 15, 1986 Phillip Greenberg or Linda Correia (202) 226-2424 MARKEY REVEALS WHITE BOUSE PRESSURE ON SEABROOK EMERGENCY PLANNING Washington - Rep. Edward J. Markey (D-MA), Chairman of the House - - Subcommittee on Energy Conservation and Power, today released information uncovered during a congressional investigation suggesting that the Federal Emergency Management Agency tested the New Hampshire emergency plans for

       'he Seabrook nuclear power plant in response to White House pressure.
                     " fem officials had told New Hampshire that they felt their emergenh plans clearly were not ready to be *.ested by an exercise," Markey added.
       "Moreover, FER had informed New Hampshire officials that in cases where two states were involved - in this instance, New Hampshire and Massachusetts -- FdMA routinely preferred te hold an exercise in which both states participated, rather than risk expending their scarce resources to hold two separate exercises. Not surprisingly, after a call from the White House FEMA jumped and proceeded to conduct the exercise."

Markey said tha't FEMA staff told the Subcommittee's investigators that FEMA had received a call from the White House suggesting that they hold an exercise with New Hampshire alone, and that regional FEMA

    %_ rsonnel were then directed to go ahead with the single state exercise.

Markey said, "Four officials of th'e Federal Emergency Management Agency have informed our Subcommittee that they believed that Governor i ~ John Sununu of New Hampshire had ' called the White House to complain a, bout the Agency's handling of the Seat $ rook emergency planning reviws end e xe rci se.. If this bell'ef is accurate, such an attempt to circumvent the real procedures of the emergency. planning review process can be vieved

                                                                                                                                                                             ~

only as an improper attempt to biing political pressure to' bear on FEMA - the Federal agency responsible for reviewing the adequacy of emergency planning around nuclear power plants. Th% I E L. . . . . . . v. . . . . . - .. . - - dispassionate evaluation, not a politically motivated shuffle."

                                                                                                                                                       . - , , - . ,              m.-., - - - - -

FEMA personnel said that ordinarily the first exercise of a state's nuclear plant emergency plans reveal a few " deficiencies" - the term FEMA uses to describe serious problem areas in a plan. In the exercise of the New Bampshire plan.they had uncovered an unusually large number of 4 deficiencies, which they felt had demonstrated to New Hampshire authorities that the New Eampshire plans badly needed improvement. Notes f rom FEMA meetings on Seabrook af ter the exercise showed that Governor Sununu had telephoned FEMA " wildly upset" about their evaluation of New Bampshire's plans, and had said that he was "on his way to the White House to clean out (FEMA's) staf f operation." Markey added, "For a Governor to say that he's on his way to the . White House to ' clean out' a Federal agency's staff operation is a bald attempt to intimidate Federal employees pursuing their mandated duties -- particularly when he's going to a White House occupied by a President of his own party who everyone knows appointed the head of the Federal agency in question. Such political maneuvering can hardly serve to bolster public confidence in Federal decisions regarding one of the most contested nuclear plant licensing decisions in history."

                                   "Notwithstanding statements in his recently concluded campaign to the contrary, it is now crystal clear that Governor Sununu has been actively involved in trying to get Seabrook licensed as quickly as possible. As a

f ar as I am concerned, no economic benefit is worth compromising the l I safety of the people who live in the vicinity of Seabrook on both sides of the border," Markey said. l l f Markey said he will pursue these questions at the Subcommittee's upcoming field bearing on Seabrook to be held Tuesday, November 18, 1986 in Amesbury, Masschusetts. Governor Sununu has been invited to testify at the hearing. l

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( 4.a._J4 f-0 ef % pew W-) QUESTION 17, IN RESPONSE TO THE SUBCOMMITTEE'S AUGUST 28, 1986 REQUEST FOR INFORMATION, NRC PROVIDED TWO MEETING CHRONOLOGIES AND TWO LISTS OF INTEPHAL AND PUBLIC DOCUMENTS (ENCLOSURES 1 - u), PLEASE PROVIDE UPDATES OF THESE CHRONOLOGIES FOR THE SUBCOMPITTEE WHICH INCLUDE ALL COMMUNICATIONS, INCLUDING BUT NOT LIMITED TO PEETINGS, TELEPHONE CALLS, AND RECORDS, SINCE THE LAST COMMUNICATIONS AND RECORDS ENUMERATED IN YOUR PREVIOUS ANSWER, ANSWER, ENCLOSED APF UPDATES OF TWO MEETING CHDONOLOGIES AND TWO LISTS OF INTERNAL AND PUBLIC DOCUMENTS INCLUDED AS ENCLOSURES 1 - 4 TO THE STAFF'S SEPTEMBER 10, 1986 LETTEP TO THE SUBCOMMITTEE. ALSO EFCLOSED ARE COPIES OF THE INFORMATION LISTED IN ENCLOSURES 3 AND 4.

        ' ENCLOSURES:

AS STATED (4) MARKEY/NRP 11/10/F6

i ENCLOSURE ] CHRONOLOGY OF MEETINGS AND DISCUSSIONS WITH UTILITY AND OTHER GROUPS - SEABROOK UPDATE 09/08-09/86 SITE VISIT BY NRC AND BNL PEPSONNEL (SEE MEETING

SUMMARY

). 09/10/86 CONFEPENCE CALL WITH NRC, BFL AND PSNH PERSONMEL TO DISCUSS POSSIBLE SITE SIMULATOR VISIT. l 09/25/86 TELEPHONE CALL BETIIEEN E. DOOLITTLE AND M. HAYES (HAVERHILL GAZETTE) TO DISCUSS ACRS MEETING. 09/26/86 JOINT MEETING OF THE ACRS SUBCOMMITTEES ON OCCUPATIONAL AND ENVIRONMENTAL PROTECT!nN SYSTEMS AND SEVERE (CLASS 9) ACCIDENTS. 10/.10/86 SEABROOK ACPS FULL COMMITTEE MEETING. 10/]5-17/86 SEABROOK SITE VISIT (S. LONG, G. BAGCHI, D. HICKFAN, W. LYON, R. YOUNGBLOOD, C. HOFMAYER, D. WESLEY, J. MOODY). MARKEY/FPR

                                                             ))/10/86

l ENCLOSURE 3 _q_ 10/29/86 Two CONFERENCE CALLS WITH NRC AND PSNH

                                   -  LONG, BAGCHI, LYON, MAIDRAND
                                   -  LONG, LYON, MOODY In/30/86 TELEPHONE CALL BETWEEN E. DOOLITTLE AND ,l. POUGHTY TO DISCUSS UPCOMING MEETINGS ON F0V. 6, 12 AND 10 l

M6PKEY/NRR 1.'/10/06

ENCLOSUPE ? i CHRONOL 0GY OF INTERNAL MEETINGS - SEABROOK UPDATE 09/03/86 MEETING WITH LONG, MATTHEWS, PERROTTI, PERLIS, SOFFEP TO DISCUSS CPITERIA FOP COMPARISON WITH NUREG 0396. 09/04/86 MEETING WITH NOONAN, LONG, DOOLITTLE, BAGCHI TO DISCUSS REVIEW STATUS. '

                                                                                      =

1 09/11/86 MEETING WITH NOONAN, LONG, DOOLITTLE TO DISCUSS PEVIEW STATUS. 09/12/86 CONFEDENCE CALL WITH NPC AND BNL, 09/16/86 MEETING WITH N0vAK, NOONAN, LONG, DOOLITTLE TO DISCUSS UPCOMING ACRS MEETING. 09/22/86 MEETING WITH NOVAK, NOONAN, LONG AND OTHERS TO DISCUSS REVIEW STATUS. 09/23/86 MEETING WITH NOONAN, DOOLITTLE, LONG TO DISCUSS REVIEW STATUS. FARKEY/PPP

                                                                          .1) /.10/86
                                                                    ,.--w       --  g   n + --, m 4             -,,-

g Ef1 CLOSURE ?

                                             -p.

09/24/86 MEETING WITH NGONAN, DOOLITTLE, l.CNG TO DISCUSS REVIEW SCOPE AND OBJECTIVES. 30/07/86 MEETING WITH ROSSI, BAGCHI, NOONAN, LONG TO DISCUSS REVIEW STATUS. 10/08/86 MEETING WITH NOONAN, NERSES, DOOLITTLE, LONG TO DISCUSS REVIEW STATUS. 10/22/86 MEETING WITH f0VAK, NOONAN, ROSSI, LONG, DOOLITTLF, BAGCHI TO DISCUSS REVIEW STATilS. 10/23/86 CONFERENCE CALL WITH NRC AND BNL (LONG, LYON, BAGCHI, PRATT).

            , 10/30/86  MEETING WITH NRC AND BNL (NOVAK, NOONAN, LONG, BAGCHI, ROSSI, PRATT, BARRY).

MARKEY/NRR l 11/10/86

e , ENCLOSURE 3 s LIST OF Pustic DOCUMENTS - SEABROOK 09/26/86 . TRANSCRIPT OF ACRS SUBCOMMITTEE MEETING, 09/29/86 LETTER FROM G. THOMAS TO V, NOONAN PROVIDING ADDITIOFAL INFORMATION, 10/08/86 LETTEP FROM S LONG TO R HARPISON REQUESTING ADDITIONAL ~-INFOPMATION, . 10/10/86 TRANSCRIPT OF ACRF FULL CnMMITTEE MEETING. ( 10/23/F6 LETTER FROM S. LONG TO P. HARRISON REQUESTING ADDITIONAL INFORMATION G, THOMAS To V, NOONAN, 10/27/86 LETTER FROM V STELLO TO J. SUNUNU PEGARDING STATE LIAISON OFFICER RICHARD STROME, l l < l 10/31/86 LETTER FROM J. DEVINCENTIS TO S. LONG PROVIDING l RESPONSES TO REQUESTS FOR ADDITIONAL INFORMATION. i i MARKEY/NRR 1]/10/86 1 l

( . o

   'fO ENCLOSURE 4
    \

LIST OF INTERNAL DOCUMENTS - SEABROOK I 09/03/86 NOTES FROM MEETING WITH MATTHEWS, KANTOP, PERLIS, SOFFER, PERROTTI, LONG CONCERNING CRITEPIA FOP PISK COMPARISONF. 09/04/86 NOTES ON MEETING WITH NOONAN, DOOLITTLE, LONG, BAGCHI

                            - REVIEW STATUS.

09/09/86 NOTES ON SEABROOK SITE VISIT BY NRC/BNL. 09/11/86 POTES ON MEETING WITH NOONAN, LONG, DOOLITTLE - PEVIEW STATUS. 09/15/86 PEMO FROM SPEIS TO NOVAK CONCERNING SCOPE OF PJ.'l PEVIEW. 09/16/86 NOTES ON MEETING HITH NovAK, F00 NAN, LONG, DOOLITTLE

                            - ACRS MEETING.

09/19/86 FACSIMILE TRANSMISCION FROM R. YOUNG 9LOOD TO S. LONG - l PAGES FROM NSAC-84 (ZION NUCLEAR PLANT RESIDUAL HEAT REMOVAL PRA), 1 09/22/86 MEMO FPnM JORDAN TO NOVAK - CRITEPIA TO EVALUATE SEA-j BROOK EPZ SENSITIVITY STUDY. MARKEY/PPR 11/.10/86

f & W g Et!CLOSUPE li 09/?2/P6 ROUTING SLIP FROM NovAK TO PAGCHI, NOONAN, LONG r.0VERING 9/15/86 MEMO FROM SPEIS TO N0vAK. 09/23/86 NOTES ON MEETING WITH NOONAN, DOOLITTLE, LONG - PEV!EU STATUS. 09/23/86 NOTES FROM PUBLIC PETING AMONG NRC, BNL AND PSNH TO EXCHANGE. TECHNICAL INFORMATION. 09/?h/86 NOTES ON MEETING WITH NOONAN, DOOLITTLE, LONG - REVIEW STATUS. 09/25/86 MEMO FROM SPEIS TO NovAK - ADDITIONAL COMMENTS ON THE ENL REVIEW OF SEABROOK EMERGENCY PLANNING STUDY.

        . 09/25/86 NOTES ON CALL WITH DOOLITTLE AND HAYES.

09/26/86 NOTES ON SEAPROOK ACRS SUBc0MMITTEE MEETING. 09/29/86 MEMO FROM HEPHAN TO NRR DIVISION DIRECTORS - AIF PAPER ON EPZ vS SOUPCE TERM. MARKEY/NPP D /.10/86

 ',     c
    . 9 ENCLOSURE 4 30/01/86 LETTER FROM R. E. PHITE TO BNL - SEABROOK PSA STUDY -

PIPING IS0 METRICS, 10/06/86 MEMO FROM LYON TO BERLINGEn - PEVIEW OF SEABROOK DOCUMENTS PERTAINING TO CHANGE IN EMERGENCY PLANNING ZONE SIZE, 30/07/86 SEISMIC FRAGILITY UPDATE. 10/08/86 NOTES ON MEFTING WITH f00 NAN, NERSES, DOOLITTLE, LONG - REVIEW STATUS, 10/09/86 LETTER FPOM FLEMING TO MAIDPAND - KEY FACTORS INFLUENCING UPDATED ANALYSIS OF INTFRFACING LOCA,

        .10/10/86  NOTES FROM ACPS FULL COMMITTEE MEETING, 10/15/86 PURPOSE AND AGENDA, IRIP TO SEABROOK STATION AND TO BROOKHAVEN NATIONAL LABORATOPY, 10/21/86 MEMO FROM BERLINGER TO NEPSES - SEABROOK STATION PISK EVALUATION PEPTINENT TO EMEPGENCY PLANNING, MARKFY /NPP.

11/10/86 6

ENCLOSURE 4 ,

                                                               -4_

10/??/86 NOTES FPOM BRIEFING FOR NOVAK OF REVIEW STATUS. 10/23/86 NOTES FROM TELECON BETWEEN NRC AND PNL CONCERNING CONTAINMENT EVENT TPEES. Y 10/22/86 AGENDA FOR BRIEFING ON REVIEW OF SEABROOK EPZ STUDY. t 10/?4/86 MEMO FROM LONG TO NOONAN - STATUS OF NRC REVIEW OF SEABROOK EMERGENCY PLANNING SENSITIVITY STUDY. 10/?9/86 NOTES ON CONFEPENCE CALL WITH PAIDRAND, SAfJCHEZ, LYON,

 )                             BAGCHI, LONG.

10/30/86 NOTES ON TELEPHONE CALL WITH DOUGHTY AND DOOLITTLE.

                     ]!/03/86  DISTANCE OF INTAKE AND DISCHAPGE STRUCTURES FROM MASS.

COAST. a J MAPKF.Y/NRP. 11/.10/06

                                                                                ---_-_-----------------__-___x-;

SEABROOK STATION Engin2Gring Offica October 31, 1986 W b We SBN- 1225 New Hampshire Yankee Division T.F. B7.1.2 United States Nuclear Regulatory Commission Washington, DC 20555 Attention: lir. Steven M. Long, Project Manager PWR Project Directorate No. 5 Division of PWR Licensing - A References (a) Facility Operating License NPF-56, Docket No. 50-443 (b) USNRC Letter, dated October 8,1986, " Request

,,                                      for Additional Information for Seabrook Station, Units 1 and 2, Emergency Planning Sensitivity Study", S. M. Long to R. J. Harrison (c) USNRC Letter, dated October 23, 1986, " Request for Additional Information for Seabrook Station, Units 1 and 2, Emergency Planning Sensitivity Study", S. M. Long to R. J. Harrison

Subject:

Response to Request for Additional Information (RAIs)

Dear Sir:

Enclosed herewith are the majority of the responses of the Requests for Additional Information forwarded in References (b) and (c). Attachment A identifies responses that are included in this transmittal. Attachment B is the responses. An additional submittal addressing the remainder of the RAIs i will be forthcoming in the near future. Very truly yours, am O* John DeVincentis Director of Engineering Attachment cc Atomic Saf ety and Licensing Board Service List ~ / l J')f(; t f* Iy Director, Office of Inspection and Enforcement \ d l United States Nuclear Regulatory Commission Washington, DC 20555 [' f ',j j) gp7 E/ % P.O Box 300 a Seobrook,NH 03874 . Telephone (603)474 9521 1 . - _ - . _ . - .

Dicn3 Curran, Esquire Petcr J. Math;ws , Mayor Harmon & Weiss City Hall 2001 S. Strcot, N.W. Newburypsrt, MA 01950 Suite 430  ; Washington, D.C. 20009 Judith H. Mirner l Silvergate, Gertner, Baker ,

 ,          Sherwin E. Turk, Esquire                        Fine, Good & Mizner Office of the Executive Legal Director           88 Broad St.                         l U. S. Nuclear Regulatory Commission              Boston, MA 02110                     l Tenth Floor Washington, DC 20555                             Calvin A. Canney City Manager Robert A. Backus, Esquire                        City Hall 116 Lowell Street                               126 Daniel Street P. O. Box 516                                    Portsmouth, NH 03801 Manchester, NH 03105 Stephen E. Merrill, Esquire Philip Ahrens, Esquire                           Attorney General Assistant Attorney General                       George Dana Bisbee, Esquire Department of the Attorney General               Assistant Attorney General Statehouse Station #6                            25 Capitol Street Augusta, ME 04333                                Concord, NH 03301-6397

• Mrs. Sandra.Gavutis -

Mr. J. P. Nadeau

.          Chairman, Board of Selectmen                     Selectmen's Office RFD 1 - Box 1154                                 10 Central Road Kensington, NH 03827                            Rye, NH 03870 Carol S. Sneider, Esquire                        Mr. Angie Machiros Assistant Attorney General                       Chairman of the Board of Selectmen j          Department of the Attorney General               Town of Newbury One Ashburton Place,19th Floor                   Newbury, MA 01950 7

Boston, MA 02108 Mr. William S. Lord Senator Gordon J. Humphrey Board of Selectmen U. S. Senate Town Hall - Friend Street Washington, DC 20510 Amesbury, MA 01913 (ATTN: Tom Burack) 4 Richard A. Hampe Esquire Senator Gordon J. Humphrey Hampe and McNicholas 1 Pillsbury Street 35 Pleasant Street Concord, NH 03301 Concord, NH 03301 ( ATTN: Herb Boynton) Thomas F. Powers, III H. Joseph Flynn, Esquire Town Manager Office of General Counsel Town of Exeter Federal Emergency Management Agency ( 10 Front Street 500 C Street, SW Exeter, NH 03833 Washington, DC 20472 Brentwood Board of Selectmen Paul McEachern, Esquire l RFD Dalton Road Matthew T. Brock, Esquire Brentwood, NH 03833 Shatnes & McEachern 25 Maplewood Avenue Gary W. Holmes, Esquire P. O. Box 360 Holmes & Ells Portsmouth, NH 03801 47 Winnacunnet Road Hampton, NH 03842 Mr. Ed Thomas Robert Carigg FEMA Region I Town Office 442 John W. McCormack PO & Courthouse Atlantic Avenue Boston, MA 02109 North Hampton, NH 03862

ATTACHMENT A Responses are included in this transmittal for the following RAIs: 1 12 23 41 55 69 2 13 24 42 57 70 3 14 25 43 59 71 4 15 26 44 60 72 5 16 28 45 61 73 6 17 33 46 62 7 19 34 49 63 8 20 35 50 64 9 21 40 51 67 10 22 53 68 11 Responses to the following RAls will be forthcoming in an additional submittal: 4 36 54 18 37 58 27 38 65 29 39 66 30 47 74 31 48 75 32 52 i

O S e ATTACHMENT B RESPONSES TO REQUESTS FOR INFORMATION 1 .i

RAI 1

Describe how the overpressurization calculations made by SMA were checked or design reviewed.

                                                                      +

RESPONSE 1 All SMA calculations on the Seabrook overpressure capacity were performed by Dr. Ralf Peek, currently on the Dept. of Civil Engineering f aculty, University of California Berkeley. All assumptions and methods of analysis were reviewed by Dr. D. A. Wesley. Numerical checks of calculations for the cepacities of the controlling f ailure modes were conducted by Dr. Wesley and other members of the SMA staff. The SMA report was reviewed by Dr. Torri and the SSPSA technical review board. These reviews were conducted and documented as part of a QA program that is described in Section 1 of SSPSA. t In addition a numerical check of all calculations and an independent l review of assumptions is currently in progress. This independent review is scheduled for completion on November 25th,1986. 1 -, - ..,r- 9.,-. v.e-.-m-m.r- - ,,-w * - - *- ,---+--t ----- - - -

f RAI 2 , A meeting should be arranged with the originator of these calculations to assist 'the BNL reviewers in following these calculations and under-standing the assumptions.

   /

i . RESPONSE 2 Assumptions need to be made in all engineering calculations. The dif f erence between the usual design calculations and this prota-

                                                                                         \

bilistic evaluation of the containment integrity is that here the median ultimate strength needs to be estimated. This requires evaluating the load carrying capacity in a limit state condition in which integrity of the liner is lost. In most cases this involves large inelastic deformations and extensive redistribution of stresses. This requires different assumptions compared to design code calculations, and these assumptions must be based on an understanding of the behavior of the structure in its ultimate condition. The intent of the calculations is to document in a conventional manner what was done and what assumptions were made, rather than to explain in detail how the structure is expected to behave, and why the various assumptions are justified. Dr. Wesley and Dr. Peek participated in a detailed discussion of the calculational methodology with your staff on October 16th and 17th at BNL.

RAI 3

Document the basis for the assumptions in the calculations. In particular, explain the uncertainty factors assigned to various pressure capacities. RESPONSE 3 The codes and standards used in. design assure large margins of safety beyond the design and test pressures. However ultimate capacities are not computed in design. Virtually all design calculations are limited to elastic conditions whereas ultimate load response involves large nonlinear ef fects. Consequently, only limited data are available to quantify the uncertainties associated with many areas of the calculated ultimate pressure capacities. Uncertainty is introduced due to inexact knowledge of the structures' material properties as well as the structure behavior at extreme loads. Where data exist, as for instance on material test properties for Seabrook, it is a straight-forward procedure to introduce it into the overall variability for a given failure mode. Other areas such as expected accuracy of analytical methods are estimated based on the judgement of experienced engineers, and are so noted in the caleplations and the report. In addition, as noted in the response to question 1, the bases for assumptions will be independently verified.

RAI 4

Explain the mechanism for transferring the load f rom the penetration sleeves to the containment wall, inparticular, the equipment hatch, when subjected to high strain conditions. Explain how the rebars around the penetrations,were assessed to assure that they can resist these loads in addition to the primary pressure induced loads. ANSWER 4 At the ultimate containment. pressure of 216 psi, the force per unit length along the circumference of the equipment hatch penetration i that must be transferred from the sleeve to the containment wall is l 18 kips /in. The 13/4 in x 10 in annular plate shown in the calcu-l lations (Fig 14.5, Section 14, Part I, as reproduced from FSAR report), is more than sufficient to support this load, thus preventing a blow-out of the penetration. ,' A typical calculation (.emonstrating that osall diameter penetration sleeves do not punch thru the containment wall at ultimate containment capacity will be provided with the response to RAI 31. l l ~! ( l

RAI 5

                                                                                                 \

The calculations use a rebar ultimate stain value of 4.7%, i.e., more than 21 feet of linear extension for the hoop bars. This linear exten-sion under the high pressure load will be accommodated by formation cracks in the concrete totaling approximately 21 feet in width. Justify , the assumption that the pressure loads will be carried proportionately I by the linear plate and the rebars (similar to the elastic condition) in this highly cracked condition. Also address the potential for developiong a crack large enough for the local extension of the liner plate to lead to its failure at that point. 4 1 l l RESPONSE 5 Since the strain is distributed in the steel and concrete between the cracks, the total elongation is not all accumulated as open cracks. The reinforcing steel assures the cracks will be essentially uniformly distributeo and the crack growth controlled until failure. Cracking of the concrete and strains above yield in the liner and reinforcing steel assure that the local discontinuity elastic stresses are proportionately less important at internal pressure which result in a stress condition above yield. Both the reinforcing steel and liner are ductile materials with flat stress-strain curves in the range of ultimate strength. This assures that brittle f' allures will not occur in isolated elements and the proportional strengths of the rebar and liner can be relied until failure. The potential for significant liner strain concentration in the vicinity of concrete cracks is considered highly unlikely. The surf ace of the reinforcing bars is deformed in order to create

a bond with the concrete whereas the liner is smooth and is in .

contact with the concrete on only one side. At the failure f pressures of interest, the required coefficient of friction between the liner and the concrete necessary to achieve the same concentration of strain in the liner as in the reinforcing steel is not considered credible, even if the thermal strains in the liner are neglected. Inclusion of the thermal strains will increase the required coefficient of friction and for the liner to f ail prior to the rebar significantly higher strains must accumulate in the liner since it has higher elongation. t

                                      ,-..._._m.,.            _ , , . , . - .

RAI 6

Was compatibility of strains in the rebars and the liner plate satisfied in the calculations? For example, the outermost hoop bars will f ail before the inside bars and the liner plate reach their respective ultimate strengths. Was this fact reflected in the calculations? In addition how is the biaxial stress-strain state of the liner plate considered. RESPONSE 6 The controlling modes of failure in the concrete structure occur due to membrane tension. For these modes, all reinforcing bars in a given direction have essentially equal strain at f ailure and the outermost hoop bars will not f ail before the inside bars. The inner bars are likely to have somewhat lower strengths due to the higher temperatures toward the inside surface of the containment. However, the strain at which the ultimate strengths of these bars is attained is not significantly altered for temperature increases up to 7000F. The bi-axial stress-strain state of the liner was considered in the calculations. A f actor of 1.73 was computed to account for the expected change in elongation properties for the liner. l

RAI 7

The combined tension, shear and bending ef fect at base and spring line levels was not considered in the calculations (Ref. p. 35, assumption 6). Verif y that the combined effect does not change the conclusions of the analysis. I RESPONSE 7 Bending and shear ef fects occur not only at the springline, but wherever the hoop reinforcement content changes. However, these stresses are secondary in nature. This means that these stresses are not necessary to carry the internal pressure loads, they merely arise from displacement incompatibilities in the membrane solution. It was judged that such displacement incompatibilities could be accommodated if necessary by plastic rotations of the containment wall (flexural yielding due to meridional bending stresses) without significant loss in the load carrying capacity of the containment wall. The hoop bars especially, which are the most critical for carrying the membrane stresses, are unaffected by these plastic rotations. Furthermore, the plastic rotations limit the shear forces which are developed. This is important because whereas plastic rotations are not expected to affect the integrity of the liner, the effects of extensive shear cracking could be more detrimental. The possibility of failure due to secondary shear stresses was considered for the junction between the base slab and the containment wall. This is the location where such secondary shear stresses were judged to be most critical. The estimated median pressure capacity for this failure mode is 408 psig as indicated in Table 3-3 of the SMA report. There-fore failure modes involving secondary stresses in the containment wall and dome are not considered critical. l

l

                                                                                                                       )

RAI 8

Since 31 cadwelds out of a total of 169 test samples failed at a stress l lower than the rebar ultimate strength and there was apparently a construc- i tion problem concerning staggering of these welds, provide justification for not uring a reduced ultimate strength for the rebar. 1 RESPONSE 8 f There was no construction problem involving the staggering of cadwelds in the containment shell. Cadwelds are distributed throughout the struc-ture as specified in design. Failure of the containment shell in the controlling hoop direction is  ; expected to begin with vertical cracks forming in the concrete. These cracks are expected to initiate at locations of vertical, rolled section steel supports for the liner which are spaced uniformly approximately 20 . , inches apart around the circumference. This results in approximately '

           .            260 vertical cracks. As the pressure is increased, these cracks wil con-tinue to op'en until the ultimate c'apacity of the combined reinforcing bars                   '

and liner plate is reached across the crack with the lowest capacity. The location of this crack cannot be predicted since the available information predicts essentially equal probability of failure of any crack location. However, it is important to realize that failure does not occur due to failure of a single bar. The reinforcing bars and liner develop a ductile system with the ability to provide significant load sharing and load redistribution between the liner and a large number of reinforcing bars. An I estimate of the number of reinforcing bars over which failure of a single l bar is averaged can be obtained by multiplying the six equivalent bars across the shell thickness by the meridional length required for the pertur-bation damp out. For the shell in the post yield condition, this length is conservatively estimated at greater than twice the shell characteristic length, or over a hundred bars. 1 The cadwelds are staggered throughout the structure and there is no way to establish the location of those cadwelds with relatively high or low capaci-ties, just as there is no way to establish the location of the crack with the lowest median rebar strength. From the cadweld test data available for Seabrook, less than 20% of the cadwelds can be expected to have capacities below the median reinforcing bar strength, whereas by definition, 50% of the reinforcing bars will have strengths below the median. Hence, the test data shows that a conservative value of the median strength was used. The relatively few cadwelds which may be expected to have lower capacities than the median reinforcing steel are only slightly weaker (less than 10% reduction). This corresponds approximately to the lower bound reinforcing steel strength so that the minimum cadweld strength and the minimum rein-forcing steel strength may be expected to be approximately equal. Away from the crack, the stress in the rebar decreases due to the concrete bond and , the effect of the cad-weld strength does not influence the strength of the bar once the load in the weld is decreased to less than the strength of the rebar across the crack. l 1 l

Even if the cadwelds were completely ineffective, the total reduction in the hoop capacity would be less than 1% due to the averaging effect of the adja-cent bars. Because of the limited number of cadwelds, there is very little probability that more than one cadweld will be located in the same crack in the area where the averaging effect occurs, and an even smaller probability that two cadwelds with low capacities would be so located. Thus, the effect of a few randomly distributed cadwelds with less than the median reinforcing bar strength is considered neglibible, and this effect is accounted for in the variabilities associated with the various structural failure modes. r

RAI 9

The containment analysis is based on an axisymmetric geometry and loading. This is not the case due to the presence of adjoining structures such as the f uel building and main steam and f eedwater pipe chase. Identif y these axisymmetric conditions and assess their impact on the failure modes and analysis. RESPONSE 9 The effect of local non-axisymmetric conditions is not expected to effect the capacities computed for the axisymmetric failure modes (i.e. cylinder hoop and meridional membrane f ailures, dome membrane f ailure, and base slab bending and shear f ailures etc.). This is because the local ef fects damp out rapidly for the inelastic case. Local f ailure modes such as interference between the f uel storage - building and the containment were evaluated and found to have significantly higher capacities than the controlling axisymmetric f ailure mode (hoop failure). Also refer to response to RAI 13. l i

RAI 10

Only (X-23,a X-26 sampleandofX-71). pipe penetrations are considered in some detail should be provided. The justification to consider only these RESPONSE 10 staff and members of the Yankee Atomic staf f. Virtually all the pipi The purpose of this i review was to identify the lines considered most likely to fail, based on support spacing from the penetration both inside and outside the containment. Based on this review " worst case" lines for a multiple pipe penetration, a thin wall (sch. 40) pipe penetration, and a thick wall (sch.160) pipe penetration were selected.

                    " worst case" lines indicated that failure of the penetratioEvaluations                     n (i.e. of these breach of liner integrity) was not likely to occur for the thin wall and multiple pipe penetrations, irrespective of relative dis-placement, could       occur. although                  fluid leaks or flow restriction in the pipes Thus, any further investiga this type of penetration was unnecessary. tion of lines associated with The evaluation indicated that sufficient potentially                failforce     could be generated in a thick-walled pipe to the penetration.

However, there are only a few schedule 160 pipes entering the containment and the expected leak path area associated with this type of f ailure is so small that even if areaall schedule is insufficient 160 pipe penetrations should f ail, the resultingeak l to prevent a continuing increase in internal pressure with eventual f ailure from an independent f ailure mode. Therefore detailed evaluations of individual penetrations were not warranted, once the contro11oing penetrations were identified.

RAI 11

A structural evaluation of electrical penetrations should be provided. ~ RESPONSE 11 The electrical penetrations were reviewed and it was determined their pressure capacity would not be a controlling mode of f ailure for the following reasons: penetration are not subject to rigid pipe reaction which interact with~ containment wall dispacement it was judged that the failure of these penetrations is dominated by thermal effect (i.e. leaks) a thermal analysis was performed in lieu of a structural analysis Consequently, a detailed evaluation was not conducted in order to concentrate on the more likely modes of structural failure modes. This is consistent with the overall approach used for both the seismic and overpressure capacity evaluations where detailed investigation is only conducted on conceivable modes of failure.

i a

RAI 12

The basis for the leakage area assigned to the flueo head at failure should be provided. 4 RESPONSE 12 The leakage area due to failure of the flued head is a rather uncertain quantity as indicated by the large variability assigned to it. As a median centered estimate, the flued head is expected to fail at a radius 0.5 in less than the inner radius of the sleeve. With a slight increase in containment pressure and the associated radially outward displacement of the containment wall, the part of the fluid head which remains attached to the pipe is pulled into the sleeve. As a result a clearance of 0.5 in is available all around, as a leak area. The logarithmic standard deviation of 0.6 indicates that the 95% confidence intervals for this clearance width are about 0.2 in and 1.3 in respectively. Thus no claim is made that this leak area can be accurately predicted. This uncertainty is taken into account in the probabilistic consequence analysis.

RAI 13

A more detailed evaluation of the impact of punching shear at the Fuel Transf er Building should be provided. RESPONSE 13 Punching shear stresses at the Fuel Storage Building begin to develop only at a (median) containment pressure of 172 psi. This pressure, at which the containment wall begins to bear against the Fuel Storage Building, is not "very approximate". The main source of uncertainty for this pressure is the uncertainty of the effect of bonding on the stress-average strain relation for the hoop reinforcing bars, and could not be eliminated by a more detailed analysis. What is "very approximate" is the existing evaluatJon of the increase in containment pressure that could be supported af ter the containment pressure begins to bear against the Fuel Storage Building. The large uncertainty in this increment in internal pressure is reflected by the large logarithmic standard deviation assigned to this quantity, and has been included in the risk evaluation process. In doing so, it is found that the failure mode is not critical. This conclusion is not expected to change even if a more detailed analysis resulted in a slightly different value for the median pressure increment or slightly reduced uncertainties for the punching shear f ailure. Furthermore, even if some of the uncertainty in modeling could be removed by a more detailed analysis, the uncertainty in strength would remain. Finally the lack of knowledge of the behavior of construction materials under such extreme loading condition renders the applicability of the results of the most sophisticated analysis open to debate and is not warranted for this evaluation.

RAI 14

Clarify the extent to which double ended piping f ailures have been considered in the overall containment performance assessment. Provide isometric drawings of all piping attached to containment penetrations. _ I

                                                                                                                                                          \

l t RESPONSE 14 The possibility of a pipe failure on the inside and outside of the containment due to the pressure induced displacements of the containment wall was considered. For some of the thick walled pipes, it is likely i that l load on thethe penetration pipe. Forwill fall prior to the pipe and thus relieve the thin wall pipes, the most likely mode of f ailure is bending of the pipe thus reducing fluid flow but not resulting in a leak. However, there is some possibility that fracture of thin walled . pipes can occur, and a much reduced possibility that a fracture on both sides of the penetration can occur for a given pipe. This possiblity was considered for Penetrations X-23, X-26, and X-71 and found to have ! a negligible contribution to the overall risk. For the flued head pipes (Penetration X-8) however, the double-ended pipe break f ailure mode contributed about 55% of the total probability of this type of penetration f ailing before the concrete st ructure membrane failure mode. Isometric drawings of all piping attached to containment penetrations were supplied to BNL in September 1986. l

   -r f * - - - - '-- --     v  - -------~'=*-'e*-         - - - - - - - --         1 +-- - -- ----- -- - - - - - - - - - - - , - - - - - - - -

RAI 15

In PLG-0465, page 2-10, Figure 2-3, the conditional f requency of exceeding whole body dose vs distance apprears to be driven by the S2 source term. If this is the case, please describe all accident sequences (internal and external events) that contribute to the frequency of the S2 source term given in Table 4-2, pg. 4-7. In particular, define how the timing and size of containment leakage was determined for each of these classes of accident sequences. Justify the appropriateness of the bounding of each of the accidents into this particular source term. RESPONSE 15 Our response to the overall question of how the contributors have char.ged from the SSPSA to recent updates, is comprised of the response to question 23 along with the following additional information: ASSIONMENT OF SEQUENCES TO RELEASE CATEGORY S2 i i As discussed in Section 11.6.3 (startingon page 11.6-5) of the SSPSA (PLG-0300), the release categories S2, S2 and 52V were originally defined t in the SSPSA to cover a class of accidents not modeled in the Reactor i Saf ety Study (WASH-1400). Of these, the category I2V was found to be a significant risk contributor and was found to dominate the 200 rem dose vs distance curve in the EPZ sensitivity study (PLG-0465). Note that for simplicity, the release category notataion was simplified in the sensitivity study. Hence S2 on the sensitivity study is actually the same as 577 in the SSPSA. l l The S2V release category was first defined to bound the releases that could occur as a result of small penetration f ailure during hydrogen burn pressure spikes that would occur shortly after the time of reactor vessel melt through (see discussion on page 11.6-12 of PLG-0300). Then

• to avoid an excessive number of release ca Jte ories (to 14), additional sequences were conservatively assigned to 52V (now S2) as well as including those in the 3FP and 7FP plant damage states (as well as some steam i generator tube rupture sequences). As noted in the response to question 23, it turned out that the 3FP and 7FP sequences fully dominated 52V.

The 3FP and 7FP sequences are dominated by station blackout RCP seal l LOCA sequences in which the release path is the 3 inch seal return line with failed open motor operated valves. There are many different sequences in but they all have the same release path. As noted above, the dominant sequences are initiated by seismic events partly because no credit was taken for operator recovery (closure of the outboard MOV manually) af ter seismic events. As discussed in Section 11.6.6.4 the assignment of seal return path sequences to II7 is conservative becuase the actual leak l rates would be much less than calculated for this category.

  , ,m. _., .__. _ . . _ _ _ _ - - - - - - - - - * ' - " - ' - " ~ ' - ' " " ~ * - * ' ' * " ' " " - * " ' ~ -^

RAI 16

Provide justification for the liner yield stress increase from the specified yield stress of 32 kai to a 'n yield stress of 45.4 ksi. RESPONSE 16 Attached are liner plate certified materials test reports (CMTR) representing ten distinct " heats" or " charges" of liner plate material. These samples were selected at random. The average yield stress in this ten charge samples exceeds the mean yield stress of 45.4 ksi which was utilized in the analysis. l l l

1 i! ,qumen - . - --,- - ,,,Jmys.3dMIIgal IFILE 810. : COImlACT IIG. : st m MlZUSHIMA KAWASAKI DORI

                                                           @uen sua camnon munmuess mm                                                                                   : , As CERTIFICATE
        ~              """^*"'*"^""                                      INSPECTION MAscn 25.1975 4AME OF CLIENT

• 1.M 1. TOILOBI ANA NO.3180E5. DATE OF ISSUE:

ASH _F &A316CM.60HT H. 3 ml STSE CO. MATEHIAL : Hmte 25.1975 , NJKT P T . DATE OF TEST : _ wsa No. 5501-yt2 CHARPY IMPACT TICST sesc Mo.  : 382E5ucxict TEMSit.E TESDC4.= 8" wa MWWtDESS may E s. OF Ti ST BEND mich : RD4 ARK' Y . P. T. 3. Min. f KLI I l* N t i 180* Ave. t i WEIGHT At i Mk Ed Man. R-l DIMENSIONS h 60.0 3 a ! CHAllCE MO. ROLL NO. 23 . 0.5 2 _ (mm) (ha) 32 0 -78 0 i 64.6 31 coon 9 60 x 3,070 m St.20$ 2,992 46.3 3-1130 spo46A - ( ss.cs.q-s) I

                                                                                                                                                                                                        \

e s1t9 CM" HEAT TREAD 4Dir : CHEMICAL ANALYSIS (%) se. Cu Ni Cr Y Cg C' Si W P S m use = sone = no w seo = no m esse u nse ,, , ,

                                                = nao = ice  = soo                                                                                                                            "8""' "

CH 2CE NO. Mas. Max. Max. E*8**18888***8058 F A' 8> Max. 13 83 ' 12 15 40 21 34 121 14 7 2-1110 30 k r.4 e 3 .. ~ . . . .

                                                                                                       -       .     ....... m_.u ,,,,. w,                     -        .

n, _

                                                                                                                                                                                  ..       "- ' > v w,, , .
                                                                                                                                                                       .:.                ge uwA sm. mamm munnu-                                                                             CONTRACT IllL :stm                             "

MtZU$lil\lA KAWASAKI DORI WRA$lilKl JAPAN NAME OF CLIENT I !!.I. YOWil4NA letD Wul&S. h.OJe a. T PITT. s .;ll-Dr.:. DUIllE.S 811I1. CO. MATERIAL : ASIS SA516Gft.&1MT DATE OF ISSUE: 88m 3 5. 39P5 _

,.r.. . i b..II aae su , :xi.: y t-rte DATE OF TEST : NAACIf 33, 1975 uu.. :.o. : ys ;. .rw..t.

                                                                                                                                                                                                                               ~

TENSiti 11.STICLgee smE M MS Cll Altl'Y IMl'ACT TEST Y . P. l T. S . E I. OF TEST BEND Notch : wenV

                                                                                                                         '      e%:     MA      e         ,   180=       Ave.              M ui.             REMARKS CIIAl(CE Nt). l(111.1, Nti.                      OIMENSIONS                WElGilT

• Est tlin 60.0 Miss k us Man. R- At t i i

                  !                                                     (snel                  iLg) 32.0         -72.0          21                            05t         i       2        3      A.e.

2,592 46.9 65 7 coup 2-17 fst. 20::f.9A 9 60 m 3,070 x 11.205 31 (su-c2-9-1) .

                                                                                     .                                                                                                           Mut litvitate g 4 W h e.

f-~~ ~Wgli ~ \ CHEMICAL ANAR.YSIS (96) HEAT TREA'iL=riT : Si Mn P S Cu NI Cr te. V C, oe C w see v one = nem = see one

                                                    , . ,l'* . im   see   = iam . i. . .. . ie CHARCE No Mas Ma s. M.s t                                                                           Plates Mat 13        83                                                                                                 Igorimm18 sed...... 3635*P Ser SS was.=
                       .             . . . .        :. t. -20_ . t an      35 '. .) . .        __

l 10 22 10A 12 8 . .

-17tt*.

l 6 _ s's/ O s j

             #"#          #                                               We Imereby cer:6f., einst she seneettet deserted herets has                                                                                    .
                                          /                           been omasic in accordance oesh ebe raise of . h                                                                    [              [h
                                                                 .W

                                                                                              %                                                   ..x
                                                                                       @wN                                                                                              .n : m,)

um mam e. s ' CWNBACT 18.: ate-sess surusumeA nWASAKI DOM

                                     "^$"""^"^*

INSPECTION iCERTIFICATE mmi.  : . . gegg g mm . I .ii.n. mAuiana no.3 unes. - .

                                            ' ;1t': .'.;:t'af,'f;" ***' '*>-                                . am. :        A-= = = =. = r                                   om or is==:          >====    v. = s
                 ~ :..                      ,ss,,,_v, FEEMisaar 7,1975 DATE OF TEST :
                 . .a 2 .:4.                  3n ._ S ::

mesats itsna, a === nett namusu cnanry aseAct Tusv Es. W TEST BE3G Nee 6 : um# l CIIAAE NOL ROLLNO. DtMENSIONS WElGilT v.t e.] r. Is. (sp AREA , , tage Ave. Men. RBSARK.C l tan 60.0 Mk dh Mem. R- At t I i (mm) (n,r} 32.0 =72.0 21 05i a a am

-2M Enon s 9.60 m 3 o p u 13.2D5 2,592 47.6 69 1 24 C005 j

I (sm-C2 9-1) is 9.60 m 3.070 m 1a.205 2.392 ( m-cz.9-1) 8  %, % , L*$ - CHI.*MICAI ANALYSIS (95) HEA1' T1tEATasswi : OtABCE Na C St Den P S C. Ni C, u. V

                                                                                                                                             ""*                  Cg pg ,g , ,

St . % 15 35 blam. *. ; st. hIAE. . t --.-p -33e --35 *l

                                                                                                          -  -                                                            Normalised.. 3652*F fler m stanstem.

I

m rn.: m: %> 4s KI STEEL CORPORATION MlZUSHNA cowraxT wo. : 817 M oose

                    .nei si nn x*w w.ni Imni
                     .                                                                   %./.     . KAD si."'"'"                                                                                                                                                                                              '===

INSPECTION CERTIFICATE Som.itii  : .

      ., a . 3 . .           ,gg.,                  I . I . '. . ... .".'.' :A h\) . ) s..               .

- .i I: 1 .. J :;. l. Co. Asag 34.,y ,gg ,,,3,gg N 87.

m _ ,p g( , DATE OF ISSUE: Y 27, DATE OF TEST : ,_

                       . . . .                    5           ,        1
                                                        ~

Il NsIl r. It % fit;t.- 13" pues RfDC }{AltlNfE55 Cil A14I*Y Ihtl' ACT TEST v . I* . i.s. se l IF TEST !!END Notch

• s.unY t il\1l .NS it >NS l u l.ls;!IT ' I'] I ' ' *8'
• AftEA , , gepy s g ,. . Mw REMAlt

                                 .. it. .I 1. *.. .                                                                                                U 8.0._                                                              t l

kle. Mis _ Mus Man. H- __^e . . i l

                                                                                          ...'mm'                       , Li:]      , 't.I. J ,, . 7:f.d             :21          _,     __        .

_01_i i 2 3 A ..

                                                          ).s,) x 3..s7J x 11,233                                                                   67 0          29                               CooD 1 .,        .       ! l .1.i ...                                                                  2,5. e.!
                                                                                                                          ,             47 6 i                          (. -c..        -1)

I I

l l g giMt Mf'N g 4'7fIf

e. i s._

                                          ' ~ ~ -
                                                                                           ~~

iWi!IdlC AL ANALYSIS (9(I HEAT TREATMENT b

                                                                                                ~

i~ ' ' E'u Ni Cr w V Plahs

                    . <, . . . . l'
                                                             ~ {h..'..        .s.e . .

Mn

                                                                                         .....     'l.Y
                                                                                                                                      ~
                                                                                                                                        .,,,,     .  ..e,      .....       ....                      C. n.,

1*ers.nll W .. 1652*F for 1.f.antinego

                                                                  ;i           g-,     ,,.j ~       51..s.
                                                                                                                     .s.  *.a * !.

_l.I;,*; ~1*.' .8

                                                                                                      ;s              .

l  !

                                                                                                                                                                                                                                                        . 7.,
                  .--         8                                      11*2%               11)         13             8                                                                                                             ,

g s.. . 1 c .7a > ,, , o ._3. ..

                                                                                                                                                                                                                                       ' m W..      l.. e . l.s       certely elsat slie seascrial aleactM herein leer                                                          ,
                                                   ,g ~, j, ' .                               l ...n     n... . l . e. . accue ilence wkle slee riules el Alte eNd                                                                /,          g,e ., o l         .7/,                                                                                                                                                         n.s no w .o.                              w-              ,,.w. ..

I _ _ _ _ _ . _ _ . . _ _ . . . _ -

                                                                                      ..                                                                           s

l 8' * *1 PEE WL: J8trFs-eaf4 Moi KAVG'KI STEEL CORPORA 110N WKS CONTRACT 100.* 2ap e g ,, MtztlSillMA KAWASAKI DORI xo As"'*' '^^" CERTIFICATE  :-- INSPECTION sum.. . hAME OF CLIENT : 1.u.I. Yosaastana No.3 vosiKs. PinsecT e PITTsmer.it. Des Motpas stssL co. MATERIAL : AsE_sA31(city _ DATE OF ISSUE: M u r.n 2 ;. n.375 SEAttaw* LINIY 1 & 2 . wesen too. DATE OF TEST : stur.it ~;, gey73 5' int -382 SPEC.MD.e )!!:$*,00001 TEMSM.E TESTICL- "" *** REDUC MARDMESS CHAMPY IMPAC r TES T Y . P. T. S . E l. OF TEST LEND Nach eenV DIMENSIONS WEIGHT t W] l iss) AREA i 180' Ave. Min. REMA{ CitARGE NO. ROLL NO. min 60.0 Mk En Max. R- At t I i I (mm) (kg) 32.0 -72.0 21 _ o.S . i r 3 1a. i ( 9.60 m 3,070 m 11.205 2.592 46.5 68.6 32 caos l t-3551 2Dzpta ( ss-cz.9-1) 1 00MW;NI 3IU W A W ts

i. - . , ;p CIIEMICAL ANALYSIS (96) HEN' TREAD 4Drr :

C Si Mn P S C. N4 C,

                                                                                         =*

ne. V Cg Mate CHARGE N0.

• Max. 1i5 55 Max. Mas Man. go ,,,.3 g a, .,3652 F for 3D at an. t o at -30 -130 35 40 s.3551 11 33 to6 14 7 , .. . ,,

N1 Tc *'. o x / O t. o We hereby certify shot she senteeinI described herein has _

f. Q. ,

g p' jff, g *g been snade in N:h imish the teams of she 35WNB988s. [ cusyff y N >-

      '-                                                                                                           _ f assa sesamm e

• MlZUSHIMA KAWASAKI DORI E*M N KURASHIKI JAPAN INSPECTION CERTIFICATE somi.  : , -

NAME OF CLIENT - 1.1J.I. Yan Ann 300 3 vois:s. i n:oRx:7 : PITrr. 2r:::ts ass peotions stuuL co. MATERIAL : M SA51%N _ DATE OF ISSUE: MCicat r$, 3 ,75 ssar xxx test? 1 a s l wr.': to.: 5,91-yts DATE OF TEST : sta c!! es s g . c :.: i.  : vmt TENSit.E TEST 1Ct.- 8= sest ' REIRE. IIM NESS CetAMPY stel'ACT TEST l Y. P. T. 3. E l. OF TEST BrJfD Needi ; sumV CHARGE NO. ROLL NO. DIMENSIONS WEIGHT t ut, I I en.o (#1 uw Rn

                                                                                                                                         %        t Man.

5 H-1e0' Ave. At t t Min. i REMARK (mm) (ha) 12.0 -73.0 31 05 i i 2 3 A 1 36o 1 ss98a 9.60 m 3.r70 x 11.3D5 s.59s 47 9 66.8 3s som 4 (ss.cs-9-1) i t I

                                                                                                                                                                                                   ====

Annetontie oralen Bise. ...8 8(Mie. tee.5) =M NIU i W1tranee ie Test. ..... ....! c .29to Di m nnia ne Isiope sties ....A easytetle ,M M"M eurteen Inspecti san. . . ....' .t"

• e M,

Y i I CIIEMICAL ANALYSIS (96) hex ThEK195 ! C Si Mn P S Ni Cr es. Ylle8.A L l = s** = ace = ses = sees = seen .Cuice = ses = ses xuse x nem . nn . _ CHAACE No. M= 15 a M= M= Mm mur.  % C ases......nss h ese a name. en .. .. s. w an to 36 ao as 111 38 8 88 38 a.34o1 C

           ~
                 / M/. G C.                                 We 'nereby cereNy thea che esseerini desseted herets has                                                                 =         O I'    o/. ,,2 e                                 - -

g(..' 4. . .

                                                                                            - ~ *                                   ~          ,4,w,,,,                          ,   %                                   '
                                            ..   ..3....,.                                                                                                                                                           ,

4.i.sie M17t*SillNIA KAWASAKI DORI

                         " " ^ " ' * ' '*""
                                                                   @ KAWSKI STEE CMPORMM MEUSWE GES                                                     CONTRACT 100.                  2139-2o003          _

INSPECTION CERTIFICATE summ  : - MAME OF Ct. LENT : 3 ** I* IC'""8AM" 88* 3 "888* resOJecT e P:TTsammenses Mortes sisuL co. MATERIAL : A98 sA5160st.6nwr

• DATE OF ISSUE: 'Wcal 21, t9q _

SE41940pK ISIIT 1 & 2 usa No. : ymin-3n3 DATE OF TEST : micti 29, im--- sitc.r.o.: 385 ywxx)1 TENSII.E Yt.NT1GI.- Ba son REIAE HARINESS Cel A MI'Y IMI'AC T TEST

                                                                                            ~T 'I' 851 Y     .l   T. 5. E l. OF isu Anry TEST BEND Imr Wich :               wenV CHAllCE NO.        ROLL NO.                   DIMENSIONS               WElGHT                     i                                         Ave.                 Min.                    REMARKS

. hin 60.0 Min in

                                                                                                                                    , Max. R-        At        tI                           i l                                                                         (.   )        (kg)  32.0       -72.0      21                         05i     i      2             3         A.

[ . l l 1-ypo6 2r885A 9.60 m 3,070 m 11,205 s.592 M.9 67 1 31 Goon l ( ss.cs-9-1) . a 9 60 m 3.070 m ine sDS s.592

• pi

( SIN:s-9-1) f p n1 KYI an

                                                                                                                                                                                   $,,~

s CHEMICAL, ANALYSIS (96) HEAT TREATMEnrT : C Si Ma P S' Cu Ni Cr 88. V Ceq

                                                     = leo = ses  = noe   w saw = w.c = ius  w see   x ses   xasse     = sees               = ias CHARCE NO.                                                                                                                                     ,

Max. 15 85 Max. Max. i 21 -30 -120 35 M.ax. o brumlined..1652*F rer 20 ministee. i 11 23 108 15 8 .,., 1-3706 pWA [ G d k,',*!IC(b _ . ,4 C / e/. O (. We lierchy certify thas the smoterial described herc.n bee i Cr p. jnMt - been saade in accordance wish the rules of she,gryengst [ k.-.mr. __ . m s . s.3 ss. , ,

                                                                     . . . wm . -~ -                                                                                                        _

                                                 !jll                                  Il                             ,    '
                                                                       ~               S                                             ,,   '

K 3', . _ 5 5- m t c 7 q9 7 R A '3 31 t e s

                                                                                                                                                                                             , a.

i 9 1 4 L' n 8 , . D W i e s

                                                                                                                                                                     ,I
        ;            8 e                        5            5 2

R c. t .N _ ._ e-t

H i

w. L S

9 C

            ~        3 t

s s I A f T S a wWw a  : T f r o T h-n gra E r T F s Vein. D *2 ( a g 0

                         .                      :            :           T m C eM 3                                     M                 5            a g           0                         E                                                                                            T                 6 1

U TS A T C . S S E P M t A E 1

                                                                                                                                                                           )              /

A E T R I I . R i. 2 T d 3 g T O N m u F O O E E F Y P M hc T A l e e L p C s

• T T A A A

I I t owt NAA a E H grNmo l e N D D C g f h s a e p i p D* N 0 5 o s n ms. E8: o i e m E b 1 R 0 G C= r e si g T S S h

      ,'                     A                                           E.ST I                    n a

d eh s e'a

       , NI DE C

• I WTf I

M 6tM r e f a s o s F I T M C U F'F H l O fn A lei o e d e e T 6 0 R n 2 s Vs= il n e u R R

• s. E W it Ma 3 s

ir r e

      .ME       T            C
                                             $S A

8 . 0 3 )

                                                                                                                                                  *sex i

s sh es e A 5 . s 6 9 eh s t 2 WF C M T i 0 7-5 6 ( S s rno hs i w M WN C O A S E j T I W 0 I S C= Y t a e ht c n - E. u P a3 5 L s a L I  : s t M3 A il e y rd E E T L A t u y 6 4 N Nx A f i t o c T C S I R Y

                                                                                                       )

g L un s o x r c e c a E E 1 T T l n 2 A C= M a A l ( C n P C 9 y i

     .$S
     .                                              M                                     I 5                                 I m       n
                                                                                                                                                               .                       b E                  2                                 M      m a0           6                   e e r

E SI= M 4 e da W A K N I s

                                        .      o c
                                                                                                       )

H C oc x 5 e a3 1 h e e m m 5 Pn t e s o b u ( m 0 2

                                                                                                                                                       =     M 1

Wene . s b t t 1 e o 0 0 i s S nn 3 1 . - 3

• N 1

x M= 51 5 - 1

                                                                                                                                                                                                       =
                                       #        s s

O I ) u2 S 01 n 0 3

                                      "         t                                          N                  7 -                                  ii        53       2 o&

N E 0. 3 >: 4 S

                                                                                                                                                        =

3 - . I

                                       "         s 1                                  M                                                          o x .

R ' sT I D u c. e O a C l M2 a1 1 1 D 'O T a. ht t s os =

6. ( ._

                        !h"  "^

• r. U i

o .

                                                                                                                                                                                             ]

3 ss n2c 9 A * ;o S^ " ms o. 0 3n o . A 1 Tu1E

                                                                - 5                         O N

W"" A K 's Ta0: I n58 Ps53 L L A 5 o m r y O s _G

                                                                                                                                                                                 .g A               T           i Mo              N                                                  R E                                                                                                                .

t . . O iw ._&

l. O N l

T. 0w g l S C b N 2 U u . E 2 7 G E 7 3

                                                                                                                                                                                 .t
                           ?.             F           J        K ktc.                           G                3                                         R                       .

l M O e O1

                                                                     -                       R A                5                                ~        A              5 E P i

W3 H - H 3 . 3 ~ C M C -i A M . g7 u { fl 1 jl jlll 1 l ,

1 I MlZl'SHIMA KAWASAKI DORI

                    """^S"'*^"^"
                                                          . . . _ _ _CERTIFICATE INSPECTION

___ C(NTRACT 110. :st3PSEIIp S=ua .; ) NAME OF CLIENT ' I.M.I. Toe 00HAMA 380.3 WOIIKS. A! Ben SAStEsn.6nwr DATE OF ISSUE: pu.E:N 25. 1975 P:nIncT FITF5nUIEH DES MD1853 STEEL Co. MATERIAL : ! SEAllll0OK UNIT 1 & 2 DATE OF TEST : ninrJI :!5. ta75 i wuftK HD.8 55nt-3ns [ SITI*. 16. 3 14'f.SmumisI TENSILE TESTICL-ga seer RDir.HAISE3S Cil AMPY 894 PACT TEST l Y . P. l T. S . E l. W TEST BEND Noech muiV l

• MS1 15, AltEA  : ter Ave. Min. REMARI DIMENSIONS WEIGIIT i i CHARCE NO. ROLL NO.

Mn 60.0 Min ist Mez. R- At t i i 05i a 3 a (snail ikg) 32.0 -72.0 21 i 1 M.4 32 C000 9.40 x 3 070 x 11. 25 2.592 48 5 l 3 632) 2C35sA j (sn.cs-9-1) , t l - ! lilli W 2.592 A..tonti , oretsi itsr. 18e.8(Mt 4.No.53 3 9.60 x 3.tv70 x 11.2 5 ..... Acceptn I.le n,#. $' p' (sa.cs-9-1) Untramar le Test. ..... ..... - Dinennta ne Isinpadiem ..... Aceartal'le siaface Inspecti sei. . ...... Acceptnl'te

        - - - - - ~

calEMICAL ANALYSIS (96) HEA T1 TEA 1 MENT: NI Cr an. Yael.A l C C la = St t** Mn

                                                          = iso P      S
                                                                  = s** t'se Ce m

• noe = see a mes =ts** maco p *1 %,

16..... 162g M m at e= CHARCE NO. new. Man. Max. 15 85 Max. Max. Max. 4D .}%._ 21 -yl -12 35 45 8 # 3I 3 6n) 11 lA 110 17 , y,

    ~~

t ;ef Q( We liereby certify that the smaserial described 1.creen has . C4 8 l,w f, l ggg, g. - been snade in aceerdassee uteh ein reise of H o w

                                                                                                                                               .                       - ,sm.m-

_ n ,. . . w cx. _ _ _

                                                                                                                                 . wi _              A                   _s

I

                                                                                                                                                                                           # y ty.5 *.?,'! ;,

nang opCuen: a..:.s. wntoen mm.3 mea .,r t s.a. . . . . :.:. -

                                                                                                                        ..; rorr=s s= za, ca.                                                       4.

64AfDUR : a s s m *' _6cour t .22 a a s - BRTE W mmE- m up. 3

               ; 3. .                                                                r t .. : '
                                                                                                                                                                                           -:jf,> .

. . . .. . .a .u. . , . . , . . . pgTE w,tsgT : aparasar m. n I imana.s Temus w= nos e==t==v mesiscy . W , .. j [ .JT,:4. Y. P. El. SM rumars iia ROI.I. 300. D84EftSatN85 WEIGHT t 8 ='3sW

                                                                                                                                                                                      ~

(#1 30 mash : ~ - muuW .. - t-t 8 i Ian .J teen' 90sse R=. lar' Asm, As ti ten ' .

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RAI 17

Indicate the correlation between containment failure sequences and the containment failure modes. RESPONSE 17 Three terms are used to describe the f ailure behaviour of the Seabrook containment. These are: o Containment Failure Mode: The term containment failure mode is used to describe the structural member of the containment system which fails as a result of an overpressure condition in the containment. For example wall hoop f ailure or feedwater penetration flued head f ailure are containment failure modes. Table 11.3-1 of the SSPSA (PLG-0300) lists the dif ferent - . containment failure modes. _ o Containment Failure Type: The term containment failure type is used to discretize the range of leak areas which are predicted for the different containment failures modes. Three containment f ailure types have been defined, namely Types A, B and C as defined in Section 11.3.1 of the SSPSA. k o Containment Failure Sequence: The term containment failure sequence is used to describe the containment failure associated with each release category. For example release category S3 is associated with late containment overpressure f ailure sequences. In the SSPSA two additional distinctions were made to characterize a release category, namely whether the containment spray system worked and whether the containment floor and reactor cavity were dry (vapori-zation release) or wet. The release categories are described in Section 11.6.4 and in Tables 11.6-1 and 2. Since all risk dominant accident sequences involved dry containment conditions without spray, the latter two distrinctions are dropped in the RMEPS study and in the WASH-1400 sensitivity study. All release categories were than based on dry containment conditions without spray and the containment f ailure sequence was the only remaining distinction for the release categories. The correlation between containment failure modes and containment f ailure type is given in Table 11.3-1 of the SSPSA. One containment f ailure type is agrigned to each containment f ailure mode. The correlation between the containment f ailure sequences (or release categories in the RMEPS and WASH-1400 Sensitivity Study) and the containment failure mode is described below. o Release category S1 model an early gross containment failure sequence. It is equivalent to PWR-1 in WASH-1400. It is always treated as a gross (Type C) containment failure. o Release category S2 models containment f ailure sequences with an early increased leak rate. It models a Type A containment l

                   .        .                   .       -- -            =          _  _    - _ _ . _ -- -

t 4 failure occuring at the time of vesselbreach. From a containment f ailure pressure point of view, the S2 release category is more conservative continues until a late overpressure failure mode occurs as in release category S3. o Release category S3 represents a late overpressure failure sequence.

It is modeled as a linearly increasing leak rate beginning at the time when the first type A failure mode is predicted to occur and building up to a full type A f ailure mode (6 square inches) at the time when a type B or C failure mode is predicted. This is followed by a type B or C f ailure type at that time. Howeve r ,

because all S3 accident sequences had calculated containment failure times longer than 24 hours they were always treated as Type C failures. The lowest containment failure time estimated in the uncertainty analysis was 22 hours and the type C failure - assumption was retained even for this case. o R'1 4 ease category S4 represnets basemat melt-through failure sequences. Because of the uncertainties in the release fractions for basemat melt-through at a rock foundation site, all basemat melt-through failures were conservatively modeled as an S3 release category. o Release category S5 represents accident sequences where the containment remains intact. A cont $r *ous release at a rate of 0.1 v/o per day was conservatively simu;a'ed as an equivalent instantaneous release at the time of vessel breach. o Release category S6 represents accident sequences where the containment is ot isolated from the beginning. It is modeled as the largest containment penetration which is allowed by - technical specification to be temporarily open during normal operation and which communicates directly between the containment atmosphere and the environment. This is the 8 inch diameter online purge penetration with a flow area of 50 square inches. Thus in size it corresponds to a pre-existing Type B containment failu:e. o Release category S7 represents containment bypass accident sequences. It is modled as an RHR pump seal f ailure with a combined leak area for both RHR pumps of 2.6 square inches. The frequency of the traditional V-sequence failure mode (RHR pipe rupture) is included , in release category S1. In summary, it is noted that the analysis did not depend on the distinction between a type B or a type C overpressure containment failure, except that the inclusion of type B f ailures shif ted the containment f ailure probability j distribution to lower pressures compared to the case where the type B f ailures are not considered. i

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RAI 18

Provide the basis for concluding that the sight glasses in the hatches will not fail under high. containment temperature and pressure conditions. ANSWER 18 The sight glass in the personnel hatch was tested by its supplier, Owen Corning Co., under the following conditions: Pressure = 150 psig c Temperature = 5500F In addition the pressure was cycled from 0 psig to 150 psig ten times at a constant temperature of 5500F. The Owens Corning data sheet is attached. We are currently persuing discussion with Corning Glass to determine if any testing has been done above these valves. t e

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RAI 19

i Document tha effect thct tha recent update in se;issic f ragilities will have on the conclusions of the PSA results. . RESPONSE 19 b* Seismic sequences dominate release categories S2 and S6 in the Risk Management and Emergency Planning Study (RMEPS). The response to question #23 discusses the principal contribution to early health risk and explains how these release categories contribute to early health risk. The following explains how the seismic fragility update

!                         is' expected to change the frequency of S2 and S6; it is expected that the ef fect on the frequency of all other release categori'es will be insignificant. A complete requentification will be includ' ed in the probabilistic safety assesment (PSA) update now in-progress and planned for completion'in 1987.                                                      '

ma In the complete performed using the seismic risk analysis, a point esEisate analysis is first , plant event trees that are qu'antified for several

              -           discrete values of ground acceleration. From the point estimate results, dominant. sequences initiated by -seismic events are identified; then, these sequences are reanalyzed using a computer code called SEIS4. This code is described in the SSPSA Section 4 and 9. In $Elp4, the seismicity curves and f ragility curves are appropriately combined and uncertainties in these curves are propagated to obtain uncertainty dfstributions on the final result, which is either a core melt orfplant damage state frequency contribution.

In the following approximate _agalysis, the point estimate step is bypassed, so some assumptions ske made about dominant sequences. Hence, these results are only rough ~ approximations ) and should only be used for order-of-magnitude estimates. Again, a complete reanalysis of seismic events is currently in progress and is planned for completion in 1987.

                                                                                                                    . i d<

1.1 RELEASE CATEGORY S2 E #'

rs*:c This release category is dominated by earthquake and trahsient initiating events. These sequences can be simply represented as OG(DT + DG + SSPS)

                                                                                                                      .: e- (1) where OG                  = Offsite Power Fragility                                                  .

l DT = Diesel Generator Day Tank Fragility DG =. Diesel Generator Fragility '# SSPS

                                                   = Solid State Protection System (SSPS) Fragility (actually 120V AC power panel required for SSPS success) and only seismic unavailabilities are included.

j Also, earthquake and large loss of coolant accident (LOCA) initiating events provide a small contribution and can be represented'as I LL*0G*(DT + DG + SSPS) (2) where l 1 LL = Large LOCA Fragility l

i l Equation (1) was quantified with the SEIS4 computer code and resulted in the following annual core melt frequency: Mean = 2.84 x 10-5 Variance = 2.24 x 10-9 Based on the f ragility update, SSPS and DT can be dropped f rom the model, based on significantly higher capacities. However, a relay chatter fragility at a relatively lower capacity has been identified in the 4,160V switchgear. This chatter could have a negative effect; e.g., trip out the diesels. Until the consequences of this chatter are evalusted, it is assumed that the chatter fails both diesels. Therefore, Equation (1) can be changed as follows: OG*(chatter + DG) (3) , where Chatter = Relay Chatter Fragility (4,160V switchgear) Quantifying equations (3) for annual core melt frequency with SEIS4 results in Mean = 1.8 x 10-5 Variance = 9.58 x 10-10 Comparing the quantification of Equations (1) and (3) shows a slight reduction (less than a factor of 2) in f requency. However, this assumes the chatter fails the diesels without recovery. An ongoing relay chatter review will determine whether this particular chatter is a real concern. In addition, this review will determine whether there are any other relay chatters that should be considered in the model. 1.2 RELEASE CATEGORY S6 This release category is dominated by earthquake and transient initiating events. These sequences can be simply represented as NOG *SSPS (4) where NOG = Of f site Power Available (negation of OG - f ragility) As described above under release category S2, the solid state protection system can be dropped from the model. Therefore, the simple model in Equation (4) would go to zero. To actually determine the new S6 - frequency, the whole plant model needs to be requantified and unraveled to obtain new dominant sequences and frequencies. However, the trend is a reduced frequency unless the ongoing relay chatter review identifies new sequences.

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RAI 20

Assess the impact on risk of assuming that the containment capability corresponds to the pressure which produces 1% strain in the containment wall RESPONSE 20 A sensitivity analysis on the risk results in the SSPSA, the RMEPS and the WASH-1400 Sensitivity Study was performed to determine whether any of the basic conclusions with respect to risk or emergency planning requirements would change substantially if the containment was postulated to fail at a deformation strain of 1 percent. It is emphasized that there is no basis in experimental data or analysis to support a 1 percent strain failure con-dition as being a reasonable definition of failure or that containment failure at this condition is any more likely than what is implied by Figure 11.3-14 in the SSPSA or Figure 4-7 in the RNEPS.

                                                 ~

Since the failure definition is arbitrary, there is no meaning in assessing uncertainty in this failure condition and it has been postulated to occur with certainty when the containment pressure reaches that pressure value where the calculated deformation strain reaches 1 percent according to l Figure 4-2 in Appendix H.1 of the SSPSA. This corresponds to a maximum radial displacement of the containment wall of 8.4 inches and it occurs at a pressure of 175 psig or 190 psia. This corresponds to the low temperature wet containment condition in Figure 11.3-14 of the SSPSA or Figure 4-7 in the RNEPS. Since in the analysis no distinction is made between type B and type C containment failures (see answer to question 17), the composite wet containment probability distribution (solid curve) is stepped from the calculated value of 0.2 at 190 psia to 1.0 (guaranteed failure). Note this j implies that the containment failure analysis in the SSPSA predicted a 20

!           percent probability for containment failure at the pressure corresponding to the 1 percent deformation strain. For dry containment conditions (dotted curve), the elevated temperature conditions reduce the strength of the innermost steel layers and the 1 percent deformation strains are reached at a pressure of 158 pois or 173 psia. An containment failures which in the original containment failure analysis were predicted at pressures below these 1 percent strain pressures were retained without modifications. The modified containment failure distributions of the 1 percent strain failure sensitivity analysis are shown in the attached Figure 20-1.

, Next, the impact of this change on the calculated risk and on the conclu-sions with respect to emergency planning were examined. An impact would result in one of two ways. Containment event tree split fractions would i change where these were determined from the containment failure pressure probability distribution. Secondly, the timing and release magnitude of , i late overpressure failure release categories (S2 and S3) would change due to l earlier containment failure times. However, changes would only result in l l cases cases. which depended on containment failure pressures above 173 psis for dry q ( I

The impact on release categories was examined first. New probability distributions for the time of containment failure were generated for dry containment sequences. These are shown in Figure 20-2. These curves do not exhibit a steep change because the uncertainty in the containment pressure versus time relationship calculated by the MARCH and COC0 CLASS 9 codes is still a valid consideration. The net effect of these changes is to reduce the time of release for release category S3B in the RMEPS from 89 hours to 70 hours and to reduce the release duration of release category $28-3 from 56 hours to 37 hours. No changes result in either the conservative source terms in RMEPS or in the source terms for the WASH-1400 Sensitivity Study, because for all these source terms, the release timing was assessed as occurring before the time when the pressure in the containment reaches the 1 percent strain level. The release fraction factor for particulate radionuclides is shown in. Figure 11.6-3 of the SSPSA. It is shown that no significant change occurs between 70 and 90 hours. Therefore, there would be no change in the release fractions as a result of the above noted change in release timing. Furthermore, in both cases (S3B and S25-3), the warning times are still auch longer than required for complete evacuation. It is

        '          thus concluded that no change whatsoever can be identified in any of the consequences esiculated either in the SSPSA or in RMEPS or in the WASH-1400 sensitivity study as a result of postulating containment failure at a 1 per-cent deformation strain.

i ' Lastly, the impact on containment split fractions was examined in Section 11.7 of the SSPSA. In no case is a split fraction dependent on a contain-ment failure pressure in excess of the 1 percent strain value for dry con-dition (173 psia). The split fractions for two top events in the containment event tree are affected. The split fraction for top event 10B on Table 11.7-8 in the SSPSA shif ts slightly to increase the probability of late overpressure failure and correspondingly decrease the probability of basemat melt through. However, as explained in the answer to question 17, all basemat melt through cases are conservatively treated as late overpressure failures and this change therefore has no impact on the results. Secondly, the split fractions for top event 12A in Table 11.7-8 in the SSPSA would shift significantly to increase the probability of type C 3 containmect failure and decrease the probability of type B containment

            ' failure. However, since in the analysis all late over pressure failures were treated as type C failures, this change also has no effect on any of the results.

Overall, it is concluded that the assumption of containment failure at a pressure corresponding to 1 percent deformation strain has no discernible ef fect on any of the results and conclusions documented in either the SSPSA or the RMEPS or in the WASH-1400 Sensitivity Study. This conclusion can be traced to three distinct reasons: (1) no early pressure transients reach a magnitude of 190 psia, (2) the reduction in release timing for late < overpressure is insignificant with respect to warning times and release fractions, and (3) conservative analysis assumptions in the containment event tree quantification absorb any effect which would otherwise be visible in the release category frequencies. l I

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RAI 21

What is the impact on risk f rom accidents during shutdown and ref ueling when the containment f unction may not be available? RESPONSE 21 The purpose of this technical note is to address the risk from accident seque that could potentially initiate during plant shutdown at Seabrook Station. Specifically, this note is intended to answer a question posed by the NRC sta during their review of the Risk Management and Emergency Planning Study (Reference 1), the companion sensitivity syudy (Reference 2), and Question Number 21 (Reference 3). All work performed to date to identify and to assess the risk of potential accidents at Seabrook station has concerned itself primarily with scenarios that could initiate at or near full power operation. In the original full scope PSA (Reference 4), the coverage of accident sequences in terms'of initiating events, the possibilities for system success and f ailure states, and the treatment of dependent events met or exceeded those of other publishe PS As . This coverage was certainly greater than was possible during the sever when the Reactor Safety Study was performed. A j udgar - normally made in a PSA, and made in the SSPSA, is that the level of risk associated with accider that could initiate during full power operation, however small,

l 15 substantially greater than that associated with accidents that occur during shutdown. There are many reasons to support this judgment including the fact that at full power there is a greater level of RCS stored energy, after-heat level and inventory of radionuclides than the case with plant shutdown. There is also generally more time available to recover from adverse situations during shutdown. Several years after t'he SSPSA was completed a research project was performed for the Electric Power Research Institute in which the ris,k of-

   . accidents at the Zion nuclear plant during plant shutdown and RHR system operation was assessed (Reference 5). The only risk parameter quantified in this study was core melt frequency. The results in comparison with the results of the Zion plant PSA (Reference 6) for power operation events that had been completed previously by the same PSA team are as follows.

Core Damage Frequency Description Mean Median Cold Shutdown (Reference 5) 1.8 x 10-5 2.6 x 10-6 Power Operations (Reference 6) 6.7 x 10-5 5.0 x 10-5 Hence, the core melt frequency from cold shutdown events at Zion is less likely but more uncertain than that from power operations. The Zion cold shutdown study did not address consequences of these events; it only l addressed the frequency of core damage events. 2 1447P102986

The Zion cold shutdown study examined plant shutdown and startup procedures in detail to identify a wide spectrum of potential accident sequences that could originate and develop during plant shutdown. It also made use of an in-depth review of in-plant records and informatior that covered 10 refueling outages, 24 maintenance outages, and some 27,888 hours of RHR system operations. Several person-years of effort went into the Zion investigation.

                                                                                                       )

It is of interest in this note to address the risk from plant shutdown events at Seabrook Station, which like the Zion plant, is a four-loop Westinghouse PWR with a large dry containment. In the brief time available, it is not possible to complete the kind of in-depth ( examination that was described in Reference 5. On the other hand, for the purpose of addressing the implications on emergency planning, it is not sufficient to measure risk simply in terms of core damage frequency With this perspective in mind, the objectives of this response are to: o Provide an order of magnitude estimate of the frequency of core damage events that could initiate at Seabrook Station during plant shutdown. e Estimate the frequency of the above events that result in containmen bypass, containment high leakage, or containment intact end states. e Account for important specific and unique features of the Seabrook plant hardware and procedures. 3 1447P102986

e Provide a suitable allowance for uncertainties associated with a preliminary level of analysis through the appropriate use of conservative assumptions. e Provide for a reasonable level of accountability of operating 4 experience with events that have occurred in similar plants during plant shutdown. APPROACH The approach taken to address ~ shutdown loss of cooling events at Seabrook Station was first to review the Zion study (Reference 5), to compare the design and operational features of Zion and Seabrook, and to identify key differences important to the determination of shutdown cooling risk. Based on this review and the key differences that were identified, a determination was made of the extent to which all or part of the NSAC-84 results for Zion could be applied to Seabrook. In cases where Seabrook specific features indicate a reduced level of risk, appropriate corrections were made to the Zion results. Finally, a quantification was made of sequences that could occur at Seabrook Station at a higher frequency than that assessed for Zion. In summary, the risk of shutdown cooling events at Seabrook Station was evaluated as follows: i Seabrook Risk = Zion Risk per NSAC-84

                                                                                              - Portion of Zion Risk Not Applicable to Seabrook
                                                                                              + Portion of Seabrook Risk Not Applicable co Zion 4                                             .

1441P102986

i I r In other words, there are some design and operational features common to Zion and Seabrook and some unique to each plant. The enhanced features of Seabrook were accounted for by reducing the risk contribution of selected dominant sequences in the Zion results. This resulted in a reduction of the core damage frequency evaluated in NSAC-84. Then, the enhanced features of Zion were accounted for by adding to those results a separate Seabrook specific analysis of accident sequences that were not important in the Zion results because of its unique enhanced features. The above process resulted in a balance'd and unbiased albeit conservative assessment for Seabrook Station that was especially designed to make maximum and appropriate use of the Zion results for core damage frequency. Then, all the resultant core damage sequences were evaluated to determine the frequeny of three types of core damsge release states: core damage with intact containment, sma;; bypass, and large bypass. Finally bounding estimates were made of the contributions of shutdown loss of cooling events to the 200-rem dose versus distance curves in References 1 and 2. COMPARSION OF ZION AND SEABROOK DESIGN FEATURES The ability to respond ti this question quickly f acilitated by the fact that key plant and systems analysts of the Zion PSA team played major roles on the Seabrook PSA team. The design features of the respective plants were compared f rom two perspectives. First, the major differences between the two plants were noted based in our general understanding of the plants, systems, components, and PSA results. Second, the 34 dominant accident sequences for Zion shutdown cooling

t s

  }

I between these systems and the f rontline systems, such as the RHR system. For one thing, there are ways to utilize equipment on Unit 2 for Unit I and vice versa at Zion that are not possible at Seabrook. These differences stem from the f act that modern design criteria, to which Seabrook was designed

<   .        and Zion was not, call for a strict physical separation between redundant trains of safety-related systems and greatly reduce the opportunities for lining up cross-train pump and heat exchanger combination. In othere words, there are more success paths in the older plants such as Zion. Ironically,
    -        the introduction of these more restrictive design criteria in Seabrook produces a relative advantage for Zion in this regard. Therefore, we would i        expect to see a higher contribution from sequences involvong cross-train combinations of electric power, service water, component cooling water, and
  )l         RHR systems at Seabrook, relative to Zion.

s 4 l ~l i l l

3. Other Plant Differences.

The remaining plant differences that were identified could be significant in the determination of the risk of power operation events, but are not found

    ]        to be significant with respect to shutdown cooling risk. Thase differences
     !       include those in the containment heat removal systems (different configurations i,            of containment spray and f an cooler systems), use of solid state versus relay technology in the safeguards actuation system at Seabrook and the ability to
    ,        utilize Unit 2 equipment for Unit 1 and vice versa at Zion. There is a high I       degree of similarity between Zion and Seabrook in the procedures that govern I      shutdown operation. Of the dif ferences in this area, there are distinct          -

I advantages to Seabrook (e.g. some of the local manual valve operations at Zion are perf ormed remote manually from the control room at Seabrook. - 1 l l l i i

UTILIZATION OF NSAC-84 RESULTS FOR SEABROOK l Following the design and procedures review and comparison, the dominan sequences from Table 6-1 in NSAC-84 were reviewed for applicability tc Seabrook Station. The following conclusions were reached, e Because of the similarity between the plants and the procedures, t dominant sequences from Table 6-1 are generally applicable to Seabrook. e The NSAC-84 sequences would be expected to occur at the same frequency at Seabrook, except for those sequences involving inadvertent closure of RHR suction path MUVs and those involving combinations of support system faults and RHR train failures. The sequences involving suction path.MOV closures would occur at a lower frequency at Seabrook because Seabrook has a separate suction path for each pump. The frequency of valve closures was calculated as part of Tc Event RM in NSAC-84. This top event asks whether RHR cooling is maintained during maintenance and refueling outages. The cause table fo this event is shown in Table 1 (adapted from Table 5-5 of NSAC-84). Al s. 8 1447P102986 h-si. . . . .

shown in the table is a correction factor that shows the effect of two drop lines at Seabrook in lowering the frequency of " hardware failures" i and " human errors." Thedert'vationofthecorregtionfactorsis explained below. For spurious valve closure to cause a loss of RHR cooling at Seabrook Station, it is necessary to postulate either a common cause event involving one valve ih each suction path, or a coincidence of a single valve closure and maintenance being performed on the other RHR train (these could also be maintenance in a support system of the other RHR train. 'but these sequences are separately accounted for below). The correction factor for this cause of RM is given by SMOV + (1-SMOV) (.5) QRHRH = .072 where Sgoy = M0V Common Cause Parameter = .043 from SSPSA Section S . QkHRM = Maintenance Unavailability of a Single RHR Pump Train During Shutdown

                                                                        = 6.1 x 10-2 Based on Zion Data in NSAC-84 i

s The factor of .5 is the chance that the maintenance is being done in one - of two specific trains. l l i, The correction factor for errors in inverter switching is given by . i i

                                                          .5 QRHRM = .031 9

1447P102986

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The result of the abcve corrections to this top event is a reduction in the failure frequency to a factor of .145. This factor was applied to i applicable sequences in Table 6-1 and the following results were obtaine I i Core Melt Frequency Mean Median i ' NSAC-84 Results for Cold 1.8 x 10-5 2.6 x 10-6 Shutdown Resul'ts Corrected for two RHR 7.6 x 10-6 1.1 x 10-6. Suction Paths

• Estimated as source factor reduction as calculated for mean results.

,                                      Hence, because of the dominance of the valve closure events in the NSAC-84 results, the effect of having two suction paths is a reduction of g

core damage frequency of the NSAC-84 sequences at Seabrook by a factor of about 2. ANALYSIS OF SEABROOK SUPPORT STATE SE0VENCES i ! 8ecause of differences in the support system interfaces with the RHR i system and because these particular differences are unfavorable for Seabrook, separate event tree analyses were performed to cover these events for Seabrook. The following initially events were selected for this analysis. 10 1447P102986 i t- -- _ _ _ . - - , _ - _ . , _ . _..____m

Designator Initiating Event LOSP Loss of Offsite Power LIRH Loss of One RHR Train LIPC Loss of One PCC Train L2PC Loss of Both PCC Trains LISW Loss of One Service Water Train L25W Loss of Both Service Water Trains

• As shown in Figure 1, these initiating events were first analyzed in support system event trees whose sequences result in one of five different plant states. The plant states denote the number of RHR trains and safety grade AC power trains rendered unavailable by the combination of the initiating event and support system failures. These states together with the sequences borrowed from NSAC-84 as corrected for Seabrook were then fed into a frontline system event tree, which considers additional events needed to resolve the end states of the event sequences in terms of release categories. This main line event tree is 4

based in the event sequence diagram in Figure 2. In this analysis, the NSAC-84 sequences were assigned to support state R2E0 (loss of both trains of RHR with both trains of AC power available). 11 1447P102986

i The event tree quantifications for LIRH, L1PC, LISW, and LOSP are shown , 1 1 in Figures 3, 4, 5, and 6 respectively. The quantifications were based on the SSPSA and RMEPS results for the support systems and initiators, 4 except for maintenance unavailability. Train 8 of all systems was assumed to be unavailable for maintenance with a conservative value of j unavailability of 0.1. This more than accounts for the higher chance of maintenance during plant shutdown. The initiating events L2PC and L25W are assigned directl'y to support state R2E0 because of a very small chance of electric power failure with no loss of offsite power. T,he results of the analysis up to the point of support state are presented in Table'2, which is organized into three types of events: Type 1 is events with one RHR train unavailable (RIEO, R1E1); Type 2, with two RHR trains unavailable (R2EO, R2El, and R2E2), and Type 3 is the set of

NSAC-84 sequences.

i When the sequences are combined according to support state, the following results are obtained. 1 Support State Mean Frequency (events per reactor-year) R1EO 1.7 x 10-1 R1El 3.3 x 10-5 R2E0 2.0 x 10-3 i 1 R2E1 2.0 x 10-7 l R2E2 3.8 x 10-7 RXEY = Sequence with X RHR trains and Y electric power train unavailable. The event sequence diagram in Figure 2 defines the possible progression for Type 1, 2, and 3 support state sequences. For Type 1 and l 12 l 1447P102986 l l

 ,.-,,n. ,,- -n -,-...,_.-- -,,~    ~,---.,_-.-_.n      ---_n-n--,-         - - - - - - -. ,- ---- ,- ,-...-_ ,                                       .n

1 2 sequences, consideration is given to operator recovery to prevent core melt., On the other hand, such consideration is not made for Type 3 because such

 -   actions are already considered in NSAC-84. Next, the RO event question whether the RCS pressu're boundary is open initially; i.e., vessel head or steam generator nanway cover is removed. For RO closed sequences, the ESD tracks the possible developement of interf acing LOCA conditions either through check valve closures or RHR system repressurization via Top Events CV, RV, and MC. For pressure boundary open sequences or all other nonbypass sequences, consideration is given to whether large and small penetrations are initially open and, when open, whether or not operator actions to secure these penetrations are successf ul.

Since the containment sprays are not tracked in the ESD for simplicity, successfully isolated sequences could result in either a containment intact (SS) or delayed overpressurization (S3). Those with large or small bypass sequences are assigned to S6 and S2, respectively.

EVENT TREE QUANTIFICATION With reference to Figure 1, the event trees were quantified in two stages. Fi rst the support system event trees were quantified for each initiating event resulting in- the quantification of the uncon-dicional frequencies of 5 dif ferent RHR support states. RIEO, R1E1, R2EO, RZEI and R2EZ (where RXEY is the state in whcih X RHR trains and and Y trains of safety grade AC electric power are rendered unavailable. Then, the main line event tree was quantified 6 times, one for each RHR state and a separate quantification for the sequences borrowed from the NSAC-84 results. The derivation of the event tree split fractigns for each event tree quantification. Support System Event Trees (Figure 3, 4, 5 and 6) The support system event trees were quantified for the following initiating events. L1RH - loss of 1 RHR train LIPC - loss of 1 PCC train L2PC - loss of both PCC trains LISW - loss of I service water train L2WS - loss of both service water trains LOSP - loss of off site power LIRH The frequency of loss of 1 RHR train during shutdown was estimated using the following model. pt,1RH = Y RHR RHR where FL1RH = frequency of the initiating event (events per reactor year) 'r skRHR

                           =

failure rate of 1 RHR train (dominated by the RHR pump) tRHR

                        =

the number of hours per year in shutdown I Note the omission of MOV closure events in the above is by design, these events are included in the " type 3 events" borrowed from NSAC-84 and

corrected for Seabrook having 2 RHR suction paths from the RCS.

4 The time on RHR, t RHR, is estimated using zion experience, whcih is viewed as a conservative assumption for Seabrook. The reason for this view is that the zion experience is worse than average for PWRs and does reflect i l

the generally higher availability factors of the Yankee system of plants (Maine Yankee, Vermont Yankee, Yankee Nuclear Power Station). In the first 16 reactor-years of experience of zion 1 and 2 there were 12 j refueling outages of average duration 1,992 hrs and 3.05 maintenance ' outages per unit-year of average duration 488 hrs. t = I **f 'I 8' RHR + 3 05 maintence events / reactor-year 16 reactor-years x 488 hrs / outage = 2982 hrs / year Hence, in the first 16 reactor-years at zion the plant was shut down about 34% of the time. . We conservatively assume the same value for the plant lifetime at Seabrook. The support system event tree quantification for L1RH is shown in Figure 3. In this event tree, it is assumed that the plant is initially being cooled

• with RHR train A and train B is in the standby. Becuase of the strict train-wise dependencies at Seabrrok Station, critical operation of RRR train a precludes unavailability of service water and PCC trains A, since both are needed to operate RHR train A. In normal power operation, the unavailability greater than 10-3of single service water and PCC trains is very low- no to 10 -2 per train. However, during plant shutdown, the unavailability due to maintenance is generally higher. For example, at sion the plant specific data shows single train maintenance unavailabilities of RHR, SW and PCC the range of .03 to .06. It is conservatively assumed in the analysis that all safety grade non-operating subsystem have shutdown maintenance unavailabilities of 10-1 This is greater than any shutdown

, maintance unavailabilities observed in the data. I LIPC The loss of one PCC train initiating event is analyzed in Figure 4. The initiating event frequency is estimated with the following model. IPC 2ke RHR ks +)pe 3

where '

l pe = failure rate of 1 PCC pump = 3.4(.5) per hour from

• PLC-0300 Section 6.

tRHR = (.34) (8760) based on L1RH analysis

                                   )s         = failure rate to start of standby PCC pump
                                             = 2.4 x 10-3 from PLG-0300 Section 6.

! [=neantimetorepairtheinitcallyrunningpump

= 24 hours per PLG-0300 Section 6.

Hence = 6.3 x 10-4/ reactor year fL1PC

The LIPC event tree is quantified using information f rom the RMEPS and SSPSA for service assumptions as withwater train event. the LIRH A and the same train B maintenance LISW The configuration and failure rate of the SW system are comparable to the SW system, i.e., each train has an operating and a standby pump. The event tree for LISW is quantified in Figure 5. frequency is the same as that for LIPC. Note that The initiating event this analysis includes the tunnel SW system only. The SW cooling tower system is considered in the subsequent recovery analysis. LOSP The loss of of f site power event tree is quantified in Figure 6. The initiatin,g event f requency is estimated using the following model: [LOSP= LOSP t RHR EPR where LOSP = frequency of LOSP = .135 events / year (SSPSA Section 6) t RHR = time on RHR = .35 (see LIRH above) EPR = frequency of non-recovery of LOSP before core damage = .01 (assumed The above assessment for EPR can be compared with EPR-1 in the SSPSA, a value of .03 for full power operation. ' The.01 value is viewed as conservative in comparison with EPR-1 since the time constants for core recovery are much longer during plant shutdown. The event tree split f raction for LOSP in Figure 6 are based on the result s of the SSPSA and RMEPS for train A and 10% maintenance unavailability for train B used for all shutdown loss of cooling events. L2PC and L2SW The initiating event f requency for loss of both trains of PCC and loss of both trains of service water are estimated with the following model. 2PC = fL2PC CRHR

/L25W =/ L2sW E po E po I

where }t t AI = frequency of same event during power operation from SSPSA po = hours of power operation assumed in SSPSA (8760) i l l

                                                                                                                                                                                  ~

l i

  , - , - , - , - -,-,m,,..,.. -, ., , , _ _ _        _,_+----,_s,.,_,,,.__,._                                .n-.n                     ____,,_,,.n.                ,,-,_ , - -

_ Top Event RH For the LIRH and other RI A0 support states sequences, it is assumed that train A is initially used to provide RHR and train B is in standby. Hence, it is the train A subsystem that is involved in the initiating event. For these conditions, the f ailure to provide continued RHR cooling is estimated from the following model. where RH = m+hs+A PR p [m=RHRpumgstrainmaintenanceunavailabilityduringshutdown

                                                 = 6 x 10-        from Zion data in NSAC-84
                                                                                                                                                                )

h ps = standby pump failure to start vote

                                                 = 3.3(-3) f rom SSPSA Section 6                                                                      -

• h pr = running pump f ailure to run vote

                                                - 3.4 x 10 5/ hour from 3SPSA Section 6
                                           -[Op = meantime to repair the initially failed pump = 21 hours from SSPSA Section 6                                                                                             _

[RH = 6(-2) + 3.3(-3) + 3.4(-5)(21) = 6.4(-2) For RIAL, the same model is used except the standby pump must run longer to cover the repair time of a diesel generator - assumed to be one week. Hence, for R2EO, R2El, and R2E2, /IRH = 1 ORH = 6(-2) + 3.3(-3) + 3.4(-5)(168) = 6.9(-2) Top Event OM RlE0 represents the most ideal conditions and minimum stress levels of these considered for OM for these conditions, OM is estimated from: 0:1 = DEI = SC = BF where del = operators f all to recognize that RCS heat removal should be restored after running RCS pump stops. SC = operators fail to align and restart a core cooling system BF = system operators fail to provide long term makeup to chargining Using appropriate values from Table 5-6 of NSAC-80, the following quantification is made: oM = 1.0(-5) + 5.0(-4) + 1.0(-5) = 5.2(-4) As shown in Table 3, higher values are used for the remaining support states to reflect different and progressively greater stress and comparison levels type 3going through events, OH =the 1 sequence RIE0, R2EO, RIEL, R2El, R2E2. For considered in NSAC-84. to avoid double counting recovery already _ _ _ _ _ _ - - - . - _ - - - - - - - ' - - - - - - - - - - ~ - - ~ ' - - - '--'-

Top Event RO The fraction of time the pressure boundary is open is keyed to the time assumed for RHR shutdown cooling. For consistency, since Zion data was used to quantify the latter, it must be used to quantify the former. From Table 3-4 of NSAC-84, the total time the RCS is opened during maintenance outages is 5,014 hours. From table 3-1, the RCS open time during ref ueling outages is above 3,000 hours. Hence, overall the 31,687 hours of the Zion outage experience, the RCS was opened (8014/31,687) = .25 of the time. Top Event CV The frequency of CV is quantified per our response to question 47. kV=5.5x104 Top Event RV This event is quantified in RMEP'S; it is estimated using:

                                                         /RV = 21 RV = 4.8(-5) wherelRV is the failure rate of each RHR relief value = 2.4(-5) from SSPSA Section 6 Top Event MC For both trains of AC power available, the top event is estimated using the following model.

MC = 2(KMOV2+B MOV 1HOV)+4()CV2 + BhCV)

                        ) MOV = f ailure rate (f ail to close on demand)
                                         = for MOVs - 4.3(-3) per SSPSA 1CV=failurerate(failtocloseondemand) for check valves = 5.5(-4) per question 75 response.

HOV; bCV = Beta factors for each type of valve = .1 hence l /MC = 2[(4.3 x 10 3)2 + (4.3 x 10 3)(.1)s + 4[(5.5 x 10 4)2

                                           + (5.5 x 10 4)(.1)s = 1.1(-3)

For one AC power bus available (RIEL and R2EI) there is only one MOV on each suction path potentially available. For these states

                          /IMC = 21MOV + 4(                                    lCVBCV)
                                         = 8.8(-3)

For two AC power basas unavailable fMC=1 f

 -,-m -

n -n--- , , , , , , - - --,,e--n , , - - - - , - , , - - - . - - . - . - - - .,>.w---._,,- - - . - -----,w- - - ,,-~

Tap Evnnt LI This event question whether any large penetrations are open initially. Large is defined as a total equivalent single opening of greater than 3" in diamter. Examples of purge containment such penetrations penetrations.are equipment hatch, personnel hatch, and unless fuel is being moved. These penetrations maybe opened during shutdown l The for the shutdown. chance that large penetrations are opened is highly dependent on ht reason If the reason is refueling, steam generator manintenance or othere maintenance on reactor coolant system somponents, it is likely that large penetrations such as the equipment hatch will open. the outage occurs due to need to repair or maintain equipment outside theIf on the other hand, containment, (e.g. turbine generator related maintenance) there would not be a compelling reason to open up large penetrations in the containment. To reflect a status of event the above RD. consider' tions, LI in assessed as a f unction of the If R0 is true (reactor coolant system is opened), it is assumed that LI is true (large penetrations are open) 90% of the time. If RO is not true (reactor coolant system is closed), it is assumed that LI . is true on 10% of the time. Note that at Zion, of the 8,014 hours during shutdown that (roughly 160 hrs per refueling outage).the RCS was opened, the fuel was being shuffled Hence 80 percent of the time that the RCS was hatch open.opened, it would have been permitted by tech specs to have the equipment Top Event OL Given a large penetration is opened initially, the event questions whether the operators situation couldsuccessfully develop. The close the penetrations before a potential release probability of successful recovery is assessed as dependent on the RHR support state, i.e. the combination of the initiating event and the response of the plant support systems. At different support states then operator wouldactions. recovery be different levels of stress and confusion to inhibit To provide an indication of the amount of time available to close the equipment hatch or other large penetrations, the time to core damage, taken as the time to uncover the core was estimated for the following cases: Cases Time to core uncovery (hr)

1. Reactor vessel head open with water level at hot leg nozzle midplane A. Loss of cooling at 2 days after shutdown B. Loss of cooling at 30 0.8 days af ter shutdown 2.6

2. RCS filled at pressure f,425 psig with A. Loss of cooling at I day af ter shutdown B. Loss of cooling at 10 days after shutdown 5.4 C. Loss of cooling at 30 days af ter shutdown 14 22 -

3. Water at refueling level with A. Loss of cooling at 5 days af ter shutdown B. Loss of cooling at 72 30 days af ter shutdown 162

a 6 It is not known how quickly the equipment hatch can be secured. Our current information is it would take several hours to attach, and up to 8-12 hours to secure all the bolts and establish a tight seal. Just how quickly this process can be accelerated is uncertain. To address this uncertainty, a base case and a bounding case are performed.' Ins the base case, it is assumed that the mean time to close the hatch is 4 hours.' In the bounding analysis, it is assumed that the hatch remains open with a probability of 1. From the Zion data in NSAC-84, there were 10 reflueing outages and 24 forced maintenance outages resulting in an average, outage duration of about 39 days. Of the entire 31,687 hours of outages, roughly 1% of the time .the RCS was drained, 5% of the time refueling was taking.Flace, 5% of the time the plant was not on RHR, and in most of the remaining 89% of the time, the reactor system was filled on RHR. Based on the above recovery time, and assuming a 4-hour hatch recovery time, it is seen that with the RCS drained to the hot leg nozzle midplanes, the chances of successful hatch recovery are not very' high. While under all other conditions, the chances are high. Therefore, the base case and low stress levels, a yalue of .01 is used for failure to isolate large penetrations. For degraded RHR states, this valve is increased to correspond with higher stress levels, as indicated in Table 3. In the bounding case, a failure frequency of 1 is assured for all states. Top Events SI and OS It is conservatively assumed that small penetrations are open 90% of the time and the chances of recovery are assessed at levels compoarable to those for OL, even though all small penetrations can be isolated quickly. Results The results of this preliminary analysis of shutdown loss of cooling events are shown in Table 4 for the base case assumptions on event OL. To bound.the consequences of these events, accident sequences were assigned to the existing PSA release categories, even though the release fractions for shutdown events would be expected to be considerably lower than those calculated for power operation events. Based on what is believed to be a very conservative set of assumptions in this base case, the tapact of shutdown events is assessed to result in no greater than a 14% increase in core melt frequency, and an 18% increase in category S6 frequency. For the bounding case of no credit for l event OL, the frequency of S6 would increase to about 5 x 10-6/ year. l The impact of these bounding estimates of shutdown events en the dose vs distance l curves for 50 res and 200 rea whole body gamma doses are shown in Figures 13 and 14 for the base case OL and bounding case OL assumptions, respectively. Our best current statement of risk levels is represented by Figure 13. As seen f rom this figure, the addition of shutdown events tapacts the right tails of these curves, but the combined results at I mile are still less than the l NUREG-0396 valves at 10 alles. Even with no credit for equipment hatch recovery l l as assumed Figure 14, the combined shutdown and power operation results fall below the NUREG-0396 10 mile levels at less than 2 miles. Hence, even a very conservative analysis of these events does not impact the conclusions of ( the sensitivity study. It is expected that a more detailed investigation of l these events would result in auch lower levels of risk than either set of results presented here. i

REFERENCES E

1. Pickard, Lowe and Garrick, Inc., Westinghouse Electric Corporation, and Fauske and Associates, Inc., "Seabrook Station Risk Management and Emergency Planning Study," prepared for Public Service Com New Hampshire, New Hampshire Yankee Division, PLG-0432, December 198

2. Pickard, Lowe and Garrick, Inc., "Seabrook Station Emergency Plannin ,

Sensitivity Study " prepared for Public Service Company of New Hampshire, New Hampshire Yankee Division, PLG-0465, April 1986. 1

3. ,

Letter from S. M. Long, U.S. Nuclear Regulatory Com ssion Staff, to R. J. Harrison, Public Service Company of New Hampshire, Wquest for Additional Information...'," Docket Nos. 50-443 and 50-444. 4 Pickard, Lowe and Garrick, Inc. "Seabrook Station Probabilistic Safety Assessment," prepared for Public Service Company of New Hampshire and Yankee Atomic Electric Company, PLG-0300, December 1983.

5. Bley, D. C., and J. W. Stetkar, " Zion Nuclear Plant Residual Heat J Removal PRA," EPRI/NSAC Report NSAC-84, July 1985.

l

6. Pickard, Lowe and Garrick, Inc., Westinghouse Electric Corporation, and Fauske & Associates, Inc., " Zion Probabilistic Safety Study,"

prepared for Commonwealth Edison Company, September 1981.

i Table 1. Q)RRECTION OF NSAC-84 RESUI.TS FOR RHR LOSS TO ACCOUNT FOR 2 SUCTION PATHS Failure Cause Seabrook l Revised Mean Value Dominant Contributor Correction Failure Factor Frequency Hardware Failures 6.08-2 Spurious Closure of RH8701,or RH8702 + .072 4.38-3 Maintenance 7.37-3 j Running RHR Pump Falls with Standby Pump x1 7.37-3 , Out for Maintenance I Human Errors 6.00-2 Errors During TSS 15.6.36 or Inverter x .031 1.86-3 i Switching (RH8701 or RH8702 close) ! Support System 2.94-6 Component Cooling Water Heat Exchanger Failures x1 2.94-6 ) Failures j Dependent Component 6.03-3 ' RHR Pumps Fail During Operation xI 6.03-3 Fail ures ' ( l Total 1.34-1 I 1.96-2 I 9

• g , .. - - _ - -- ._ . _ ____ __ _ _ _ _ _

I ! TABLE 2. CLASSIFICATION MAIN LINE MODEL QUANTIFICATION OF SUPPORT MODE CASES Impact Vector Power Class f cation A B A B ' Type 1 Events - One RMR Train Nede Unavailable _ LIRH ' 1.7-1 LIPC9A X RIE0 5.0-5 X L0$P43r01R 2.0-5 Rito X X , LOSP4A191R Alt 1 1.3 5 X L15r$R X R1t1

• 5.1-4 X LOSP4BM9R RIE0 3.4-6 LOSP9Br$4 X 3.5-7 Rit0 LO$P9A24R X X 8.9-8 R1(1 LOSPNA3*$4 X 5.5-4 RIE0 X X LIPCNA3*$R RIE1 1.0-9 ,

                                                                                                                  ,   X RTE 0 Type 2 twents - Two RHR Trains Mode Unavaflable L1RN*PSM9R        '

1.8 3 \ I X LIRPW8mSR 2.0-4 X R2C0 LIPCVSM9R X X 5.5-6 R2to Li$rWSM*$R X X 4.3-7 R2E0 L1$rP8r$R X X 5.7-7 R2t0 X X LOSP*0Al*G8r02R 3.5-7 R2E0 X X X X L0$P4A199r01R9R 1.4-7 X R2E2 X X L 0$P9A2*G8M*01R 6.0-8 R2E1 X X X LOSP4A19BF01R*$R 1.5-8 X R2t1 X X X L0$P9A293rPR 9.4-9 R2E2 X X L0$P9A39Br$R 6.9-9 X R2E0 LOSPNA398M*$R X X X R2t2 6.2-9 X X X L0$PNA343r01R*$P 3.9-9 X R2E1 X X X L0$P9A243r$R 1.1-9 R2E2 LIPCNA398r$R X X X R2t1 2.210 - X X LIPCNA3*P9M*$R R2E0 2.0-10 X X R2fG Type 3 Events - From NSAC-84 7.6-6 X X R2E0 NOTE: Exponentfal notation is 1.e. 1.7-1 = 1.7 x 10-1. indicated f a abbreviated form; 1448P102886

TABLE 3.

SUMMARY

OF SPLIT FRACTIONS FOR MAIN LINE EVENT TREE Main Line Event Tree Quantification Cases Event Tree Type 1 Type 2 Type 3 Top Event ' RIE0 (NSAC-84) R1El R2E0 R2E1 R2E2 Rif 6.4-2 6.9-2 1 1 1 1 W 5.2-4 1.0-2 2.0-3 1.0-1 1 1 RU .75 .75 .75 .75 .75 .75 IV 5.5-4 5.5-4 5.5-4 5.5-4 5.5-4 5.5-4 RV 4.8-5 4.8-5 4.8-5 4.8-5 4.8-5 4.8-5 hT 1.1-3 8.8-3 1.1-3 8.8-3 1 1.1-3 ITlR0 .90 .90 .90 .90 .90 .90 ITlE .10 .10 .10 .10 .10 .10 IR- 1.0-2 3.0-2 1.0-2 3.0-2 .10 1.0-2 3T .90 .90 .90 .90 .90 .90 E 1.0-2 .10 1.0-2 .10 1 1.0-2 X = Event Success T = Event Failure NOTE: Exponential notation is in i.e., 6.4-2 = 6.4 x 10-2. dicated in abbreviated form; 1448P102886 I w_ - - - - - - - - - - -

TABLE 4 KEY RESULTS OF SHUTDOWN SEQUENCES FOR SEABROOK STATION Release Category Event Tree

         ;                         55 or S3     S2       SS RIED              5.6-6      5.0-8   1.9-8 RIE1           . 1.9-8      1.7-9   2.0-10 R2E0              4.0-6      3.5-8   1.4-8 R2E1              1.8-8      1.8-9   1.9-10 R2E2              3.2-8     3.3-7    1.2-8 Type 3            7.5-6     6.8-8    2.6-8 I

Total for Shutdown 1.7-5 4.9-7 7.1-8 Events i Total for Power 1.1-4 2.0-5 3.2-7 Operation Events

l Percent Increase 13.3 2.4 18.2 with Shutdown t-Events i

NOTE: Exponential notation is indicated in i abbreviated form; i .e., 5.6-6 = 5.6 x 10-6, I ll 1448P102886 i,

l !l 1l S E ET CA H HS HS NT T S I T T A S ES U E N"" O' EW EWP I A I EWP QS _ O" _ RE OG S RE Y RE Y E A S E O CA S OG CA L S B2 OG E B6 CA S T L NE EM NT* M A MA A L MG AR V ED y D D M S DA L N E m" m"

                                                                                     )

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i. INITIATING SERVICE SERVICE EVENT PRIMARY PRIMARY WATER WATER pygiCLENCY L1RH TRAIN A CowoENT COWOENT NUFSERO TRAIN B MNTS PER WA COOUNGA COOLNGB RHR TRAIN WB PA EACTOR PB MADE YEAR) UNAVAILAB.'

         .20/ YEAR a                     1 m      .90     m1 me                        me                                     m      .90
                                                              \1                      r,                1       .165             1
                                                                                              .10       2    1.8 (-2)           2 0                    m 3        0
                         ,                                             .              v                                        1 4         0             2
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_, , y, 5 2.0 (.2) 2 0 _ ,,, s 0 2 o _ y_ _o 7 o 1 8 0 2 _ g_ _ g, _ 9 0 2 GF= GUARANTEED FAILURES FIGURE 3. SUPPORT SYSTEM EVENT TREE QUANTIFICATION LOSS OF ONE RHR TRAIN (L1RH)

• - p --

                                                                       ---,-.-w.,,.-m

INITIATING SERVCE SERVCE EVENT WATER PRIMARY PRIMARY C SEORJCE WATER L1PC TRAIN A CCM2ONENT CCM20NENT IHEQUEPCY TRAIN B NUMBER WA COOLINGA (EVENTSPER WB PA COOLNG B RHR TRA PB REACTOR MADE YEAR) UNAVAILA 6.3 (4) g - 1 Q.90 0 90 1_ 0 0

                                                                                         .10             2        0             1 1

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                                                     .10

_ , gp ,, _ _ gp_ - 9 2.2 ( 8) 2 GF= GUARANTEED FAtLURES FIGURE 4. SUPPORT SYSTEM EVENT TREE QUAnTIFICATION FOR LOSS OF ONE PCC TRAIN (LIPC)

                                - - - - - - - ~ - ' ' ' '       '

INITIATING SERVICE SU MCE SECLENCE EVDg PRIMARY PRIMARY WATER WATER FRiOLENCY L1SW TRAIN A Cowote(T CO W OtENT t NUf4EROF TRAIN B COOUNGA (EVENTS PER RHR TRAINS WA WB CDOLNGB PA NEACTOR MADE PB YEAR) UNAVAILABLI 6.3 ( 4)'#n0

                          ^           ^            ^
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                                              ., g        g.,     9     6.3 ( 5)        2 GF = GUARANTEED FALURES FIGURE 5.

SUPPORT SYSTEM EVENT TREE QUANTIFICATION F LOSS OF ONE SERVICE WATER TRAIN (LISW) t l l I

l l l l l INITIATING PRMARY PRMARY I ' g PONER N SEGANCE TRAIN A PONER WATER WATER N N WER W TRAIN B TRAIN A COQUNG CKCUNG GA TRAIN B (EVENT 5 PER RHR TRAINS 08 WA WB TRAIN A TRAIN B PA FEACTOR MADE PS YEAR) UNAVAILABLE 46 (4) m .93 m 90 m .98 v v 6 90 m a 1 we v we m .90 1 3, 3.1 f 4) 0

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                                                         .10

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_ , y _ _ ,, , , _ 15 30f6) 2 10 _ ,, ,, p , ,p, _ , p,,,,,,,,,,g,, 18 3 4 f 61 2 LOSP . LOSS OF OFFSITE POWER WITH NO RECOERY SEFORE C GF. GUARANTEED FAILURE FIGURE 6. SUPPORT SYSTEM EVENT TREE QUANTIFICATION F LOSS OF 0FFSITE POWER (LOSP)

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EVENT TREE QUANTIFICATION FOR R1[1. EVENTS TIMT FAIL ONE TRAIN OF $AFETY GRADE AC POWER l i

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• 1J 8101 FIGURE 10. EVENT TREE OUANTIFICATION FOR R2E1 - EVENTS THAT FAIL TWO RNR TRAINS AND ONE TRAIN OF SAFETY GRADE AC POWER

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1

                                  .                ....is              .    .    .
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                 >                                                                            FOR EVENTS ATPOWER
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RAI 22

It is the staf f's understanding that preexisting violations of containment integrity were " included" in the PSA by assuming the average effect was to raise the containment leak rate to the design basis value of 0.1%/ day. a. Compare this assumption with the containment integrity violation data presented in NUREG/CR 4220. b. What contributions would these containment integrity violation dat a make,to the probabilities for each of the release categories (assume the S5W category is redistributed over all the appropriate categories by the conditional probabilities of preexisting leakage paths of the size appropriate for each category). RESPONSE 22 As discusse'd in Section 11,6.3 of 'the SSPSA (PLG-0300), release category SS represents accident sequences where the containment remains intact (does not fail). All containment failures and bypasses, including failure to isolate S2, S3, the S4, containment and S6). are included in the other release categories (S1, Preexisting containment leakage is quantified in S5 by assuming an effective average leak rate equivalent to the containment design leak rate of 0.1% per day. As shown in Section 11.6.4.1, this is a reasonably good approximation when compared to Weinstein's Contain=ent Integrity availability work (reference 11 6-18 of PLG-0300 - also referenced in NUREG/CR-4220). This leakage is .also imposed on category S3 but is increased to reflect the predicted higher containment pressures. Weinstein's work found PWR containment integrity to be available 97.3% of the time and an average leak rate of 31 times design basis the rest of the time. a. NUREGICR-4220 suggests that PWR containment integrity availability could be as low as 71% with leakage in the range of 1 to 10 times allowable 29% of the time. The probability of larger leaks (28 square inches) is estimated to be in the range of 0.001 to 0.01 with a point estimate of .005. Section 9 of NUREG/CR-4220 states that these are upper bound estimates of containment unavailability. Section 6 1 indicates that the 0.29 unavailability is based on 215 events in - 740 reactor years where type B and C leak testing results exceeded 60% of allowable leakage; a leakage duration of one year is simply assumed for each event. Section 6.2 indicates that the .01 to .001 chance of a 28 square inch leak is based on 4 events in 740 reactor years ranging from "small drilled holes" to an open six inch valve; basis. a There again one year duration is simply assumed with little, if any, is insufficient information presented in this document to assess the applicability to Seabrook of each event. We believe it is highly conservative to assume that 0.29 events per year will occur at Seabrook in which 60% of allowable leakage is exceeded to a level one to ten times allowable leakage and that such conditions will exist 29% of the time. It is also highly

conservative to assume that 0.005 events per year ranging "f rom small drilled holes to an open six inch valve" represents a 0.01 to 0.001 chance of a 28 square inch leak. The reasons why we believe the application of these assumptions to Seabrook to be conservative are as follows. Fi rs t , it seems extreme to assume i that evidence of exceeding Technical Specification limits with  ! LERS will mean that the extent of the leakage is ten times the limit. Also, events involving mispositioning of manual valves are subjected to monthly surveillance testing at Seabrook, therefore, one year is an inappropriate f ault duration for Seabrook. Many of the possibilities for mispositioned valves will be covered by automatic containment isolation actuation, whose failures are included in the PSA. With regard to the f.our large 1e'akage ev'ents noted in sections 4.1.7 and 6.2 of the NUREG, the information provided is insufficient to compare directly with Seabrook; however, it may be that none of these events are applicable to Seabrook (especially for durations of , one year) for the following reasons.

1. The containment purge valves at Seabrook are leak tested every six months or less; their position is checked monthly; they are actuated valves which receive containment isolation signals (failure of containment isolation signals / valves is included in the risk models).

2. Seabrook's containment is three to four feet thick; it is

, difficult to imagine that a hole could inadvertantly be drilled through it, never mind go undetected.

3. At Seabrook, valves, flanges, penetrations, airlocks, etc. , are leak tested and position checked after any maintenance activities l on them.

t

4. Manual isolation valves outside containment are position checked every month. All isolation valve positions are checked before

return to power operation and at least once per 18 months.

The upper bound frequency of pre-existing leakage derived in NUREG/CR-4220 is greater than that included in the Seabrook PSA; however, they are overly conservative estimates which are inappropriate for use in the Seabrook risk models.

b. Ragardless of the above, the ef fects of applying the NUREG/CR-4220 preexisting leakage estimates to Seabrook was evaluated; it was found that even these upper bound estimates would not result in any of the emergency planning risk criteria to be exceeded. ~

The effect of an assumed small preexisting leakeage, one to ten times allowable, was evaluated as follows. Assume a preexisting leak of ten times design leak rate 100% of the time. This would increase the source terms for SS and the first part of S3 by a factor of ten. Since the source terms for the first 24 hours of S3 are greater than S5, S3 will be evaluated. (For the first 24 hours, S3 and SS are the same leak size but the driving containment pressure is higher for S3.) ~ Using Tables 4-14 of PLG-0432 and 4-4 of PLG-0465, it can be seen

that the first 24 hours of the release for S2C clearly envelopes ten times the first 24 hours of S3W in terms of source terms; release timing is also conservatively enveloped. Table D-1 of PLG-0432 shows that S2C is an insignificant contributor to early f atality risk and, in f act, S2C has zero consequences (early fatalities). Table 1, which follows, is a partial reproduction of the no evacuation case risk summary table for the EPZ Sensitivity Study (PLG-0465) which was provided in the response to question 23; however, the S3 and SS releases have been I replaced by S2C. (The S2C source term represents 50 to 100 times the Seabrook maximum design leakage rate and assumes it exists 100% of the time. In other words, Table 1 represents a 100% chance of five to ten times the preexisting " Technical Specification Violations" leakage predicted to occur in the NUREG 30% of the time.) As Table 1 shows, this extremely conservative case still shows sero early fatalities and a small contribution to early injuries. NUREG/CR-4220 estimates an upper bound of SE-3 for "large leakage" which is conservatively assumed to be a six inch valve (or 28 square inch hole). We can conservatively bound the risk contribution of this by assigning this frequency to release category S6W. In Figure 1, we plot the upperbound effects of both small and large leaks in the 200- and 50-rem dose versus distance curves based on the assignment of small leaks to S2-CH and the large leaks to S6W accordir.g to the NUREG-4220 probabilities. While we don't agree that these results are reasonable for Seabrook, they show that the NUREG-0396 results at ten miles do not occur at Seabrook at one mile or less. Hence, a conservative interpretation of the NUREG/CR-4220 experience with pre-existing holes has no impact on the conclusions of the sensitivity study. 4

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RAI 23

i

a. Provide a narrative description- that quantitatively delineates the dominant contributors to the dose prob-ability vs distance curves and the early fatality prob-ability curves. The dominant release categories should be specified and the dominant accident sequences contrib- i uting to each of these release categories should be s pecified . The probability of occurence of each release category should be stated. These data should be provided for the current study and for the original PSA results.

Changes between the two studies should be attributed to . specific differences in the analysis.

b. Provide a set of early fatality conditional probability curves for each release category, assuming evacuation distances of 1 mile and 2 miles.

c. Provide the conditional mean risk of early f at'ality for each of the curves provided in b. ,

RESPONSE 23 1 The following describes the principal contributors to early health risk at Seabrook Station, as determined in the original probabilistic safety assessment (PSA) in 1983 (PLG-0300), in the PSA updates of 1985 (PLG-0432), and in the sensitivity study (PLG-0465). The risk measures of interest here are the early f atality risk curves and the frequency of exceedance of dose and distance curves for the whole body dose of 200 rem. The three (3) parts of this question are addressed collectively.

1. SSPSA RESULTS (1983) 1.1 EARLY FATALITY RISK CURVES The only results available for early health risk in the Seabrook Station Probabilistic Safety Assessment (SSPSA) (PLG-0300) assume a 10-mile evacuation zone. Significant and costly analysis would be required tu produce these results assuming evacuation distances of 1 and 2 alles. The contributors to risk can be expressed in a number of different ways. Alternative ways to group accident sequences are to group the individual sequences by initiating event, by plant damage state and by release category. A graphic display of how sequences grou, ped by release category contribute to the mean risk of early f atalities in the original PSA is shown in Figure 1, which is taken from Figure 13.2-la in PLG-0300. As seen

• in this figure, release category S6 (large isolation failure) makes a small contribution, and all other categories make negligible contributions that are at frequency levels below 10-9 per reactor year. Note that the mean risk curve, whose contributions are being i discussed here, is the mean of a family of curves that characterize uncertainty in the risk estimate. This f amily is shown in Figure 2, which is reproduced from Figure 13.1-Sa in PLG-0300. The fact that the meanlarge indicates curveuncertainties.

f alls well outside the median (.50) risk curve - These uncertainties are due to uncertainties in estimating the accident frequencies, source terms, and site model parameters. The conditional risk curves for each release category can be found in Figure 3.2-2A of the SSPSA. (See page 13.2-78 of PLG-0300.) A tabular representation of the information in Figure 1 is provided ' in Table 1, which is adopted from Table 13.2-7a in PLG-0300. This table .! shows that more than 99% of the mean risk curve comes from release category S6. Most of the remaining contribution comes from S2. Only l in the extreme right hand tail, at frequencies below 10-9 per reactor year, does another category appear, S1 (early containment rupture due to steam explosion, early overpressure, or external missile). The next step in breaking down the SSPSA risk contributors is to ' l examine the contribution of sequences grouped by initiating event. Because nearly all of the early health risk comes from S6 and a sna11 contribution from, S2, it is more efficient to confine our l search to these release categories. The initiating events that make significant contributions to S6 and S2 are provided in the table below, which was adapted from Table 13.2-Sa in PLG-0300. As can be seen, release category S6, which indicates early fatality risk, is, i n turn, dominated by the interfacing system loss of coolant accident

(LOCA) (V-sequence).

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Initiating Event Percent Contribution S6 S2 V, Interfacing LOCA 76. O ET, Seismic Transient 22. 95 EL, Seismic LOCA 2. 5 Others < 1. <1 i l Total l 100. I 100. l In a similar f ashion, accident sequences can be grouped with respect to plant damage states (sometimes referred to as a bin). The following is the plant damage state breakdown of the S6 and S2 release categories.

• Plant Percent Contribution l l Damage State
• I S6 S2 IF 78 0 IFF 0 5 3F 21 0 3FP O 35 7F 1 0 7FP O 60 Others

                                                                        <1                            <1 Total                       100                     1      100 In comparing the previous two tables, note that 100% of the interf acing                                  .

systems LOCAs were modeled as IF states. Hence, of the 24% of the seismic contribution to S6, 21% terminated in plant damage state 3F, 2% in IF, and 1% in 7F. Hence, most of the overall risk contribution comes f rom the interfacing system LOCA initiator and plant damage state 1F. ' Essentially, all the remainder are seismically initiated sequences ending in plant damage state 3F. Note that all the FP states, and have more thsn 99% of the release category S2 and are dominated by similar sequences - with the same release path and characteristics, namely Station Blackout ' with a RCP seal LOCA and failed open seal return line isolation valves. m The result is high early leakage and delayed ove pressurization of the containment.

• See Table 1-2 in the Seabrook Station Risk Management and Emergency ~

Planning Study (RMEPS) (PLG-0432) for definitions. Numbers denote containment and reactor coolant system conditions at time of reactor vessel melt-through; letters denote status of containment systems and leak paths. F states are isolation f ailures or bypasses more than 3 inches in diameter; FP states are isolation failures less than 3 inches in diameter. . t

The final step in breaking down the early health risk is to examine specific accident sequences. In the SSPSA, an accident sequence is a single path that can be traced through the plant event trees from the point of entry (the initiating event) to the point of termination (the plant damage state). As with other PRAs, the interfacing system LOCA was analyzed as a single sequence. That is, the event was analyzed as an initiating event and assigned directly to the most severe plant damage

 . state considered in the study and denoted as IF.       This reflects the conservative assumption that multiple failures of the interf acing valves automatically result in a core melt and early large containment bypass.

All other initiating events were modeled through the plant event trees, which include more than 4.5 billion sequence's counting all the initiating events, t he plant damage states, and paths connecting them through the plant event trees. TSerefore, the above contributions from the V-sequence are from a single sequence, whereas the seismic contributions , come from many sequences. i ( The nature of the specific sequences initiated by seismic events is next described. Of the seismically induced transients that make up 22% of the i f requency of release category S6, the single sequence having the greatest i f requency makes up only about one-fourth of this contribution and was analyzed as follows.

                                                            'l                      l Event                                     Frequenev Earthquake Occurs (.3g)                            1.1 x 10 4/ year       ~

Off site Power Does Not Fail .35 Solid State Protection Fails .041 Charging Pumps Fail .88 Containment Is Initially in Purge Mode .10 Emergency Core Cooling System, Containment Isolation, and Containment Sprays Fail [ dependent failures resulting i from loss of solid state protection system (SSPS)W 1.0 i Other Equipment Does Not Fail .86 Total I 1.2 x 10 7/ year l i l The remaining three-fourths of the seismic sequences in S6 are made up from a large number of sequences, some involving loss of of fsite power and others involving f ailures cf other equipment.

In a similar f ashion, the seismic contributors to release category S2 are also spread over many sequences. The single most frequent sequence in this category is a seismically induced loss of offsite power and f ailure of both diesels due to either seismic causes or independent causes. This sequence appears several times in the scenario identification tables (in Section 13.2 of PLG-0300) once for each discrete range ,of ground accelerstion. The total frequency of this sequence summed over all values of ground acceleration is 6.9 x 10-6 per reactor year, or about 40% of the total release category frequency.

     -    In summary, the early health risk curves in the SSPSA, which were only performed for a 10-mile evacuation zone, were dominated by the interf acing LOCA sequence (about 76% contribution to the mean exceedance frequency in the risk curves). Most of the remaining contributions come from seismically induced sequences with release paths either through the purge lines (the S6 sequences) or through the reactor coolant pump seal return line (the S2 sequences).

1.2 DOSE VERSUS DISTANCE CURVES The 200 rem and 50 rem dose versus distance curves that correspond with the SSPSA results are compared with the RMEPS and sensitivity study in Figure 3. The SSPSA curves are dominated by release categories S2 and S6.

2. PSA UPDATE RESULTS (RISK MANAGEMENT AND EMERGENCY PLANNING STUDY. 1985) 2.1 EARLY FATALITY RISK CURVES In the RMEPS update of the Seabrook PSA, the following changes were made that had an impact on the risk levels and the ordering of the risk contributors.

Plant Model Changes Item 1. The single sequence interfacing LOCA model was replaced by a two-event tree model, one for suction side and one for injection side residual heat removal / reactor coolant system (RHR/RCS) interf acing valve ruptures. This led to a reduction C in the frequency of plant damage state IF and the addition of three new plant damage states (1FV, IFPV, and 7 FPV). Two plant damage states (1FPV and 7FPV) were added to model new scenarios with a submerged RHR pump seal bypass. This, in turn, led to the introduction of a new release category, S7, which takes credit for decontamination and scrubbing in the source term dete rmina tion. Plant damage state 1FV contains interfacing LOCAs resulting f rom unsubmerged piping failures. ( Item 2. A conservatism in the treatment of certain seismically initiated sequences in release cateFory S6 (plant damage states IF, 3F, and 7F) was eliminated. In the updated results, credit was taken f or loss of instrument air to the air-operated valves ( A0Vs) in the purge lines on loss of of fsite power; hence, a high probability of purge isolation valve closures in these instances. This resulted in a shift in some of the frequency of release category S6 to S2 because, when the large purge valves are assumed to close, there remain small open lines with motor-operated valves that fail in these same sequences. There still remain some seismic sequences in S6 with the purge isolation failure. Those that remain either involve a no loss of of f site power condition or mechanical f ailure of the purge valves. Item 3. In support of the effort to optimize plant technical specifications (PLG-0431), the PSA systems modeled were revised to incorporate revision to the technical specifications and a more complete treatment of common cause failures. This led to many minor changes to individual sequence frequencies, with the most significant change being an increase to the unavailability of the primary component cooling system. This led to a slight increase in core melt frequency from 2.3 x 10 4 to 2.7 x 10 4, most of which occurred in plant damage state BD. Item 4. The updated results take credit for recovery of certain containment systems (principally the containment building spray) during core melt scenarios initiated by Icss of of fsite power and

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involving a station blackout. Both the original and updated results took credit for recovery of electric power prior to and in prevention of core melt. This new recovery action results in a small shif t in frequency from release category S3 (gradual containment overpressure) to SS (containment intact). This l change does not appreciably affect the results of the RMEPS or sensitivity studies since neither S3 nor S5 contributes to early health risk with at least 1 mile of evacuation. , Containment Model Changes Item 5. Uncertainties in source terms were reasssessed for all release categories with the net effect of a reduction in the mean source terms for,all categories. Item 6. A new release category and three new plant states for interf acing system LOCA scenarios were added. Item 7. - Interf acing system LOCAs resulting in unsubmerged RHR piping f ailures (plant state 1FV) were reassigned from release category S6 to S1 (small conservative effect). Site Model Changes Item 8. Site model uncertainties were reassessed (minor ef fect). Item 9. The evacuation distance and sheltering assumptions were varied. Item 10. thi population The Unit 2 construction workers were eliminated from distribution. Although all the above changes contribute in some way to differences in the updated results, the ones that had the most significant impact on early health risk are items 1, 5, 6, and 9. A more quantitative picture of the significance of each change is provided below. The results in the RMEPS update for early fetality risk are presented in Tables 2, 3, and 4 for evacuation cases of no evacuation,1-mile evacuation, and 2-mile evacuation, respectively. These are the comparison tables for Table 1 and the original SSPSA results. The conditional risk curves for cach release category and evacuation distance can be found in Appendix C of PLG-0432. There are two kinds of differences exhibited in the new tables. One is that the risk Icvels (exceedance frequuncy values) are lower although less evacuation is assumed, and, as expected, the levels decrease as the evacuation zone is increased from 0 to 2 miles. The other difference is that several new release categories, in addition to S6 and S2, appear as making significant contributions: S3, S7, and SI. Release category S3 contributes only under the assumption of no evacuation. This result is viewed as purely academic because the time of release for S3 is some 89 hours af ter the initiating event, during which even ad hoc protective actions would be affective. For 1 or 2-mile evacuation, release categories S2, S6, S7, and S1 are significant. The contribution of S2 only appears in the low consequence,

relatively high frequency portions of the risk curve. In comparison of these results with Table 1, the shift in the ranking of contributors is due to the following.

1. The frequency of S6 in RMEPS is lower because of the deletion of the interf acing LOCA and some seismically initiated sequences with station blackout.

2. The frequency of S2 increased slightly from the same seismic sequence noted in 1.

3. Some of the old V-sequence frequency formerly categorized in S6 is now in S7. While source terms in S7 are lower than S6, they are still great enough for potentially fatal doses.

4. The frequency of S1 increased due to the addition of the pipe break type V-sequences formerly categorized in S6 and to a smaller extent by a reassessment of some turbine missile scenarios that was done since the RMEPS.

The contributions of plant damage states and initiating events to all updated release 2ategories are shown in Tables 5 and 6, respectively. Tables 7 and 8 define the codes used for initiating events. About 78% of the scenarios in category S1 are pipe break type interfacing LOCAs that are assigned to plant damage state IFV. The remaining scenarios in S1 include aircraft and turbine missile scenarios that fail the containment in plant damage states IFA, 2FA, and 6FA and a wide spectrum of transient and LOCA scenarios with containment failure due to reactor vessel steam explosions. The contributors to category S2 are the same as those in the l SSPSA; namely, seismically induced station blackout with a failed open small penetration. Rel 'ase category S6 is now dominated by seismically induced accident sequences with no loss of offsite power and failure of the SSPS system with an assumed containment purge in progress. No credit for operator recovery of any system or component is taken for any seismic sequence, including those that now dominate S2 and S6. Release category S7 is composed wholly of new interfacing LOCI scenarios in which the RHR piping remains intact and the bypass occurs via a degraded and submerged RHR pump seal. In assessing the uncertainty on the source term for S7, a 10% probability was assigned to the possibilitly that the leak path would not be submerged. From the information provided in RMEPS , i t is clear that there would be no contribution to early health risk from S7 if only best estimate (submerged) source terms had been used. Similarly, had the conservative source terms not been used for the remaining release categories, the risk levels calculated in RMEPS would have been much lower than they were. In f act, on the basis of using the best estimate source terms only, release category S1 is the only category that produced any potential for 200-rem doses and, hence, any potential for early fatalities. 2.2 DOSE VERSUS DISTANCE CURVE In the RMEPS results, there was found to be very little potential for 200-rem doses, even close to the site. As seen in Figure 2-9 of RMEPS

(PLG-0432), the frequency of exceedance scale had to be extended f rom .001 to .0001 to pick up the mean risk of exceeding the 200-rem dose shown on the curve. The median curve for the 200-rem dose was off-scale. The contributions to the mean risk at various distances are indicated in the table below. Percent Contribution to l Release 200-Rem Exceedance Frequency Category (Figure 2-9 in PLG-0432) ,

 -                                                                                                                                         l 1 mile                     1.5 miles    2 miles' l                          l S1               1                            3          6 S2              13                           34          0 S3                67                            3          0 SS                0                            0          0 S6                17                           50         73 S7                 2                            9         20 1             1 By comparing there results with those in Table 2, it is seen that the 200-rem risk has the same set of release category contributors as the early fatality risk curve for no evacuation. Category S3 dominates at 1 mile. Again, this result is largely academic. It is difficult to envision, even if no emergency plans existed, that any ir.dividual would be in a position to receive a large dose more than 3 days after the initiating event. At 2 miles, the 200-rem curve is dominated by S6, with smaller contributions by S7 and St.

Hence, the overall picture of the risk contributors is the same for the early fatality risk curves and the 200-rem dose versus distance curves. 2.3 CONDITIONAL MEAN RISK Appendix D of PLG-0432 provides mean risk summary tables for early fatality risk as well as cancer risk.

                                                                                                                                                          )

3. SENSITIVITY STUDY UPDATE In the Seabrook Station Emergency Planning Sensitivity Study (PLG-0465, 1986), there were no changes made to the plant model. Source terms were revised to reflect the Reactor Safety Study (WASH-1400) source term methodology; 1.e., were calculated using the CORRAL computer program for the Seabrook plant configuration. These CORRAL source terms had been developed during the original SSPSA. CRACIT computer program runs were made using best estimate (median) modeling assumptions, and median accident frequencies were used for consistency with NUREG-0396 and WASH-1400. However, unlike NUREG-0396 and WASH-1400, the full treatment of dependent and external events in the Seabrook results was lef t unchanged. ,

3.1 EARLY FATILITY RISK CURVES The early fatality risk curves for 0,1 and 2-mile evacuations are

• plotted in Figure 2-1 of PLG-0465. The conditional early fatality risk curves for each release category are found in Appendix B of this report. The contributors by release category are shown in the tables below for no evacuation,1-alle evacuation, and 2-mile evacuation.

l l RESULTS FOR NO EVACUATION Release Percent Contribution to Category Early Fatality Risk Curve 1 Fatality l 100 Fatalities 1.000 Fr.talities S1 <1 <1 <1 S2 100 99 99 S6 <1 1 1 Others <1 <1 <1 Total 1 100 100 l 100 RESULTS FOR 1-MILE EVACUATION Release Percent Contribution to Category Early Fatality Risk Curve I I Fatality 1 100 Fatalities 1,000 Fatalities S1 <1 <1 <4 S2 99 95 0 S6 1 5 96 Others <1 <1 l

                                                                                                                                                                                                                        <1 l                                      l
                                           ,_ Total       i 100                                                                               !                          100                  100

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ^

_ = . -- - RESULTS FOR 2-MILE EVACUATION Percent Contribution to Release Early Fatality Risk Curve Cat ego ry l 1 Fatality 100 Fatalities 1,000 Fatalities l S1 2 2

• S2 0 0
• S6 98 96 8 Others <.1 <1
• Total 100 100 100 -

I l l 1

• Results below 10-9 per reactor year not shown.

l By comparing these results with the RMEPS results, it can be seen that one chief difference is that S2 r.ow has a more dominating impact than it did in RMEPS, especially for the 0 and 1-mile evacuation cases. Category $6 dominates the low frequency tail of the 1-mile curve and i completely dominates the 2-mile results. The other chief difference is that categories S3 and S7 no longer make a significant contribution to early health risk and the percent of contribution of S1 is reduced somewhat. These differences stem from the fact' that application of the WASH-1400 source term methodology did not have uniform impact on all the source terms. The application of this methodology appears to have increased the S2 source term more than the others. In addition, the RMEPS results for early health risk are heavily influenced by the conservative source terms used in that study. For category S7, the conservative RMEPS source term a;sumed no credit for a flooded RHR vault, while such credit was taken in the sensitivity study to make thr analysis consistent with WASH-1400. In WASH-1400, credit was taken for suppression pool scrubbing in some boiling water scenarios. In Figure 3, the 200-rem and 50-rem dose versus distance curves are compared between NUREG-0396, the Sensitivity Study, RHEPS, and the original SSPSA. scale. The 200-rem curves for the latter two studies are of f As can be seen f rom this figure, the Sensitivity Study results f ully bound the RMEPS and SSPSA results. 3.2 DOSE VERSUS DISTANCE CURVES The 200-rem dose versus distance curve is fully dominated by release category S2, with very small contributions from S6 and S1. The contri-l butions of S6 and S1 occur below the level of conditional core melt frequency at which the curves are cut off in NUREG-0396 (.001). 1

3.3 CONDITIONAL MEAN RISK Table 9 provides a mean risk summary table for the sensitivity study results. Column number 5 provides the information requested in Part C of this question. The corresponding information for RMEPS can be found in Appendix D of RMEPS. There are no results in the SSPSA that assume tither a 1-mile or 2-mile evacuation. The results presented in Table 9 confi rm that release categories S2W and S6W completely dominate the risk at Seabrook Station if the WASR-1400 source term methodology is used to define source terms. The only additional small contribution to risk is made by release category E1W for the early fatality risk with a 2-mile evacuation distance. i Figures 4 and 5 show the decrease in the early fatality risk as a f unction of evacuation distance, comparing the mean risk results from the RMEPS study (PLG-0432) and the EPZ sensitivity study (PLG-0465), which used WASH-1400 based source terms. Figure 4 compares the risk , reduction for the two cases on an absolute basis, and, in Figure 5, ~ the risk reduction is normalized to the no-evacuation case for each study. The results indicate that, for the WASH-1400 source terms (PLG-0465), the acute fatality risk decreases even more rapidly in the first 2 miles than for the RMEPS baseline case. Furthe rmore , \ goalrisk the risk. for all WASH-1400 source term cases remains below the safety

i-TABLE 1. OF EARLY FATALITIES AS CALCULA Number of Early Fatalities (percent contribution of release category) 1 10 ' 100 1,000 ' 10,000 56 (98.98) S6 (98.8) 56 (99.4) 56 (99.4) 52(0.92) S2 (1.10) 56 (99.5) S2 (0.52) 52(0.49) Others (< .1) Others (< .2) 51 (0.5) Others (< .1) Others (< .1) ((',$,"j",CY 'I Others (0) 4.60-7 3.87-7 i__ 3.14-7 1.78-7 6.26-10 NOTE: Exponential notation is ind

       .                                                        i .e., 4.60-7 = 4.60 x 10-7.icated in abbreviated form;

  , =

l i,

 .i TAHLE 2.

FATAtlTIES BASED ON RMEPS UPOA Number of Early Fatalities (percent contribution of release category) 1 10 100 ' 1,000 10,000 52 (40.1) $3 (62.7) S3 (50.0) S6(67.6) 53 (35.4) S6 (27.6) S7(98.5) ! 56 (32.7) 57 (24.9) 51 (1.5)

                                       $6(18.3)        S7 (6.7)       S7(13.3)              51 (7.5) i Others (< 7)    Others (< 3)                                                                 ,

Others (< 4) Others (= 0.0) Ilp.I. t. .I Others (= 0.0) Frequency at 1.40-/ Exceedance 7.91-8 2.98-8 4.41-9 2.53-11 gg,, 4. 0-7 3.o7 7 3.>4-7 i.70-7 '

                                                                                                               . 2.-10 NOTE:                                                                                      i Exponential notation is indicated in abbreviated                  .e.,

form; 1 i 1.40-7 = 1.40 x 10-7 O I

 =

1 . TABLE 3. CONTRI8UTIONS OF RELEASE CATEGORIES TO RISK OF EARLY FATALITIES BASE 0 CN RMEPS UPDATE MILE EVACUATIO  ! l Number of Early Fatalities (percent contribution of release category) I 10 100 - 1,000 10,000 lS2(71.7) S6 (65.5) 56 (50.7) S7 (62.4) S7(98.5) S6 (18.6) S7 (26.4) 57 (41.2) 56(28.3) S1 (1.5) S7 (7.4) S1 (8.1) S1 (8.1) S1 (9.3) Others (< 3) Others (= 0.0) Others (= 0.0) Others (= 0) Up la t ...I Others (= 0) Frequency of 1.44-8 1.64-8 8.25-9 Exceedance 1.59-9 2.53-11 SSPSA Results 4.60-7 3.87-7 3.14-7 1.78-7 6.26-10 i NOTE: - g Exponential notation is indicated in abbreviated form; f.e., 7.44-8 = 7. 44 x 10-8 , if tr .-

i 8 j 3 TABLE 4. i CONTRIBUTIONS BASE 0 DN RNEPSOF RELEASE

                                                               !!POATE                  CATEGORIES MILE     S EVACUATION                      TO20

\ Number of Early Fatalities (percent contribution of release category) i 1 10 100 , j 1,000 10,000 S2(46.0) S7 (55.2) S7*(65.3) 56(72.9) S7 (43.3) 56 (33.3) 56 (23.9) 57 (19.1) 51 (10.7) 51 (11.5) 1 S1(10.8) $1 (8.0) 1 Others (= 0) Others (= 0) 6 Others (= 0) Others (= 0) tip.t.. ...: Frequency of 4.96-9 Exceedance 3.07-9 1.11-9 2.28-10 0.0 gg,, 4.60-1 3.er-, 3.1<-> l.i ., 6.u-10 NOTE: y Exponential notation is indicated in abbreviated form; i e

                                                                       . ., 4.96-9 = 4.96 x 10-9

1 i i e i e i 1 i j ,; l - i > l I TABLE 5. i MAKING MAJOR CONTRIBUTIONS TO RISK RMEPS UPDATE RESULTS 1 l { Major Risk and Core Melt Frequency Contributing Release Categories 51 (percent contribution of plant damage states to release categories) i 52 53  ! ' 55 56 S7 IFV (77.5) l 7FF (49.5) 80 (72.0) 8A (82.1) { 3F (92.5) IFA (7.3) 3FP (43.8) IFPV (69.5) 5 F0 (14.6) 4A (15.0) 8A (5.0) 7F (7.3) 7FPV (31.5) IFP (6.6) 30 (11.3) 2A (1.6) 2FA (3.7) i 3 6FA (1.3) i i Others (5.2) Others (< 1) Others (< 3) Others (< 2) j Release Others (<'l) Others (0.0) j Category 6.00-9 '. Frequency 2.02-5 1.43-4 - 1.17-4 3.00-7 1 3.93-8 NOTE: Exponential notation is indicated in abbreviated form; ie

                                                                                                 . . 6.00-9'= 6.00 x 10-9

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TABLE 7. BINNING OF INITIATING EVENTS TH New Initiating Events Title Binned SSPSA initiating Events Frequency (events / year) Title Frequency (events / year) EXTAC 2.70-6 FSRAC FCRAC 5.19-7 2.10 6 FL2SG EXTLP 8.50-8

• 1.20-3 FTBLP FLLP 6.00-4 3.20-4 TCTL EXTCR 2.76-4 5.43 7 TMCR MCR 3.98-7 ACR 5.80-9 TLPCC 1.39-7 1.82-5 LPCC 1.39-6 FSRCC FCRCC 3.60-6 9.00-6 FPCC 4.20-6 TMPCC MPCC 1.27-8 .

TLSW 5.46-9 6.22-6 LOSW FCRSW 2.52-6 2.10-6 FLSW TLCV 1.60-6 4.18-1 LCV 4.18-1 TMLCV 8.30 5 NOTE: . Exponential i.e., 2.70-6 = 2.70notation x 10-0 is indicated in abbrenated form; i o

TABLE 8 INITIATING EVENT CATEGORIES SELECTED FOR QUANTIFICATION OF THE SEA 8R00K STATION RIS

                                                                           ~

i Group Sheet 1 of Initiating Event Categories Selected for Separate Quantification Code Designator e Loss of Coolant 1. Excessive LOCA Inventory 2. Large LOCA ELOCA 3 Medium LOCA LLOCA 4 Small LOCA MLOCA 5 Interfacing Systems LOCA SLOCA(a) 6 V Steam Generator Tube Rupture SGTR(a ) - e General 7. Reactor Trip Transients 8. Turbine Trip RT

9. TT(b.)

Total Loss of Main Feedwater TLMFW(C)

 '                                   10. Partial Loss of Main Feedwater

11. PLMFW(C)

Excessive Feedwater Flow EXFW(b)

12. Loss of Condenser Vacuum LCV(b)

13. Closure of One Main Steam IMSIV(b) 14 Isolation Valve (MSIV)

Closure of All MSIVs AMSIV

15. Core Power Excursion

16. Loss of Primary Flow CPEXC

17. LOPF(b)

18. Steam Line Break Inside Containment SLBI

19. Steam Line Break Outside Containment SL80 20 Main Steam Relief Valve Opening ' MSRV Inadvertent Safety Injection SI e Common Cause Inittattna Events

            - Support System Faults      21.

22 Loss of Offsite Power LOSP(d) Loss of One DC Bus L1DC 23. 24 Total Loss of Service Wa:er LOSW Total Loss of Component ;ooling LPCC Mater

           - Seismic

25. 0.7g Seismic LOCA Events 26 E.7L 1.0g Seismic LOCA

27. 0.2g Seismic Loss of Offsite Power E1.0L

28. 0.3g Seismic Loss of Offsite Power E.2Tle)

29. 0.4g Seismic Loss of Offsite Power E.3T(e)

E.4T(e) a. b.

c. Transient without scram scenarios are,, represented ASLOC.

ATT.

d. Transient without scram scenarios are represented by a separate code

e. Transient without scram scenarios are represented by a ,,separate ALOMF.

ALOSP. code xTransient

           = .2, . 3, . without  scram 4, .5, . 7,  1.0. scenarios are represented by a ,separate ExA,                    code j

_________.____.______...___J

i TABLE 8 (continued) Group Sheet 2 e Initiating Event Categories Selected '

                                    -                                            for Separate Quantification                           Code Designato 30                                                                    _

31 0.5g Seismic Loss of Offsite Power 0.7g Seismic Loss of Offsite Power E.5T(e)

32. E.7T(e) i- - Fires 33 1.0g Seismic Loss of Offsite Power Cable Spreading Room - PCC Loss El.0T(e) 34 Cable Spreadin FSRCC

35. Control Room g Room - AC Power Los s FSRAC

36. PCC Loss FCRCC 37 Control Room - Service Water Loss FCRSW 38 Control Room - AC Power Loss Electrical Tunnel 1 FCRAC

39. Electrical Tunnel 3 FET1 407 PCC Area FET3

41. FPC,C Turbine Building - Loss of Offsite Power FTBLP

                  ~
                                    - Turbine                       42 Missile                     43        Steam Line Break Large LOCA                                         TMSLB 44                                                          TMLL Loss of Condenser Vacuum

45. Control Room Impact TMLCV 46 THCR 47 Condensate Storage Tank Impact Loss of PCC TMCST

                                  - Tornado                                                                                 TMPCC i                                                                   48.
!                                     Missile                             Loss of Offsite Power and One              . MELF Diesel Generator

49. Loss of PCC

50. Control Room Impact MPCC MCR

! - Aircraft $1, Containment Impact Crash 52 Control Room Impact APC 53 ACR Primary Auxiliary Builetng Impact APA8

- Flooding 54 55 Loss of Offsite Power

' FLLP Loss of Offsite Power aas One Switengear Room i FLISG

56. Loss of Offsite Power a 3 Two Switchgear Rooms j 57 FL2SG 4

i LossPumps Wat*- of Offsite Power a 2 Service FLSW i - Others S8. ! i Truct Crash into Transmission LinesTCTL

a. .

b.

c. Transient without scram scenarioso e,e,are ASLUC.

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e. Transient without scram scenarios are represented e code, ALUMF.

e code, ALOSP. by a sep xTransient

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RAI 24

Provide a quantitative description of the effects of the following differences between the original PSA and the current study:

a. reduction in probability of core-melt V sequences

b. factor of 1000 scrubbing of releases through RHR seals

c. change of release category (S6 to SI) for unscrubbed event V sequences. .

The earlyeffects should fatalities be 200 and for described rem vsindistance. terms of differences in risk curves for RESPONSE 24

Reference:

Response 23 Re: 200 REM and early fatalities In response to part a) of question 24, the following highlights the key factors that result in a major reduction in risk levels for the core melt V-sequence in the updated Seabrook probabilistic safety assessment (PSA) results [per the Risk Management and Emergency Planning Study (RMEPS), PLG-0432,19858 in comparison with the Seabrook Station Probabilistic Safety Assessment (SSPSA) (PLG-0300, 1983). Qualitatively, the key differences fall into three main areas of the analysis: initiating event f req ue ncy. The response to 24 b. will be provided in the response to 30. plant response to various types of interf acing loss of coolant accident (LOCA) scenarios, and operator actions to prevent core melt and isolate the bypass. Description of each of these areas follows. 6

1. INITIATING EVENT FREQUENCY The initiating event frequency model in the SSPSA cons 4dered four residual heat removal (RNR) cold leg injection paths, each having two 4

series motor-operated valves (MOV). series check valves, a The check valve model considered outboard check valve. successive, independent ruptures of, first, the inb The second failure was assumed to occur at the the next test (refueling).same rate as the first at any random time ruptures (inboard, then outboard), as well as the possibility outboard valve is already open when the inboard valve fails. model included all the above failure modes plus several more. The updated check valve failures, it was conservatively assumed that the For the outboard + inboard. sequence could also occur and that it would occur at the same rate as the inboard + outboard sequence. In addition, the possibility of instantaneous failure of the second valve in a sequence at the time the first valve failed was also considered. Hence, the model used in the update is more complete. The net'effect of those model enhancements given leak size. is worth about a factor of 2 increase in the frequency of a A second initiating difference between the two studies was the definition event. a major rupture leading to RHR overpressurization.In In the RMEPS, any the SSPSA pump) was considered an initiator.ruoture with a leak flow exceeding Such flows are not capable of overpressurizing properly. the RHR system when the RHR relief valves operate A third of checkdifference valve data. in the initiating event frequency was in the treatment - The SSPSA used check valve rupture data--actually zero failures in a large sample of component hours per population--taken from the Indian Point 2 and 3 PSAs. In RMEPS, a different approach was based on a frequency-magnitude correlation of nuclear grade RHR and reactor coolant system (RCS) check valve experience in U.S. pressurized , water reactors (PWR). separate submittal. These data are documented in RMEPS and in a RHR overpressurization events can be compared as follo SSPSA: 1.8 x 10-6/ reactor year. RMEPS Update: 7.1 x 10 / reactor year (leak > 1,800 gpm). Thus, the net effect of the model differences (which have an increasing effect in the update) and the data treatment (which has a decreasing effect in the update) is a reduction in the frequency of valve ruptures leading to RHR pressurization by a factor of 2 to 3. Hence if no other changes would have been made to this analysis, the V-sequenc,e risk contribution decreased by(and its early this same release frequency contribution) would have factor. , 1 d 1 J

PLANT RESPONSE 2 The plant response to RNR interfacing valve ruptures in the SSPSA and in most previously published probabilistic risk assessments (PRA) on PWRs has been treated rather simply, according to the following assumptions, without consideration of their incremental probability. e Valve ruptures produce a shock wave with peak dynamic pressures significantly greater than the RCS pressure that travels down the low pressure RHR piping. e RHR piping ruptures outside the containment. e RCS and refueling water storage tank (RWST) inventories leak outside the containment via a piping break. e Core melt occurs with unsubmerged bypass. e No credit is taken for any operator actions. Therefore, as a single the sequence. plant response to the V-sequence in the SSPSA was treated in turn, was assigned to release categoryInS6.It was assigned to plant d the RMEPS update, a large number of alternative scenarios were identified to provide a more complete picture of plant response. The most important variables introduced in the update to consider alternative plant responses are the size of the valve ruptures that initiate the event, the response of the RHR relief valves inside the containment, the pressure capacity of RHR low pressure piping, the response of RHR pump seals to overpressure, and the configuration of the RHR pump vaults with regard to source term implications. Tne major differences in plant response to RHR interfacing valve ruptures, in Figureas 1.modeled in the SSPSA and the RMEPS upcate, are illustrated This figure is a highly simplified version of the event sequence model that was developed and quantified in the RMEPS update. , The chief differences in the update in this regard are a lower frequency of unsubmerged pipe rupture-type bypasses because of the high capacity of the RHR piping. A more likely outcome is a submerged bypass via the RHR pump seals. 4 f 1 I

OPERATOR RESPONSE

                                                                                                                )

Because of a different treatment of hardware and plant response, the potential for operator actions to mitigate the effects of the interfacing valve ruptures was appropriately considered in the update. The two key actions, to isolatewhich the bypass. are illustrated in Figure 1, are those to prevent melt and chance, as assesed in RMEPS, that the operators would prev whether or not the bypass was isolated. The key is to diagnose the bypass the RWST. at the RHR pump seal and to provide long-term make'up of coolant t If this action is not successful, there is some chance that the operator valve can isolate rupture case (VI). the bypass, but only for the discharge check The net effects of the major update factors are sunenarized in Table 1. e l t i 9

                                                                                                                                                                      ~

TABLE 1. IMPACT OF KEY FACTORS IN UPDATED V-SEQUENCE ANALYSIS b Factor SSPSA RMEPS Frequency of RHR 1.8 x 10 4 7.1 x 104 System Overpressurization r Reac r y ,p p,,g y ,p Percent of Overpressurization

• Events that End with:

e No Core Melt 0 93 e Melt with No Bypass 0 ~1 I e Melt with RHR Seal Submerged 0 Bypass ~5 e Melt with Unsubmerged RHR Pipe 100 Rupture Bypass ~1 1 l l l

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RAI 25

Provide a list of all paths for loss of RCS inventory outside containment. Show how these have been considered with respect to LOCA and with respect to containment bypass for radioactive materials following core damage. 9 RESPONSE 25 Loss of RCS inventory outside containment could occur via a flow path which directly links the RCS to systems outside containment or in conjunction with the secondary side of the plant assuming steam generator tube rupture has occurred. In either case, a specific containment penetration can be identified as being associated with that flow path. The Containment Penetration Analysis (PLG-300, Table D.13-3) in the SSPSA accounts for all such penetrations and either provides or references a description of the equipment f ailure or saloperation necessary to 4 bypass containment. Penetrations associated with flow paths which 4 would be operating during an ac:ident are also listed. Penetrations which are associated with flow paths that could line the RCS directly with systems outside containment are: X-9, 10, 11, 12, 13, 24 through j 31, 33, 35, and 37. ! Penetrations which are associated with flow paths that could link the RCS with systems outside containment assuming a steam generator tube rupture are: X-1 through 8, 63 through 66. An evaluation of LOCA outside containment was performed for the SSPSA, , examining each line which communicated with the RCS and penetrated the cont ainme nt. Based on that evaluation, six lines were considered in detail - four RHR cold leg injection lines and two hot leg suction Ifnes. (See SSPSA 56.6 3.2.1). These are the classic "V-sequence" lines first discussed in WASH-1400. This evaluation of LOCA outside containment was not documented explicitly in the SSPSA but can be reconstructed from the SSPSA GD.13. Table D.13-3 contains a list of all the containment penetrations with the related , isolation valves and affected system. In order to be of interest in the l evaluation of LOCA outside containment, the line must not only penetrate i the containment to the atmosphere but also communicate with the RCS. Each penetration is discussed below. Reference is made to Table D.13-3 in SSPSA, PLG-300. l i I

LOCA OUTSIDE CONTAINMENT Penetrations Cont ai nment Quantified Penetration System in SSPSA Comments X-1, X-2, Main Steam Yes Quantified in 55.3.11, X-3, X-4 Steam Generator Tube Rupture with secondary side leak to atmosphere. X-5, X-6 Main Feedwater No See Table D.13-3 X-7, X-8 " Comments" X-9, X-10 RCS/RHR Yes Quantified in 56.6.3.2.1, RHR hot leg suction. X-11, X-12 RHR Yes Quantified in $6.6.3.2.1, RHR cold leg injection. X-13 , RRR , No Discussed in 56.6.3.2.1, RHR hot leg injection. X-14, X-15 Containment No No direct communication Building Spray with the RCS. X-16, X-18 Containment No No direct communication Online Purge with the RCS. X-17 Equipment Vent No No direct communication System with the RCS. X-20, X-21, PCCW No No direct communication X-22, X-23 with the RCS. X-24, X-25, Safety Injection No Each line has at least X-26, X-27 two check valves and one normally closed MOV in series. X-28, X-29, CVCS No See Table D.13-3 X-30, X-31, " Comments. X-33 1 X-32, X-34 Floor and No No direct communication l Equipment Drain with the RCS. X-35 SI No 3/4" test line, has two normally closed A0Vs in series. RCS No 1/2" sampling lines X-36 Deminieralized No No direct communication l Water with the RCS. l l e l L

Penatraticns Containment Quantified Penetration System in SSPSA Comments Nitrogen Gas No One check valve and four normally closed valves. Reactor Makeup No Two check valves and two Water normally closed valves. X-37 CVCS No See Table D.13-3

                                                        " Comments".

X-38 Combustible Gas No No direct communication Control Fire with the RCS. Protection

     -  X-39          Spent Fuel Pool          No       See Table D.13-3 Cooling and Cleanup              " Comments".          -
   . X-40        '

Nitrogen Cas , No No direct communication with the RCS. RCS Sampling No No direct communication with the RCS. X-4 3, X-47 RCS, SI No See Table D.13-3 X-50 " Comments". X-52 Post Accident No See Table D.13-3 Monitoring " Comments". X-57 SS, SI No Ste Table D.13-3

                                                       " Comments".

X-60 CBS No See Table D.13-3

                                                       " Comments".

X-63, X-64, S/C Blowdown Yes Quantified in 15.3.11.4 X-65, X-66 Steam Generator Tube Rupture with secondary side leak. See top event IV (p.5.3-95). X-67 Service Air No No direct communication with the RCS. X-71, X-72 Combustible Cas No No direct communication Control with the RCS. l HVAC-1, Containment Air No No direct communication HVAC-2 Purge with the RCS. I

Nominal Diameter Penetration # Isolation Valve Inches Comments X-72 CGC-V-10 1" Hanual locked closed valve that isolated Train A H2 analyzer input line. X-72 CGC-V-3 l' Hanual locked closed valve that isolated Train A H2 analyzer return line. X-39 SF-V-86, 87 2" Hanual locked closed valves. Would only be utilized in conjunction with the refeeling canal skimmer pump.

 ~

X-67 SA-V-229,Id42 2" Hanual locked closed valve outside containment, normally closed valve inside containment. Note: Test connections were not considered in this analysis. Test connections are 3/4" with normally closed manual valves and a pipe cap. Per the Seabrook Station Technical Specifications there are verified closed every 31 days. o

l 1 RAI 26 I Indicate the extent to which the effect of local deflagration / detonation of hydrogen gas concentration is localised areas both inside and outside the containment has been considered in the assessment of risk. Include a discussion of how weak areas of containment have been considered in your assessment, for example, the containment is considerably weaker in its resistance to pressure loading from outside the containment.

;                               RESPONSE 26                                                                                                          4 j

A separate probabilistic analysis of the effects of hydrogen combustion has 4 been performed for each plant damage state. The analysis accounted for uncertainties in the hydrogen generation, release from the primary system, ignition, and containment atmospheric conditions. Vessel breach discharge j

burns and global burns at different times in the accident progression were

• treated separtely. All hydrogen burns were treated as instantaneous adiaba- 1 tic combustion events. The analysis is documented in Section 11.5.2 and in Tables 1.7-1 and 11.7-7 of the SSPSA (PLG-0300). Local hydrogen deflagra-

!                               tions or detonations were considered and dismissed as requiring conditions i

of nearly stagnant or quiescent atmospheres which are not considered cre- i dible under accident conditions. In a large dry containment thermal and l mass transfer induced mixing of the containment atmosphere under accident  ! j conditions is considered assured particularly on those accident phases where t rapid releases of hydrogen into the containment are possible such as at vessel breach. i f Weak areas in the containment with respect to the capability to contain hydrogen burns could not be identified. A maximum adiabatic post burn pre-sure of 128 psia was determined for a limiting vessel breach discharge burn. j The lowest containment failure pressure was identified at 181 psia for the type A failure (0.5 square inch leak area). Hydrogen burns outside the containment in the annulus between the primary j and secondary containment could be postulated. Such hydrogen burns would 4 have no impact on the calculated risk or cc,nclusions with respect to emergency planning requirements for two reasons: i 1. In order for hydrogen to accumulate to a flammable mixture in the annu-j lus region, the hydrogen must be released from the containment and this i i requires that the containment is already failed. Furthermore, the con- , centration in the annulus would be lower than in the containment due to ' l the additional mixing with the annulus air. I i I 2. . i A hydrogen burn in the annulus would impose an equal load on the enclo-  ; I sure building and on the containment building. Even though the contain-1 ment may be weaker for external loads than for internal loads, the j external load capacity is definitely much areater than .the internal load

!                                   capacity of the containment enclosure bui13ing which was evaluated in the $$PSA Appendix H.1, Section 6. The weak elements in the enclosure                                            c 4                                    pressure boundary are identified as the HEPA filters and the sheet metal l                                    duct work, each of which is expected to f all at a pressure between 1 and                                        ,

! 2 paid. 1  ! 1 l .' i

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

l

RAI 28

i Identify any penetrations connected directly into the containment atmospheric ' which rely on any remote manual or manual valves for isolation. 4 Response 28 ' Nominal Diameter

Penetration # Isolation Valve Inches Comments

. X-14 CBS-V-11 8" To CBS spray rings. Normal t X-15 CBS-V-17 8" closed motor operated valves i open on Hi-3 containment pressure or manual. Valve closure is ' manual.

       .                        X-60                      CBS-V-14          16"                      Containment sump return line MOV.

4 X-61 CBS-V-8 16" l I Normally closed. Valves open automatically on ECCS/CBS recircu-lation signal indicated by 2/4 ' Lo-Lo level in the RWST. Valve closure in manual. . X-38 CGC-V-43 2" Compressed air line to containment OGC-V-44 2" for pressure testing. Manual locked CGC-V-45 10" closed valves. Line is used only for containment pressure testing. ! X-38 FF-V-592 4" Manual locked closed valve outside containment, normally closed valve inside containment. X-71 CGC-V-36 2" Manual locked closed valve. This valve is in series with CCC-V-28 located inside containment which  ! will isolate on a containment t isolation signal. X-71 CGC-V-32 1" Manual locked closed valve that isolates the Train B H2 analyser , input line. 1 X-71 CGC-V-24 1" Manual locked closed valve that . isolates the Train B H2 analyzer return Ifne. 1 X-72 CGC-V-15 2" Manual locked closed valve. This 1 valve is in series with CCC-V-14 i located inside containment which will isolate on a containment

isolation signal. -

l e j i

Nominal Diameter Penetration # Isolation Valve Inches Comments X-72 CGC-V-10 1" Manual locked closed valve that isolated Train A H2 analyzer input line. X-72 CGC-V-3 1" Manual locked closed valve that isolated Train A H2 analyzer return line. X-39 SF-V-86, 87 2" Manual locked closed valves. Would only be ut111:ed in conjunction with the refueling canal skimmer pump. X-67 . SA-V-229, 1042 2" Manual locked closed valve outside containment, normally closed valve inside containment. Note: Test connections were not considered in this analysis. Test connections are 3/4" with normally closed manual valves and a pipe cap. Per the Seabrook Station Technical Specifications there are verified closed every 31 days. I

l l l

RAI 33

Confirm that a complete and independent check will be perf ormed for the containment strength calculations that served as the basis for the EPZ sensitivity study. . l i l RESPONSE 33 A complete and independent check will be performed for the containment strength calculation. This ef fort will be completed on approximately November 25, 1986. e i I 9

l RAI 34 - Fully address the ef fect of uncertainty in the ultimate strength of Cadweld splices on the pressure capacity of the containment. As discussed in the meeting your response the Cadweld splices. should address potential, non-ductile f ailure of l 1 RISp0NSE 34 See response to RAI 8. 4

RAI 35

Assess the response of the containment sump encapsulation vessel on the containment integrity. ( RESPONSE 35 l l The sump encapsulation vessel is not a part of the primary containment pressure boundary. See attached drawings for cenetration X-60 (typical of X-61 also) 9763-SL-X60-01 9763-D-801212 l Inside containment, CBS line 1212-1-151 is welded by partial penetration welds to a flat plate penetration closure. In turn, the flat plate is welded to the l penetration sleeve. Containment pressure is present in the piping up to l Motor Operated Valve CBS V-14. The encapsulation vessel surrounds this valve and is present only as a secondary boundary to contain valve stem leakage, if any. l This penetration is also not subject to the large shell deflections estimated l in the $11A ultimate capacity analysis since the line penetrates containment l below t he basemat elevation, (-31'-6"). I

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RAI 40

What indication are available if RHR is lost during shutdown (e.g. spurious closure of suction valve)? RESPONSE 40 This is dependent primarily on the type of f ailure that causes the loss of the Residual Heat Removal System. The list below indicates some various alarms and indications for two scenarios; 1 RHR Pump trip on overcurrent (faulty overcurrent relay) RHR PUMP TRIP BREAKER LOCKOUT alare on the Video alare system (VAS) , 2 Spurious f ailure CLOSED of an RHR suction valve. RHR LOW-LOW FLOW alare'on the VAS (335 GPM) RHR LOW FLOW on overhead alare annunciator (555 GPM)

          -     RHR HEAT EXCHANGER LOW DISCHARGE FLOW alare on the VAS (1500 CPH)

Abnormal current indication on the Main Control Board Valve Position on the Hain Control Board Other indications and alares of a malfunction in the RHR ava!!able to the operator at the Main Control Board are; INDICATIONS (for both ' A' and 'B' train) RHR pump asperage (MCB and MPCS) RHR pump discharge flow (MC8 and MPCS) RHR pump discharge pressure (HC8 and MPCS) RHR heat exchanger inlet tesperature (MPCS point) RHR outlet exchanger outlet temperature (NPCS point) RHR pump motor winding temperature (NPCS point) RHR pump radial bearing temperature (NPCS point) RHR pump thrust bearing temperature (MPCS point) RHR pump PR1HARY COMPONENT COOLING WATER flow (MPCS point) RHR heat enchanger PR1HARY COMPONENT COOLING WATER flow (MPCS point) ALARMS (for both 'A' and 'B' train) RHR pump motor winding high temperature (VAS) RHR pump radial bearing high temperature (VAS) RHR pump thrust bearing high temperature (VAS) RHR pump discharge pressure high (VAS) RHR pump PR1HARY COMPONENT COOLING WATER low flow alare (VAS) RHR heat enchanger PR1HARY COMPONENT WATER low flow alare (VAS)

I  ! L l ,

RAI 41

What indication is available for vessel level during shutdown and refueling modes? l RESPONSE 41 As the plant is brought out of the standby condition and into hot shutdown and cold shutdown, the Reactor Vessel Level Instrumentation Systes (RVLis) indicates at the esin control board the level in the vessel. This systes is available both with the RC pumps operating (dynamic-range) and without the RC pumps running ( full-range). The RVLIS has two electrically-independent trains of instrumentation and indication. The RVLIS is not available when the head is off the vessel. During shutdown when the vessel head is of f the reactor vessel, or the RCS is otherwise open to the atmosphere, two methods of vessel level determination are available. First, a clear plastic tube is connected to one of the loop drain lines and routed vertically to an elevation above the reactor vessel head to serve as a level guage. The RCS is drained either to just below the vessel flange (refueling) or to the midpoint of the loop piping (steam generator esintenance). The loop drain line connection is made at the bottoa point of the RC piping. Therefore, this method of toeparary level measurement is available throughout the range of draining required f or all eatntenance which may take place while fuel is in the vessel. The second method of vessel level detereinstion is by level transitter RC-LT-9405 which indicates on the NCB (SF) by RC-L1-9405. This method is also available only when the RCS is open to atmospheric pressure, and is valved-out during normal operation. A final metod of vessel level detereination is by the Ref ueling Cavity Level Transeitter SF-LT-2629. SF-LI-2629 indicates on the NCB (BF). This method is available and applicable only when the refueling cavity is in communication with the RCs: that is, when the cavity level is above the reactor vessel flange with the head of f and the f uel transfer tube gate open.

RAI 42

Does loss of power to the pressure transmitter that provides input to the autoclosure interlock for RHR suction valve cuase the valves to close? RESPONSE 42 Loss of power to the pressure transmitter that provides input to the autoclosure interlock for RHR suction valve will not cause the valve to close.

4 l l

RAI 43

To what level (s) is the RCS drained for maintence activities while shut-down with fuel in the vessel? What level is necessary to maintain connection with the ultimate heat sink? RESPONSE 43 For maintenance and refueling the RCS may be drained to the level required for the activity. The two typical stopping points are 6-inches below the reactor vessel head and to the midplane of the reactor coolant loop piping for steam generator, reactor coolant pump and valve maintenance. The midplane of the reactor coolant loop piping is the minimum level to ensure proper operation of the Residual Heat Removal System. Estimates of the time that night be expected at the different water levels based on Zion data is provided in response to RAI 21. r _..,,._-._-r- -

RAI 44

Describe the availability of the SI pumps while shut down. How difficult would it be to restore the SI function to respond to transients during shutdown and refueling conditions? Consider maintenance of the SI system in your response. k RESPONSE 44 The Safety injection (SI) pumps (intermediate head safety injection) are to be SAFETY TAGGED (with a CAUTION TAG) with the motor circuit breakers open and rack-out within four hours after entering H0DE 4 (average RCS temperature <325 degrees F but >200 degrees F) Ojt any one of four RCS cold legs temperatures <325 degrees F, whichever comes first. To restore the SI function of these pumps, the SAFETY TAGS aust be removed f rom the breaker (s) and the breaker (s) racked into the OPERATE position. They may then be started manually or automatically by a Safety Injection Actuation Signal. A more complete discussion of this subject is in response RAI 21. ~ 1 5

4

RAI 45

Provide the procedure for establishing cold overpressure protection when shut ting down. RESPONSE 45 Cold Overpressure Protection at Seabrook does not require manually arming. It is armed automatically on Reactor Coolant System temperature decreasing to less than 342 degree F. The attached illustration below shows the logic employed in arming and operating the LTOP system (Low Temperature Overpressure Protection) during an RCS cooldown and depressurization-f or Protection Train ' A', Train 'B' logic is similar. PORV (Power Operated Relief Valve) Block Valve auto-open (NOTE: this is a NORMALLY OPEN valve.); Power available to valve AND valve control in remote AND valve control in automatic AND LTOP train 'B' armed (suctioneered low wide range RCS cold leg temperature is less than 342 degrees F.) PORV Train ' A' will auto open if; Valve in automatic AND Wide Range Reactor Coolant pressure channel (PT-403) is greater . than the LTOP setpoint (see attached graph, Protection Train 'A' LTOP setpoint generated by suctioneered low RCS hot leg temperat ure.)

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QUESTION 46 1s the primary system made water-solid during shutdown? RESPONSE 46 There is at present no plans to operate the RCS for long periods of time in the water solid condition. If the plant is shutdown to refuel, or for other maintenance, the pressurizer will be taken solid per operating procedure to collapse the steam bubble, then drained to the required level. Another condition when the RCS is water solid would be when drawing a pressurizer steam bubble following filling and venting of the RCS. This would only be a transitory condition. Another possible mode of operation is to have the RCS filled and vented to the PRT.

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RAI 49  ; The FSAR gives RHR relief valve flow rate as 900 sps with a set pressure of 450 psi. The flow rate does not agree with the value used in Reference 1, section 3, page 6. Please explain. D RESPONSE 49 The correct rated flow of each RHR relief valve is 900 GPM at 450 PSI. Although the value of 990 GPM stated is incorrect, the V-sequence probability modeling used a total of 1800 GPM for both relief valves. It was assumed that a valve rupture flow of greater than 1800 GPM would result in RHR

pressurization to 2250 psia. For the flow calculations in MAAP a pressure dependent flow capacity was used as described in RMEPS. This is somewhat conservative since the initial driving pressure is over 2000 PSI.

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RAI 50

Please describe the mechanism for assuring that plant changes and new knowledge form a part are promptly factored into the technical considerations which of the foundation emergency planning zone radius.for staff consideration of a reduced RESPONSE 50 The PSA vill knowledge. be kept up-to-date with regard to plant changes and new significantlyWherever affect major plant modifications are performed which the PSA, the report will be updated accordingly. To ensure prompt and effective input to plant change decisions, the  ; PSA will be considered in the design change review process. PSA considerations will include the impact on public health risk. The RMEPS results already reflect some of the changes to plant Technical Specifications that have been made since the SSPSA was completed in 1983. Recently, to a review the plant since has been completed by the PSA team of changes made 1983. the current PSA results in any significant way.No changes were identified t j G

RAI 51

Reference 1, page 3-7, paragraph 5) references both high and low level sump alarms. What is s sump low level alarm? RESPONSE 51 The CBS pump cubicle sump pump is tripped on low sump level and an alarm is annunciated at the waste management system control panel (CP-38A) and remotely in the main control room via the VAS. The pump trip and low level alarm are set to protect the pump from cavitation. I l l i

                                                                                                                                                                                                                  +

2 4 .4-- + 4 .- < m RAI 53 i Reference is made on page 3-7 to the RHR system crosstie line and RHR l systemlines. bypass respor:se due to flow in this line as well as in the miniflow The conclusion is drawn that the RHR system pressure will tend to be unifore as a result. realistic? What is the impact Are flow conditions such that this is to the discussion. of this assumpation on conclusion pertinant { Response 53 Page 3-7 states that

• uniformly af ter valve f ailure (neglecting the time it...the entire RHR system takes for pressure valves to traverse the system)".

provides a diagram of the, RHR system.The figure included with this response The configuration of the Residual Head Removal System when it is aligned as part of the emergency core co fsystem rom theisref forueling both RHR water pumpsstorageto tank. be aligned through separate suction paths The discharge side of the pumps are ' connected downstream of their respective heat exchangers by an 8-inch crosstie line containing two open motor operated valves. Each pump has a the sinfaum pump suction flow if piping. ne that connects the outlet of the RHR heat exchanger to a flow restricting orifice (750 GPM atThe minaus flow line is a 3-inch pipe containing operated valve;- 185 PSI) and a normally open motor water storage rank, the containmentIn the RHR pump suction piping, branches from 2 the RHR pump so.-tion pipe. It sump and the reactor coolant loop join coolant loop that contains the 3x4 inch relief valve setis the pump suction pipe a rated flow of 900 gps. at 450 psig with provides additional details of RHR flow paths. Table 4-6 and Figures 3-1 and 3-Except for the RHR pipe failure case (which maps to the most severe release category, 51), the interfacing LOCA flow rates through to RHR system are expected to be moderate (a few thousand GPM) relative to the RHR piping i The pressure losses through the RNR system willthenot be lar pressure should be fairly uniform. We see no algnificant " impact of this assumption on conclusions pertinent to the discussion" for the following reasons: 1) As discussed throughout PLG-0432, the RHR system overpressure failuresmodes. failure modeled conservatively envelop the spectrum of possible 2) The RHR pressure boundary failure (s), if any, are expected to be caused by the initial pressure wave or pulse, not f rom flow induced s pressure drops. ' 3) i The precise pressure distribution through the RHR is of little leportance since the risk results are not extremely sensitive to the exact RCS blowdown /RHR pressur2astion rate. 1

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RAI 55

What is the justification for the statement on page 3-10 that the first sign of trouble will be pressurizer low level or low pressure alarms? We suspect a number of other indicators may be first, such as abnormal indication f rom the PRT or even a smoke alarm. RESPONSE 55 The indications of an RHR Interf ace LOCA would primarily depend on the magnitude of the inleakage f rom the RCS as was stated on page 3-10. If the RCS to RHR leak flow rate is less than the capacity of the RHR suction relief valves, the first indications the operator may receive are;

                                   -             PRT (pressurizer relief tank) high pressure
           ~
                                   -             RHR discharge pressure seters readings abnormally high, 450 PSIC (0-700 PSIG scale)
                                   -             RHR heat exchanger outlet temperature recorders pegged high (0-500 Degrees F)
                                   -             Pressurizer Level low (-5% deviation f rom programed level)
                                   -             Pressurizer pressure low (2210 PSIG)

For in-leakage less than 1800 GPM, the RHR system would pressurize to approximately 450 PSIC (the setpoint of the suction relief valve) and stay at this point. The RHR suction relief valve (s) would cycle to maintain the RHR system at 450 PSIG. It is important to note that this would entail no threat to the RHR system integrity, the mechanical seal package would remain INTACT and no leakage would be seen outside containment. If the RCS to RHR in-leakage was greater than 1800 GPM, the indications the operator would receive would be somewhat dif ferent. Based on the worst case RHR Interf ace LOCA sequence run on the MAAP code by Westinghouse, ! the following alarus/ indications would be received by the operator. (please l refer to plot of RCS Pressure vs Time for basis of alara sequence) These alarms / indications are listed in CHRONOLOGICAL order and DO NOT take credit ! for any radiation monitoring systems alaras, although there are both sres and ventilation process monitors that monitor the RHR equipment vault area; RHR DISCHARGE PRESSURE HIGH alarm train ' A' ('B') (560 PSIG)

                                     -            RHR dischstge pressure meters pegged high (0-700 PSIG)
                                     -           RHR heat exchange outlet temperature recorders pegged high (0-500 Degrees F)
                                    -             PRT Pressure High alarm (4 PSIG)
                                     -            PRT Pressure and level increasing with no indication that the

, pressuriser PORV(s) or Safety Velves are discharging. l Pressurizer PORV(s) and Safeties have the following devices to monitor their position and process flow thru them;

                                                 -       Valve position for PORV(s) l                                                 -       Tailpipe temperature for PORV(s) and Safety Valves
                                                  -      Acoustic flow monitoring for PORV(s) and Safeties PRESSURIZER PRESSURE LOW alarm (2210 PSIG)
                                     -            PRESSURIZER LEVEL LOW alarm (-5 deviation from program)

RRR EQUIPMENT VAULT HIGH TEMPERATURE alarm Train ' A' ('B') RHR EQUIPMENT VAULT HIGH SUMP LEVEL alara Train ' A' ('B') l

RAI $7 1 The statement on page 3 - !! that "As soon as the pumps begin to produce flow to the RCS, valves in the miniflow lines close and all RHR pump 4 flow is injected into the reactor vessel via the RHR cold leg injection lines" is not correct. The sensors are not located at the RCS to detect flow at that location. Further, one is postulating a break in the RHR system, and a significant portion of the pump flow may never reach the RCS (as it stated in a later paragraph). RESPONSE 57 The referenced paragraph states that "As soon as the RHR pumps begin to produce flow to the RCS, valves in the miniflow lines close and all RHR flow is injected into the vessel via the RHR cold leg injection lines". This paragraph describes the response of the Emergency Core Cooling Systems to a standard LOCA, and is correct. The RHR flow transmitters are not located on the RCS proper, but are indicative of RHR injection flow via the cold leg injection lines. l

RAI 59

An ites under consideration for advanced nuclear power plants is the ability to monitor pressure on the low pressure side of check valves. This could provide early warning of check valve leaks and would pro-vide monitoring capability to help assure check valves were operating properly. The same monitoring capability with respect to RHR suction line valves could identify if individual valves were af spositioned or malfunctioning. Would such a system for Seabrook be of significant benefit in reducing risk in a reduced size emergency planning zone? RESPONSE 59 The f ailure model and quantification for leaks greater than 150 gpa is summarized below for the four cold leg injection paths (two check valves - in series) and the two hot leg suction paths (two MOVs in series). 4 COLD LEC INJECTION PATHS CHECK VALVE NO. OF FAILURE CONTINUOUS PATHS MODEL QUANTIFICATION (ANNUAL TESTING) MONITORING 4 2 ( T/2) 4[(4X10-4)2+ Variances = 3.5 x 10-6 o 4 2 d 8X(4X10-4 )X(2.7X10-4 ) 0.9 X 10-6

                                                                                                                                    =

0.9 X 10-6 2 HOT LEC SUCTION PATHS NO. OF FAILURE i PATHS MODEL QUANTIFICATION (18 MONTH TESTING) ' 2 2 ( T/2) 2[(4X10-4)2+Varianceux1.5 = 2.7 X 10-6 2.7 X 10-6 2 2 d 4X(4X10-4)X(2.7X10-4) = 0.4 X 10-6 0.4 X 10-6 2 g 2X(4X10-4)X(1.1X10-4) = 0.1 X 10-6 0 1 x 10-6 7.6 X 10-6 4.1 X 10-6 The last column above assumes perfect continuous testing (monitoring) of l the two series check valves. As shown the total f requency changes very little (less than factor of 2). However, quantification of the other failure modes and treatment of MOV discs as check valve discs are believed to be conservative.

                " Perfect continuous" leak testing of the RCS series check valves would require significant modification and is not practical.                                                                     Frequent, periodic testing can be performed to verify that excessive leakage is not occurring through the inboard RCS isolation valves. Excessive leakage through the i                outboard check will possibly be detected by increased pressure in the RHR t

system and reduced accumulator level depending on the magnitude of the leakage. Currently the normal RHR suction motor-operated valves cannot be tested as described above because permanent test lines are not connected to the process lines which the valves isolate.

I in addition, as discussed in the response to RAI 48b, the interf acing LOCA event contributes approximately 12% to the total early release f requency. Therefore, if this was reduced to zero it would be a minor reduction to total release frequency. There is no significant benefit to reducing risk. b o e 4 9 h I W a I l

RAI 60

Please elaborate on the page 3-23 list of actions an operator can take to sitigate the accident. This list appears to be short. Include iden-tification of what has been incorporated into operator training and proce-dures at Seabrook. RESPONSE 60 At present, the following isolation sequence is being incorporated into training:

1. Check proper valve alignment RC-V22 RHR pump "A" Th suction CLOSED RC-V23. RHR pump "A" Th suction CLOSED RC-V87 RHR pump "B" Th suction CLOSED RC-V88 RHR pump "B" Th suction CLOSED

2. Identify and isolate leak CLOSE THE FOLLOWING VALVES

          - RH-V21
          - RM-V22 OBSERVE THE RESPONSE OF THE RHR SYSTEM
          - Faulted train follows RCS pressure / temperature
          - Non-faulted train depressurizes CLOSE FAULTED TRAINS TO DISCHARGE VALVE
          - Energine NCC 522/622
          - Close RH-V14/RM-V26 STOP FAULTED TRAIN RHR AND CBS PUMPS CLOSE FAULTED TRAINS RWST SUCTION VALVE
          - CBS-V2
          - CBS-V5

3. Check if break is isolated:

RCS Pressure - increasing or Reactor Vessel level - increasing and Faulted RHR loep pressure - decreasing

4. If break is not isolated:

     - Energize MCC 522 and 622
     - Close RH-V14 and RH-V26 RETURN TO STEP 3 This strategy allows:

a. a quick reduction in leakage flow by isolating the faulted train from the non-faulted train

b. prevents cycling the RH injection valves if not necessary

c. allows better diagnosis by separation of trains.

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RAI 61

What is the fregnancy of failures in the pipe tunnel that is mentioned on page 3-23, and which led the authors to conclude they are very lovt RESPONSE 61 As noted on page 3-24, top event pI in the YI and is event trees. I represents any failure of piping or the heat exchanger doe to the RRR systen high pressure challenge. Any piping fai2eres are assigned to plant damage state 1 F7. As Table 4-17 shoes, damage state IFY always maps to release category $1- the seat severe release. In other words, any EER pipe f allars is soeumed to result in the most severs release, regardless of failure location; therefore results are insensitive to the pipe failure location. O 9

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RAI 62

Page 3-27 refereness situationa abere the combloed sump pump espacity is sufficient to remove leaks and keep the vaults free floodias. In t hes e cases, the REE SI, and CS pumps are sessmed not to be impacted by flood-i ng . What consideration was givso to is11are of one (or both) samp pumpet

RESPONSE

As shon la Figure 3-4 of PIA-0'432, the sequences referred to in this question beve frequencies on the order of IE-8 er lass; these fregosacies, when multiplied by the chance of samp pump failure clearly make such sequences afsportant contributors. l i

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RAI 63

What is the maximum flow rate that can be injected into the RCP pump seals? (Of potential interest since it may be an alternate path for injection into the RCS.) e

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RESPONSE 63 Flow to the reactor coolant pumps seals is significantly less than that required for core cooling.

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RAI 63

What is the maximum flow rate that can be injected into the RCP pump seals? (Of potential interest since it may be an alternate path for injection into the RCS.) RESPONSE 63 Normal seal injection flow into the reactor coolant system through the reactor coolant pump seal injection flow path is 5 GPM per reactor coolant pump or 20 GPM total. This flow path is judged insuf ficient by itself to satisfy core cooling requirements. Other flow paths such as the normal charging line, the charging / safety injection flow path to the cold legs or the safety injection pumps to the cold legs (at reduced

• RCS pressures) will provide greater flow capability. It should be noted that there are no risk significant core damage sequences in the ,

PSA results in which the charging pumps are available.

RAI 64

Shutting an RHR system crosstie valve is identified on page 3-35 as an action to help isolate a LOCA outside containment involving the RRR/SI systems. Has a careful evaluation of these systelus been performed to assess isolation strategy? If so, are procedures in place at Seabrook Station which reflect the work? RESPONSE 64 See Question 60. e

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RAI 67

Page 3-37 contains the wording "End state DLOC contains sequences in which the interf acing LOCA has been terminated, and the ECCS has been degraded (D) (RHR or SI pumps have f ailed)...The point estimate frequency of DLOC is 4.0 x 10-7 per year. The additional failures required to achieve core seit would lower this frequency by at least one order of magnitude." What is the justification for this conclusion? (We have already lost a portion or all of the' ability to inject water into the RCS via the usual paths). RESPONSE 67 End state DLOC represents sequences in which the interf acing LOCA has been terminated (isolated) and ECCS has been degraded but some core make up capability remains available. (see top of page 3-27). "The additional failures required to achieve core seit would lower this frequency by at least one order of magnitude" means that, given the degraded ECCS state, the unavailability of remaining mitigation equipment is better (lower) than 0.1.

l

RAI 68

The bottom of page 3-37 contains a statement to the effect that failure of one charging pump will lead to core melt. Why is this the case? Our perception is that sufficient flow might be provided by alternate means to keep the core covered, such as use of the remaining two charging pumps, and perhaps the tactor. makeup water pumps. < RESPONSE 68 For the DILOC sccident sequences discussed at the bottom of page 3-37, the f ailure of one charging pump would not be expected to lead to core melt. This is a conservative assumption to estimate the frequency of f ailures with the rest of the plant that would be needed to produce a core melt.

RAI 69. 70. 71. 72. 73

RAI 69

What i is to be the status of the " temporary" 34.5 kV power lines which are ' identified on page 3-457

RAI 70

What

'                                     is to be the status of the mobile power supplies which are identified on page 3-46?                                                                                     i

RAI 71

What capability has been provided to connect external pumps as identified in the second and third paragraphs of page 3-467 (This was briefly mentionsed on page 3-48). Use of a pump to simply inject water into containment via the sprays on a short ters basis (no recirculation) does not appear to be identified. Has this been considered?

RAI 72

l Page 3-46 identifies a number of possibilities for recovery of various saf ety f unctions. Are there specific plans? If so, please provide them.

RAI 73

i There have been several references to purchase of a mobile electric penerator by pooled resources on the pages prior to page 3-49. What is the likelihood that such a generator would be needed by several plants at the same time, and hence night not be available to Seabrook i Station when needed? Siellarly, where is the generator to be stored, i ' and how is it to be transported to Seabrook? Include consideration of post seismic and post severe stora in the response. } RESPONSE 69. 70. 71. 72. 73 These question refer to section 3.2 which discusses Containment Recovery following an extended loss of all AC power. The various potential recovery seasures discussed are related to recovery which could prevent t late containment failures predicted to occur one to several days after the initiating event. These late containment failures, and therefore, the i recovery measures discussed have no effect on early health risk. These l recovery measures are not related to emergency planning decisions. Reducing the chance of late containment failure could impact latent health effects _ which are not sensitive to evacuation assumptions. Alternate ways of i recovering late containment failures are still being evaluated. I I

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TABLE OF CONTENTS Page

1.0 INTRODUCTION

. . . . . . . . . . . . . . . . . . . . . . . . . .                                             1-1 1.1 Background                  . . . . . . . . . . . . . . . . . . . . . . . .                            1-1 1.2 Method                . . . . . . . . . . . . . . . . . . . . . . . . . .                          1-1 2.0 ACCIDENT SEQUENCES STUDIED . . . . . . . . . . . . . . . . . . .                                      2-1 2.1 Sequences Without Operator Actions . . . . . . . . . . . .                                       2-1 2.2 Sequence With Operator Action . . . . . . . . . . . . . . .                                      2-1 2.3 Uncertainty Analyses . . . . . . . . . . . . . . . . . . .                                       2-1 3.0 SEABROOK SPECIFIC INFORMATION . . . . . . . . . . . . . . . . .                                       3-1 3.1 MAAP 3. 0 Pa rame te r Fi l e . . . . . . . . . . . . . . . . . .                               3- 1 3.2 Operator Actions . . . . . . . . . . . . . . . . . . . . .                                      3-1 4.0 RESULTS            . . . . . . . . . . . . . . . . . . . . . . . . . . . .                            4-1 4.1 No Opera tor Ac tions . . . . . . . . . . . . . . . . . . . .                                    4-1 4.2 Operator Actions . . . . . . . . . . . . . . . . . . . . .                                       4-5 4.3 Influence of Uncertainties . . . . . . . . . . . . . . . .                                       4-13

5.0 CONCLUSION

S . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1

6.0 REFERENCES

. . . . . . . . . . . . . . . . . . . . . . . . . . .                                      6-1 1

APPENDIX A: MAAP 3.0 Seabrook Parameter File . . . . . . . . . . . . . A-1 APPENDIX B: Steam Generator Tube Integrity Analysis . . . . . . . . . B-1 B.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . B-1 B.2 Tube Degradation . . . . . . . . . . . . . . . . . . . . . . B-1 i

TABLE OF CONTENTS (Continued) Page B.2.1 Tube Properties

                                                                                  . . . . . . . . . . . . . . . . . . . B-2 B.2.2 Thinning / Cracking Type Defects . . . . . . . . . . . . B-2 B.2.3 Tube Denting . . . . . . . . . . . . . . . . . . . . . B-5 B.3 Creep Ru ptu re . . . . . . . . . . . . . . . . . . . . . . . . B-9 B.4 S u ma ry . . . . . . . . . . . . . . . . . . . . . . . . . . . B - 12 B.5 References                                                                                                                                 '
                                                       . . . . . . . . . . . . . . . . . . . . . . . . . B-14 APPENDIX C: Estimation of Steam Generator Tube Wall Temperatures . . . C-1
)

mm LIST OF FIGURES Figure No. Page 1 -1 Hot leg and steam generator natural circulation flow . . 1-2 4-1 Base case primary system pressure . . . . . . . . . . . . 4-3 4-2 Base case primary system structure temperatures . . . . . 4-4 4-3 Base case and seal LOCA case primary system Cs! re-tention . . . . . . . . . . . . . . . . . . . . . . . . 4-6 4-4 Base case and PORV case primary system pressure com-parison

                     . . . . . . . . . . . . . . . . . . . . . . . .          4-8 4-5       Base case and PORV case steam generator inlet gas temperature comparison . . . . . . . . . . plenum
                                                               . . . . . . 4-9 4-6        Base case and PORV case steam generator average tube tempera ture compa rison . . . . . . . . . . . . . . . . .        4-10 4-7        Base case and PORV case in-vessel hydrogen production compa ri son . . . . . . . . . . . . . . . . . . . . . . .        4-11 4-8        Base case and PORY case primary system Cs! retention . .          4-12 4-9        Base case and high eutectic temperature case hottest core node temperature comparison . . . . . . . . . . . .          4-14 4-10       Base case and high eutectic temperature case steam generator inlet plenum gas temperature comparison        . . . . 4-15 4-11       Base case and natural circulation uncertainty cases steam generator inlet comparison . . . . . . plenum gas temperature
                                     . . . . . . . . . . . . . . . . . . 4-17 4-12       Base case and natural circulation uncertainty cases steam generator tube temperature comparisons . . . . . .          4-18 4-13       Base case and steam generator PORY case primary sys-tem pressure comparison . . . . . . . . . . . . . . . . . 4-21 4-14       Base case and steam generator PORV case water level comparison . . . . . . . . . pressurizer
                                                       . . . . . .  . . . 4-22 4-15       Base case and steam generator PORV case steam gen-erator inlet plenum gas temperature comparison . . . . .          4-23 4-16       No blockage case core-uoper plenum flow      . . . . . . . . . 4-24

                                                                        - iv -

LIST OF FIGURES Figure No. Page 4-17 No blockage case upper-plenum gas temperature . . . . . 4-25 4-18 No blockage case steam generator plenum gas temperature . . . . . . . . . . . . . . . . . . . . . . 4-26 B-1 High temperature tensile properties of annealed (1600'F/1 hr.) hot-rolled plate (B-1) . . . . . . . . . . B-3 B-2a Burst data for 0.875 x 0.050 uniform thinnin mens - defect length variation (B-1) . . . .g .speci- . . . . . . B-4 . B-2b Burst data for 0.875 x 0.050 uniform thinning speci-mens - defect depth variation (B-1) . . . . . . . . . . . B-4 B-3a Burst data for 0.875 x 0.050 EDM slot specimens - defect depth variation (B-1) . . . . . . . . . . . . . . . B-6 B-3b Burst data for 0.875 x 0.050 EDM slot specimens - defect length variation (B-1) . . . . . . . . . . . . . . B-6 B-4 Burst pressure data of 0.875 x 0.050 uniform thinning specimens with and without denting (B-1) . . . . . . . . . B-7 B-5 Creep and creep-rupture comparisons (B-4) . . . . . . . . B-10 B-6 Master creep rupture curve for 316 stainless steel. taken from Ref. (B-7) ..................B-13 C-1 Natural circulation flows on the inside and outside of a tube carrying fluid from the inlet to the out-l e t p l e num . . . . . . . . . . . . . . . . . . . . . . . . C- 2 I l l \

                                                                                                                                   ~~   ~ '
                                                                                    - ------~ ---- -- - ~                     -

                                                                          -v-LIST OF TABLES Table No.

Page l I 4-1 Blackout Base Case Ke Seal LOCA . . . . .y Event Times, Without and With

                                                                         .................. 4-2 4-2               Operator Action Figures of Merit and Event Times                              ....      4-7 4-3               Key Event Time Comparison f Open Case . . . . . . . . .or Steam Generator PORV
                                                                               ............... 41g B-1 Comparison of Burst Pressures of Elliptically Wasted
                     .875 00 x .050 Wall Tubi (B-1) . . . . . . . . . ng With and Without Denting
                                                                           .................                 B-8 .

B-2 Creep Rupture Data for Inconel-600, Hot Rolled, and Annealed at 1600*F

                                                                      ................... B-ll O

t

v1 - 1 l 1 I i i

l-1

1.0 INTRODUCTION

l

1.1 Background

l A frequently-studied postulated severe LWR accident is the station blackout sequence (TMLB'). In this sequence, all off-site and on-site AC power are assumed to be lost. When analyses of this sequence are performed, high reactor vessel upper plenum temperatures are computed (1). This is caused by two factors, the heating of the upper plenum due to core-upper plenum natural circulation, and that caused by volatile fission products (chiefly iodine isotopes) which are released from the core and deposit in - the upper plenum. Experiments performed at Westinghouse (2_) in a one-seventh scale test facility have shown that these high upper plenum temperatures will cause natural circulation to initiate between the upper plenum and steam genera-tors. This process, which is shown schematically in Figure 1-1, would result in an increase in steam generator inlet plenum temperatures. Conse-quently, concern has been expressed that temperature-induced failures in the tubes or in the tube-to-tubesheet welds could result in a discharge of fission products to the secondary sides of the steam generators. Continued blowdown of the primary system to the secondary system through such failures could lift the steam generator safety valves, bypassing the containment. The principle purpose of this study was to assess the likelihood of these failures in the Seabrook plant, given a station blackout sequence. 1.2 Method Several variations on station blackout sequences were simulated using the Modular Accident Analysis Program (MAAP) version 3.0. This code is an integrated thermal-hydraulic and fission-product analysis code for severe accidents developed in the IDCOR program. Earlier version have been used by utilities, NSSS vendors, and consultants for approximately 4 years. The code has been verified and has been extensively benchmarked against experi-mental data, actual plant transients, and detailed code calculations (3).

1-2 Waa " ' [ (TOTAL

                          'OUT"              TUBE                   FLOW)[= <          1 "BACK' TUBE                                                      :
             ',                                  SEC                     =-

r Wst f T" eo TUP ' W TH Tc gt i I w% i Figure 1-1 Hot leg and steam generator natural circulation flow.

1-3 MAAP is particularly suited for this study since it contains fully integrated models for:

a. Core overheating, oxidation and melting.

b. Natural circulation between the core and upper plenum, between the upper plenum and inlet steam generator plena, and between the inlet and outlet plena of the steam generators. Calculations using the natural circulation models have compared well to the Westinghouse one-seventh scale tests (1,2_).

c. Fission oroduct release, transport, deposition, and revapori- -

zation. In addition, MAAP allows arbitrary operator actions to be applied so that the efficacy of the likely operator responses can be studied.

2-1 2.0 ACCIDENT SEOUENCES STUDIED In' all the sequences presented below, it is assumed that off-site and on-site AC power remains lost indefinitely. In such a case, the emergency response procedures (ERPs) (4) instruct the-operators to initiate auxiliary < feedwater (AFW) using the turbine-driven pump. As long as sufficient AFW' flow can be maintained, decay heat would be removed and the accident would not progress further. For the cases studied here, it has been further assumed that all AFW has been lost. This reflects a major conservatism in-the analysis. 2.1 Seouences Without Operator Actions 4 Two sequences were studied in which no operator actions were credited. In the base case, no additional failures were assumed. In the second case, loss of cooling to the main coolant pump seals was assumed to result in a leak area corresponding to a flowrate of 50 gal / min of water per pump... - This leak area is the same as that assumed in the IDCOR program (5_). 2.2 Sequence' With Operator Action The Seabrook ERPs.(5_) specify that when AC power is available and all j ~ other measures have been attempted, if core thermocouple temperatures exceed i 1200*F, the pri try system should be depressurized using the pressurizer PORVs. This enables the accumulators to be used to inject coolant into the core. While this action is not invoked for blackout sequences, it was considered that this action might well be recomended by the Technical Support Center. Accordingly, this action was simulated in one care. 2.3 Uncertainty Analyses Several variations on the base case were studied to investigate the impact of uncertainties in phenomenological parameters and accident pro-gression. In these cases, parameters were varied in such a way as to increase the potential for high steam generator temperatures: I l

                                                                        .    .    .  . - ,   i

2-2

a. A high (3000'K) core melting temperature was assumed compared to the nominal value (2500*K). This delays the onset of core geome-try degradation and thus enhances the potential for the core to heat the rest of the primary system.

b. Low values of axial and cross-flow friction factors were assumed in the core modelling. This also tends to maximize convection of heat out of the core.

c. Lower values of steam generator natural circulaticn flow W3g (see Figure 1-1) relative to hot leg natural circulation flow WHL " '

assumed to minimize cooling of the steam generator inlet plenum by flows from the outlet plenum. This was accomplished by choosing lower limit values of the number of steam generator tubes partic-ipating in the flow from the inlet to the outlet plena, guided by observations in the Westinghouse experiments (2). j

d. A run was made where it was assumed that the steam generator PORVs stick open. Depressurizing the secondary side tends to decouple the tubes from the shell after the steam generator dries out.

This could potentially increase tube temperatures.

e. A run was made in which it was not assumed that coolant channel blockage occurs in a core node when melting begins in that node.

MAAP ordinarily makes such an assumption to represent the rapid reduction in flow area that would occur after melting commences. In this run, complete blockage of the flow area in a node was only, assumed to occur if the node completely filled with molten ma- , terial. By allowing continued core oxidation and core-upper plenum flow, this assumption maximizes the potential for the core to heat the rest of the primary system.

3-1 3.0 SEABROOK SPECIFIC INFORMATION 3.1 MAAP 3.0 Parameter File Since MAAP 3.0 was used for this analysis, it was necessary to develop a Seabrook-specific parameter file. The major part of this effort was completed by using the MAAP 2.0 Seabrook parameter file which had been developed in support of the Emergency Planning Zone study (1). Additional changes were made to the Seabrook MAAP 2.0 deck to reflect the model revisions in MAAP 3.0. The major additions and changes were in, the areas of reactor vessel heat sinks, peaking factors, fuel rod ballooning data, model parameters, and auxiliary building data. The new MAAP 3.0 data for this analysis was derived from Seabrook plant drawings, data deleted or modified from MAAP 2.0, the Westinghouse IMP data base, and other generic and Seabrook specific documents. A copy of the ~ Seabrook MAAP 3.0 parameter file is included in the Appendix. 3.2 Operator Actions In order to get a complete picture of steam generator tube response ' during a station blackout transient, operator actions were also considered. The potential operator actions were derived from the Seabrook ERPs (4.j). l In the event of a station blackout, the operators are instructed to initiate auxiliary feedwater (AFW) operation using the turbine-driven pump. This action was not modelled in the MAAP 3.0 runs, since it would result in the removal of decay heat and would effectively prevent core uncovering from occurring. In sequences with AC power available, the symptom-oriented procedures (!,) call for depressurizing the primary system using the PORVs if core ~ temperatures exceed 1200*F and all other means for cooling the core have L _ __ _ _ _ - . - ---

3-2 been attempted. While this action is not invoked in the procedures if AC power is not available, it was considered that such an action might well be recomended by the Technical Support Center in a blackout in order to obtain flow from the accumulators. For this reason, this action was simulated in one of the comparison cases. f l l l _ _ _ _ _ _ _ , _ , _ , _ _ - - - - - - - - -_ - - - - - ' - - - - " - " - ~ - ~ ' ' ' ' ' - - ' ~ ~ ~ ~ ~ ' '

4-1 4.0 RESULTS 4.1 No Operator Actions Two cases were run with no operator action, the base case and the base case with a pump seal LOCA at 45 minutes after initiation. Since the results of these sequences are quite similar, the base case will be de-scribed here and only differences will be noted. Major events are listed in Table 4-1. After the blackout is initiated, the steam generator inventory begins to be depleted. As the water level in the steam generators de-creases, heatup and expansion of the primary coolant leads to the pressur - izer " going solid", i.e., completely full of water. The quench tank rupture disk then breaks after discharge from the pressurizer overpressurizes the

tank. The steam generators dry out completely soon afterwards.

The seal LOCA case has slightly different timing until steam generator dryout because of the loss of primary system inventory through the failed seals. Thus, in this case, the pressurizer goes solid later because there 4 is less water which can expand. Steam generator dryout, on the other hand, is determined simply by integrated decay power and the available secondary side inventory and differs little from the base case. Loss of primary coolant leads to eventual core uncovery, meltdown, and vessel failure. The primary system remains at high pressure until vessel failure as dictated by the pressurizer relief valve setpoints (Figure 4-1). Strong natural chtulation occurs between the core and upper plenum after the core uncovers; tW in turn sets up circulation between the upper plenum and the steam generators. Fission products leave the core during the heatup and can migrate through the upper plenum to the hot leg. The circulation and fission product transport are affected by temperature differences, and j feed back to influence region temperatures (Figure 4-2). After vessel failure, the primary system blows down, the pump bowls clear, and the accumulators empty. Most core debris and water in the lower plenum at that time are entrained into the lower coinpartment, which is

 -,r- -

r..-- - , - , . - . . ~,_ --_m , _ , . - , -

4-2 Table 4-1 BLACK 0UT BASE CASE KEY EVENT TIMES. WITHOUT AND WITH SEAL LOCA Event Base Case Seal LOCA Initiation 0.0 0.0 Seal LOCA - 2714. Pressurizer Solid 4829. 5920. Quench Tank Disk Rupture 5486. 5944. Steam Generator Dryout 5527. 5517. Steam in Pressurizer 6319. 6269. Core Uncovery 7280. 7124. Vessel Failure 11648. 11054. Accumulator Depleted 11700. 11102. End of Simulation 20000. 20000. I 4 i

i l 4 1 l i 1 1 4 SEABROOK TM_B BASE i e e i ON 2'''I''''I''''l''''l''''l''''l''''l''''l''''l'''' t X  : _~ i i ' C - . A - _

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l l l I STRUC. TEMPERATLRE C K)  ; l 500 600 700- 800 300 1(Xe i100 1200 0...1..

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4-5 connected to the reactor cavity. Airborne fission products are swept from the primary system at that time. With the pump bowls clear, natural circu-lation can occur throughout the primary system, and some long-term revapor-ization of fission products is possible. In the case of the seal LOCA, slightly more Cs! leaves the primary system due to leakage through the pump seals and the lower system pressure (Figure 4-3). 4.2 Operator Actions Operator action to depressurize the primary system by opening the pressurizer relief valves was assumed to occur when the maximum core temper-ature reached 1200*F, as discussed earlier. This resulted in complete . blowdown of the primary system before bottom head failure. Thus, the time during which steam generator tubing was exposed to both high pressures and high temperatures was significantly reduced by operator action. Key e'.'ents and figures of merit are sumarized in Table 4-2 and com-pared to the base case. It can be seen that not only is the peak steam generator inlet plenum gas temperature lower after the operator action, but the primary system pressure is lower as well. The action occurs at 8000 seconds, about 15 minutes after core uncovery, and leads to accumulator dis-charge well before bottom head failure. This discharge occurs over a period of several thousand seconds, as seen by the gradual decrease in primary systempressure(Figure 4-4). The steam generator inlet plenum gas tempera-ture (Figure 4-5) is high only after depressurization, and the tubes them-selves are relatively cool (Figure 4-6). While, as shown in Figure 4-6, tube temperatures continue to increase after vessel failure, simulations continued for a longer time than those presented here show that the tempera-tures do not increase much beyond 700*K due to heat losses to containment. Since the differential pressures are much lower after vessel failure, these moderate temperatures are not limiting. l The accumulator discharge causes only slightly more hydrogen production (Figure 4-7) through availability of steam. However, the large flows caused by steaming of accumulator water flush fission products from the primary systemintothecontainment(Figure 4-8).

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4-7 Table 4-2 OPERATOR ACTION FIGURES OF MERIT AND EVENT TIMES Base PORY Seal Case 9 1200'F LOCA

1. Peak SG inlet plenum gas tem- 858. 682. 865.

perature while primary system is at high pressure ( K)

2. Primary system oressure at 17.1 1.83 14.9 bottom head failure (MPa) .

3. Peak SG inlet plenum gas 858. 843. 865, temperature (*K)

4. Primary system pressure at 17.1 4.23 16.9 l

time of #3 above (MPa)

5. Time of peak T #3 (sec) 9230, 9290. 9000.

6. PORV open (sec) --

8005. --

7. Timeaccumulatorsdepleted(sec) 11701, 10643. I 11102.

8. Vessel failure (sec) 11648, 12601. 11054.

9. Approximate peak SG tube wall 750 low 760 temperature (see Appendix C)

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• figure 4-8 Base case and PORV case primary system Csl retention.

l

4-13 4.3 Influence of Uncertainties Five sensitivity cases are presented here which show the influence of high core melting temperatures, enhanced core to upper plenum natural circulation, reduced steam generator tube circulation, blowdown of the steam generator secondary sides, and neglect of core blockage phenomena. Each of those factors individually can increase hot leg and steam generator plenum temperatures relative to nominal cases. I A high UO2 -Zr eutectic temperature of 3000*X (versus the nominal . 2500*K) was used for uncertainty analysis of the base case presented above. In the base case, vessel failure occurred at about 11,600 seconds (Table 4-2), while in this case vessel failure was delayed until 12,300 seconds. ' This is because more time is required to reach the higher eutectic tempera-ture and cause melting. Other thermal-hydraulic behavior and event ti dng is similar to the base case, with exceptions relating to higher core temper-atures. The driving potential for natural convection can be compared between this and the base case by considering the hottest core node tempera-ture (Figure 4-9), which shows the effect of the input eutectic temperature. Only a small impact on the steam generator inlet plenum gas temperature is caused by this parameter change (Figure 4-10), with a slight delay in l reaching the peak values. Thus, considering an uncertainty of 500'X in the core melting temperature, a difference of only tens of degrees in steam generator temperatures results. This is because the heatup rates in the core are very much more rapid than the heatup of the steam generator plena j once temperatures are high enough to cause rapid Zircaloy oxidation. Natural circulation between the core and upper plenum was enhanced by lowering the axial and cross-flow core friction factors in one sequence. Circulation between the steam generator tubes and the inlet plenum was reduced by lowering the fraction of tubes carrying flow out of the plenum in another case. Each case acts to increase the steam generator inlet plenum i temperature over the base case value. i There is a small effect of these changes on sequence timing: vessel failure occurs at 11600 seconds in the i base case, 12100 seconds in the high core circulation case, and 11200 seconds in the low steam generator circulation case. In the high core _ _ _ . _ . . ~ _ . _ . . _ _

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

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SEABROOK (NCERTAINTY COMPARISfN EanSE AIII TFD ' ] t Q os i

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            @0                 2          4          6           8             10       12         14       16         18       20 a

TIME S x10' - i l Figure 4-10 Base case and high eutectic temperature case steam generator i l inlet plenum gas temperature comparison. I t 4 s

  - - .                 -    . _ _ - _-   -    .-. = _ _ _ _ _ - .           -. -   . - _ . -        - - _ .

4-16 circulation case, more heat transfer occurs to the steam generators and i vessel failure is slightly delayed. In the low steam generator circulation case, less heat transfer from the plenum to the tubes occurs, and vessel ' failure is slightly hastened. The steam generator inlet plenum gas tempera-ture (Figure 4-11) is significantly higher (about 150'K) only for the steam generator circulation case ("$G NC"), while it is only slightly higher for  ; the core circulation case (" CORE NC"). Tube temperatures (Figure 4-12) l follow the same trend. It should be noted that the low steam generator circulation case, which assumed that only 10 percent of the tubes carried flow in the "out" direction, is in all likelihood unrealistically severe. I More typical extremal values of the flow split observed in the Westinghouse ' experiments (about 20 percent of the tubes carrying out flow) result in only about a 900'K peak inlet plenum gas temperature. A blackout case with stuck open steam generator relief valves on all the units was run to minimize the heat transfer capability of the tubes, and

                                                                                                             '{

thus increase the steam generator' inlet plenum temperature. In the base  ! case, steam on the secondary side serves as an efficient heat sink for the small recirculation flow through the primary side, and also couples the tube mass with the steam generator shell heat sink. When the secondary side is blown down, and this heat capacitance is impaired, the tube outlet tempera-ture is higher and thus the mixed mean steam generator inlet plenum tempera-ture is higher. t The early behavior of this sequence differs dramatically from that of + j the base case due to enhanced cooling by the secondary side early in the , transient. Comparing events (Table 4-3), the steam generators go dry much , earlier in the steam generator PORV case, and the primary system cools down i enough that the pressurizer is drained due to contraction of the coolant. ' ) Later, of course, with the heat sink lost, primary system fluid reexpands I l and the pressurizer goes solid. Core uncovery and vessel failure occur

!       slightly earlier in the steam generator PORV case because the overall                                  "

integrated heat removal is lower with the steam generators blown down. - i i i

i l

 )

{ a SFABROOK 1NCFRTAINTY COWARISON .

                                                                       .4 CORE NC                                ySC NC                                                                                                    '

a:;cNSE

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O- 0 20 0.40 0 60 0-80 12 14 1_6 1

i 1 1-8 2_O 4 i i TIME S x10 j Figure 4-11 Base case and natural circulation uncertainty cases steam i generator inlet plenum gas temperature comparison. I, - l i i

i i f .i 4 j SEABROOK IJNCFRTAINTY COtFARTSON i o aON ..o ig..n ni g A N MC

                                                                   ........g .. u.

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Table 4-3 KEY EVENT TIME COMPARISON FOR STEAM GENERATOR PORV OPEN CASE Event 3 Rv Base Case 0p Initiation 0.0 0.0 SG PORY Open -- 0.0 SG Dryout 5527. 2738. Pressurizer Drained -- 350. Pressurizer Refilling -- 2000. Pressurizer Solid 4829. 4205. Quench Tank Disk Rupture 5486. 4937. Pressurizer Has Steam 6319. 5920. Pressurizer Empty 7000. 7400. Core Uncovery 7280. 6626. Vessel Failure 11648. 11045.

4-4-20 Primary system pressure (Figure 4-13) shows the cooling achieved while the secondary side blows down, resulting in a 10 MPa transient over 4000 seconds. The pressurizer water level (Figure 4-14) illustrates the con-traction of primary system coolant during the cooldown, and later reexpan-sion. Since the core uncovers earlier in the steam generator PORV open case, high gas temperatures in the steam generator inlet plenum are shifted to slightly earlier times (Figure 4-15). The initial high peak, prior to degradation of core geometry, is the same in each case both in magnitude and rise behavior. Thereafter and before vessel failure, the inlet plenum gas temperature is about 50*K higher for the steam generator PORV case. How- . ever, it is still below 800*K for most of the high temperature period. The last uncertainty case assumed that nodal core flow channel blockage does not occur at the onset of melting in the node. MAAP normally assumes such blockages to represent the rapid reduction in flow area that would occur after melting begins. In this case, complete blockage was credited only when the geometry would allow no flow, i.e., when the node was com-plately full of refrozen eutectic. As shown in Figure 4-16, a rapid reduc-tion in core-upper plenum flow still occurs after the beginning of accel-ersted oxidation. While the flow continues at relatively low values, heat

removal from the upper plenum due to the hot leg flows about equals heat convected from the core and that due to fission product heating, resulting in a stabilization in upper plenum gas temperature (Figure 4-17). Steam 1

generator inlet plenum gas temperatures (Figure 4-18) also stabilize. The j peak sustained plenum gas temperature, which occurs just before vessel I failure (12,000 secs) is about 1060'K. Thus, even when blockage in the core is essentially neglected, only relatively moderate increases in plenum gas

temperature are seen over the base case.

e i i 3 4

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t 1 i i BASE SEABROOK THLB COHPARISON l

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-{                                                                                                                                                          2 TIME S                                                   xlO k             figure 4-13

) i Base case and steam generator PORV case primary system pressure comparison. W 4

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g n O. 2 4 6 8 10 12 14 IG 18 20 3 TINE S x10 figure 4-15 Base case and steam generator PORV case steam generator i inlet plenum gas temperature comparison. i V

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5-1

5.0 CONCLUSION

S l 1 l Various blackout sequences have been analyzed for Seabrook which cover a range of potential operator actions and phenomenological uncertainties. An important result of the base case is that for the period when the primary system is at high pressure, the peak steam generator inlet plenum gas temperature is only s 850*K, and average tube temperatures are well below 700'K. Operator action to depressurize the system can successfully reduce - the system pressure prior to the time high temperatures are reached in the steam generator. When uncertainties in circulaticn between the core and upper plenum or in the core melting temperature is considered, negligible increases in the steam generator temperatures are observed. Similarly, only slight increases in temperature were obtained in a case where the secondary sides of the steam generators were depressurized. For an unmitigated case, considering a large variation in the steam generator inter-plenum circula-tion indicated that gas temperatures can briefly reach about 1000*K in the steam generator plenum. Similar temperatures (s 1060*K) are achieved when the blockage of core node coolant channels due to melting is neglected. Both of these latter cases are considered unrealistically extreme. MAAP does not contain a detailed model for the change in tube wall tamperature as one leaves the inlet plenum and moves toward the outlet plenum of the steam generator (i.e., only average wall temperatures are , l computed). However, by using the average secondary side gas temperature and the temperature of the primary side gas entering the tubes, the wall temper-atures can be estimated by knowing the value of the heat transfer coeffi-cients on the primary and secondary sides. As shown in Appendix C, these heat transfer coefficients are approximately equal. By equating the heat flux convected to the tubes by the primary side gas to that convected away from the tubes by the secondary side gas, the analysis in Appendix C leads to the conclusion that the peak tube wall temperature will be the average of the inlet gas temperature and the average secondary side temperature. In the worst case discussed above, this results in a peak tube wall temperature of approximately 850*K and is only 750'K in the best estimate case. As

5-2 shown in Appendix B, these are considerably less than the temperature values which would challenge the integrity of the tubes. Therefore, steam gener-ator tube rupture is judged to be very improbable for the sequences ex-amined. e l t

6-1

6.0 REFERENCES

(1) Fauske & Associates Inc., " Technical Support for Issue Resolution", s IDCOR Report 85-2 (July,1985). (2_) W. A. Stewart, et al., " Experiments on Natural Circulation Flows in Steam Generators During Severe Accidents", Proc. Inter. ANS/ ENS Topical Meeting on Thennal Reactor Safety, San Diego, California (1986). (3_) R. E. Henry, " Benchmarking of Severe Accident Codes: How Should It be Done and How Should it be Used?", Proc. Inter. ANS/ ENC Topical Meeting on Thermal Reactor Safety, San Diego, California (1986). . (4_) New Hampshire Yankee Station Emergency Operating Procedure ECA-0.0,

     " Loss of All AC Power", (May 16, 1986).

(5) Comonwealth Edison Company. " Zion Station Integrated Containment Analysis", IDCOR Report 23.12 (1985). (6_) New Hampshire Yankee Station Emergency Operating Procedure FR-C.1,

     " Response to Inadequate Core Cooling", (May 16, 1986).

(,7,) Pickard, Lowe, and Garrick, Inc. , "Seabrook Station Risk Management and Emergency Planning Study", PLG-0432 (December,1985).

A-1 APPENDIX A MAAP 3.0 Seabrook Parameter File m G e

                                   ?

m a w-

A-2 jag g g g g g PLANT: TENTATIVE PARAntTER FILE FON MAAP 3.0 as s AW EfW s s sassssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssas as i as6ENERAL INFORMATION: ! ss

          !!1.FORTHEGE0METRICALRELATIONSHIPM1WEENVARIOUSCONTAINM as      SYSIEM N0 DES: SEE THE REDIUM SUBROUTINE WRITE-UPS IN VUL 2 0F USER'S MANUAL as      FOR LARGEe DRY CONTAINMENTS, SPECIFY ZERO VOLUME FOR ICE-CONENSER AND ss
          **      UPPER PLENUM-fMIS CAUSES ALL OTHtR UPPER PLENUM AND ICE CONDENStR PARAMETERS TO M IGNORED sa 882.

88 NOTE OUTER WALLS IN COMPARTMENIS A AND D SEPARATE THE CONTAINMEN1 FROM

'         as      THE ENVIRUNNENTI THE UUIER WALL IN COMPI 3 SEPARATES CUMPTS 3 AND Di

' ** THE OUTER WALLS IN COMPTS I AND U (ICE CON M NSENS ONLY) ANE NOT MUDELED as SINCE THESE WALLS ARE INSULATEDI THE UUIER WALL IN COMPT C IS ASSUMED

          ** TO  THEM     ADIASATIC INNkR FACE) ON ITS FAR SIDE (IE NU HEAT LOST FNOM THE SIDE OPPOSITE   '
          **     TO REPRESENT A FREE STANDING STEEL CONTAINMENT WITH A SHIELD DUILDING, ss      TREAT THE SHIELD BUILDING AS THE WALi AND THE CUNINT PRESSURE BOUNDARY ss      AS A ' LINER *i ENTER Th1E 'GY' DICTANCE KIKEN THL TWU WHkRE CALLED FOR

483.

as INTERNAL OR INTERIOR WALLS ARE WALLS TOTALLY CONIAINED IN A COMPT ss PROPERTIES (TNERMAL CONDUC ETC.) 0F IN1ERIOR WALLS IN A AND 3 ARE ss e.SSUMED IME SAME AND ARE ENTENED IN THE LOWEk COMPT SECTION

         **      IF (AS IS USUALLY IME CASE) YOU MUST LUMP WALLS OJ SEVERAL THICKNESSES ss     TOGETHER. YOU SHOULD LUMP ONLY RELATIVELY THILK WALLS (EG SkEATER THAN
         **     ABOUT 1 FOOT OR .3 METER IN THICKNES$) AND EN1ER IME THICKNESS OF THE THINNEST WALL CREDITED sa ss4.

sa DECK REFEks TO THE FLOOR (AND VERTICAL WALLS IN ICE CONDENSER PLANIS) 88 fHAT SEPTRA 1ES THE UPPiet COMPARTMENT FROM INE COMPARTMENTS LOWER IN THE CONTAINMENT ss l **5. TWO WAYS TO HANDLE CONTAINMEN1 FAILUkEI

         $$   A. MECHANISITIC MUDEL:

as

         **        ENTER 0 FOR THE FAILUkE PRESSURE (ACOMPT NO. 34) ALSO SUFtLY:

as (1) CONCREIE: SUPPLY ALL THE MAIERIAL DATA: CUNCRElE PARAMS 13-22 ETC. 88 (2) FREE STANDING STEEL SHELL: ENTEN THE WALL THICKNESS IN THE UtFER ss AND ANNULAN COMPARIMENTS IN THE ' LINER' THICKNESS EN1 RIES ss AND SUPPLY ONLY TE LINER MATERIAL FRUPERTIESI THL NUMBER OF 8: TENDONSeAND AMUUNT UF REBAR SHOULD BE SLT TO 4ERO IN THIS CASE

3. SIMPLE MODEL:

ss ss SUPPLY ACOMPT NO. 34 AND 3/f FAILURE AREA ENTERED AS MUDEL PARAMAfER NO.23NEED NUT SUPPLY IHE OTHER PARAntTERS ss sse. as SEDIMENTATION AREA' IS THE TOTAL UPWARD-FACING AREA IN A ss GIVEN COMPARTMENT UPON WHICH FISSION PRODUCT AEROSULS CAN SETILEI THIS SHOULD INCLUDE (WHERE APPROPRIATE), FLOORSe CABLE TRAYbe EUUIPMENI ETC ss as7. as AS DESCRIMD IN THE SCON1ROL SECTION, THE AUXILIARY BUILDING MUDELS Akt ACTIVA 1ED sY SUPPLYING A NONIERO NU. OF AUX. NUpts TO BE MODELLED. 85 THE MODEL CAN BE RUN SIMULTANEOUS WITH A RUN OF THE CONIM) AND FkIMARY

A-3 as SYSTEM MON LSe QRe BY SUPPLYING A NUNZERO INPUT FILE NO.e THE

     **        AUX MODELS ONLY CAN K RUN USINU AN INPUT FILE OF T/H DATA FROM AN as       EARLIER MAAP RUN.

as ) i as8. FISSION PRODUCT REMOVAL SY INERTIAL IMPACTION IS MODELLED ONLY IN

     **        ONE CONTAINMENT COMPARTMENT. IN LARGEe DRY'S SUCH PARAME1ERS 1

i ss SHOULD CHARACTERIZE GRATES WHICH ARE ASSUMED TO K IN THE ANNULAR

     **       COMPARTMENT. IF MORE IHAN UNE LEVtl 0F GRA1ES EXISISe SUPPLY IHE TOTAL ss        IMPACTION AREA 0F ALL THE GRATES: AND THE MAXIMUM FLOW AREA AT ANY
     **       OF THE GRATE ELEVATIONS N        IN ICE CONDENSER PLANIS THESE PARAMETERS (EVkN THOUGH LOCATED as
     **        IN THE sANNULAR COMPARIMENT DATA SECTION) SHOULD REFLECT IMPACTION AND sa       FLOW AREAS AND STMAP WIDTHS IN THE ICE 80X--SEE EG FOS 1MA'S REP 0kr as?. THE UNITS FOR PARAMETER INPUIS ARE SPECIFIED BY EITHEN A *S1 (ETRIL) st OR 88R (BRITISH) UNITS CARD. ALL PARAMETERS FULLUWING SUCH A CARD sa      ARE ASSUMED TO HAVE THESE UNITS UNTIL THE NEXT UNITS CARD IS INLLUDED.

( THUS A PARAMETER FILE CAN HAVE SECTIONS WITH DIFPERENT UNITSe IF 38 DESIRED. THE LAST UNITS CARD IN 4 PARAMETER FILE CONTROLS Tit UNITS

     **      UF OTHER PROGRAM INPUTS IN TAPE $ (EG START AND FINAL f!MES ETC.) AND
     **      THE UNITS TO K OUTPUT IN THE OU1PUT FILE AND PLOT FILES.

as MkTRIC UNITS ARE M-KG-SEC-DEGREE KELVIN-PASCALS-Mas 3/SECeETC. as BRITISH UNITS ARE FEEL-LBM-HOURS-DEGREE F-PSI-GPM as EXAMPLES: sa IN METRIC UNIT $e FLOWRATES SPECIFIED TO BE VOLUMEIR1C SHOULD K ss Mas 3/SECB OfMER FLOWRATES IE ALL 1 HOSE NOT LXPLICIILY STATED TO K 3 VOLUMETRIC SHOULD K KG/SECl HEADS SHOULD BE IN MI PRESSURES IN pal as IN ENGLISH 1HE UNITS ARE RESPECTIVELY OPMeLSM/HRe>Te PSIA--

    **       NOTE TO MAAP/8WR USEkS--GPM IS USED IN MAAP/PWR INSTEAD OF FTss3/HR sa        fHE UNLY EXCEPTION TO THIS PROCtDURE IS THAT IHE TIME STEPS ARE ENTERED
 ,  as       IN THE 8 TIMING SECTION ALWAYS IN SECONDS i

as as10.IN

    ** SUCTION  LARGEe DRY CONTAINMENIS ' FANS' REFEkS TO FAN COOLERS WHILH TAKC FROM INE UPPER COMPT AND DISCHARGE TO EITHER THE LOWER OR
    ** ANNULAR          COMPT. AS SPECIFIED BELOW. THE SAE INPUIS IN ICE CONDEN5ER
    ** PLANTS AME          USED TO LHARACI'ERIZE IHE AIR RETURN FANS.

Es

    * * ***s338883388***83388 *s333333*s t s s *** *s s s * * * * ** s s *

• s s a t s s s s ss s s s s s s t a s 8FR st assatassassstatssasssstss**stssstass**sssssssssssssssss**stsss**stssssas SUPPER COMPARTMENT (OR 'A' COMPT) assass**ssassss&;sssssss-asasssssssssssssssassassssssssssss****ssssssss**

! 01 2.138D6 FREE VOLUME I 02 1415. AREA 0F REFUELING POOL 03 141. HEIUNT OF CONIAINMENT SPNAY HEAD ABUVE BOT 10M OF COMPARTMtNT 04 3437. FLOW AREA FROM UPPEN COMPARTMEN1 IN10 ANNULAR COMPT OS 15394 CHARACTERISTIC CROSS-SEC AREA 0F COMPT FOR BURN TIME 8: CALCS--EG THE SURN TIME IS THE SOUAkE ROOT OF THIS

   **                   AREA DIVIDED BY THE SURN VELOCIlY 06       0.0         CURS HEIGHT IN REFUELINO POOL TO ALLOW OVERFLOW--N0kMALLY ss                   0 UNLESS YOU ASSUME REFUELING POUL DRAIHS ARE BLOCKED (A

, as CLASSICAL ICE CONDENSER SEQUENCE), THEN MAKE If LARGE 07 72122. SURFACE AREA 0F UUTER WALLS IN UPPER COMPARIMENT 08 .0357 LINER THICKNESS ON OUTER WALL 09 0.28 UUTER WALL LINER GAP RESISTANCE--SEE NOIE IN

• LOWER COMPT l

   **                   FOR HOW TO MODEL FkEE STANDING STELL CONTMTS WITH A SHIELD l

l l l

A-4

      **                WALL 10    41         OUTER WALL TOTAL THICKNESS 11    0.92
      **                fMERMAL CONDUCTIVITY OF UUTER WALL (FOR CONCRL1E STRUCTURES WITH A LIN(Rs THIS REFERS TO THE CONLRE1E PANT) 12    0 157       SPECIFIC MEAT OF UUTER WALL 13    144        DENSITY OF OUTER WALL 14   0
      **               ENTER A 1 IF THE OUIEN WALL IS SOLID STEEL (IE A SIEEL CONTM1 15    11116. W11H NO SHIELD BUILDING)e 0 FOR CONCREIE WIIH OR W/0 LINER 16    0          HALF AREA (WALLS MODELED AS 1-P SLASS) 0F INIERNAL WALLS LINER 1HICKNESS ON INTERIOR WALLe IF ANY 17   0 LINER GAP RESISTANCE IN INTERIOR WALL 18    4.          fHICKNESS OF INIERNAL WALLS 19   13990.      DECK AREA 20    0           LINER THICKNESS ON DECK 21    0           LINER GAP RESISTANCE ON DECK 22    4.2         DECK THICKNESS                                                -

23 0.92 THERMAL CONDUCTIVITY OF DECK 24 0.157 SPECIFIC HEAT OF DECK 25 144. DENSITY OF DECK 26 0 27 3.434D4 ENTER A 1 IF THE DECK IS SOLID STEEL: 0 FOR CONCRElk METAL E0PT MASS 28 2.0A0305 EOPT HEAT TkANSFER AREA 29 0. NUMBER OF IGNITION SOURCES IN UPPER COMPT (A COMPARTMENT) 30 0. AVERAGE DISTANCE OF THESE FNOM THE CEILING OF A asFULLOWING PARAME1ERS ARE USED TO DtTERMINE WHICH IGNI1ERS OR IGNITION

     ** SOURCES IN THE LOWLRe ANNULARE OR UPPER PLENUM CAN IN111A1E DukNS IN stTHEIR RESPECTIVE COMPARTMENTS WHICH CAN THtN PROPAGATE INTO IHE UPPER
    ** COMPARTMENT--IF NO IGNITERS IGNORE 31-3J 31       0.        NO. OF IGN!)ERS/IGN SOURCES IN & WHICH CAN BE STEN FROM A 32      0.        NO. OF IGNITERS /IGN SOURCES IN D WHICH CAN DE SEEN FROM A 33       0.       DISTANCE FROM INE 10P OF A IO THE DECK 34       .90
    **                FRACTION OF UPPEN COMPT SPRAY WATER THAT RUNS IN10 1HL REFUELING POUL (VS. CUNTINUING ON DIRECTLY INTO LOWER COMPT) 35       .35      FRACTION OF WATER DRAINING 001 0F REFUELING PUOL 1HA1 g                 ggTU LOWER COMPT (REMAINING FRACTION RUNS INTO THE 4sINPUTS FOR SIMPLE (FAILURE PRESSURE SUPPLIED) OR DLTAILED (CON 1MI STkAINS
    **CALLULATED) 36      187. MODELS FUM CONTAIMtNT FAILURE--SEE GENERAL NOTES AB0VE 37      0         FAILURE PRESSURE OF CONTAINMENT OR 0 TO USE DETAILED MODEL i
    **                ENTER A 1 IF CONfMT FAILS IN UPPER COMPTIO FOR FAILURE IN THE ANNULAR COMPT (USED ONLY FUR THL SIMPLE MUDEL) saastst*******sssss***NEW8888888888888:

38 64.4 CONTAINMENT RADIUS FOR STkESS CALCULATIONS 39 48 1.67D-4 EUU1 VALENT AREA TO CALCULA1E CONTAINMENT NORMAL LEAKAGE-- 40 0.D0 NORMAL LEAKAGC IS ASSUMLD TO COME FROM THL ANNULAN COMrT 4: MASS OF WATER IN NEU1RUN SHIELD BAGS--WHEN BAGS RUPluRE 41 THEY DROP THEIR CON 1ENTS INIO REFUELING FOOL 10263. SEDIMENTAT!UN AREA FOR FISSION PRODUCT SETTLING s*THE REST OF THESE ARE NEW

   **ss  THESE ARE ONLY REGUIRED IF THE DETAILED CONTAINMENT FAILUhE MODEL Ib USdu.
   ****sts****stsssast**SNEWattsattsas***:

42 425. as NUMBER OF TkNDOMS IN HOOP DIRECTION IN THL LENGTH OF WALL GIVEN IN ITLM 43 l 43 .04390

   **                VOLUME OF REBAR PER UNIT AREA 0F OUTER WALL (EGUIV THICKNL:iS)

RUNNING IN fME HOOP DIRECTION 44 .0512 sa VOLUME OF RESAR PER UNIT AREA 0F 00lER WALL (EQUIV THICKNESS) ' KUNNING IN IHE Z DIRECTION 45 .1969 DIAMETER OF HOOP TENDONS 46 164. as HEIGHT OF fME CYLINDHICAL PART OF THE CONTAINMENT WALL ABOVE THAT PART OF THE WALL REPRESENIED IN DCOMPT ITtM NO. 5 i

A-5 88 (EG APPROX IHAT ABOVE THE OPERATING DECK) 47 16.4 Hk!GHT OF INTENNAL WALLS 48 .984 DISPLACEMENT IN AXIAL DIRECTION WHICH IS SUFFICIENT TO TEAR i 88 THE CONTM1 WALL (EG A1 A PENETRATION) ! 49 .984 SAME AS 46 FOR 1HE RADIAL DIRECTION 88 88 888888888888888888888888888888888888888888888888888888888888888888888888 SLOWER COMPARTMENT (OR 'B' COMPT) 888888888888888888888888888888888888888888888888888888888888888888888888 01 45. DISIANCE FROM FLOGR TO 10P OF S COMPARTMENT 02 3375. AR(A 0F CORIUM POOLI THIS MUST BE LEb51HAN THE CEA 0F 88 THE FLOUR (ENIERED BELOW) 1 03 2.162 HEIGHT OF CURS ON FLOOR (OVER WHICH WATER OVERFLOWS TO C) 04 1840 CHARAC. CRUSS-SEC AREA OF LOWER COMPT FOR BURN TIME CALCS 05 2.523E5 FREE VOLUME 0A 40. VERTICAL DISTANCE FROM THE CAVITY BYPASS FLOW AkEA Sa (EG AREA ARQUND VESbEL NUZZLES BUT SEE DEFINITION 88 IN CAVITY SECTION BELOW) TO THE CENTEN OF THE CAVITY END 88 0F fHE luiNetL FLOW AREA 07 25. DISTANCE FNOM THE FLOOR OF A TO THE OPENING FROM 3 IN10 0 08 0.00 FUR CASES WHERE INE 001ER BOUNDARY OF CONIMT IS A j 88 STEEL SHELL SEPERATED FROM A CONCRETE SHIELD WALLe i 88 ENTER DISTANCE BElWEEN THE TWO AND IREAT 1HE STEEL 88 SHELL AS A LINEN (ACOMPT AND DCOMPT OUIER WALLS)--

  **                                                         EMIER 0 OfMERWISE                                              _
 *a00TER WALL OF S DIVIDES IT FROM COMPT D 09     1125.0                                               AREA 0F OuitR WALL 10   0.0                                                   OuftR WALL LINER THICKNkSS 11   0.0                                                   GAP RESISTANCE OF OUTtR WALL LINER 12    4.                                                    THICKNESS OF OUTER WALL 13   0.92                                                   fMERMAL CONDUCTIVITY OF OUTER WALL 14    0.157                                                 SPECIFIC HEAT OF OU1EN WALL 15   144.                                                  DENSITY OF OUIER WALL 16    0                                                     ENTER 1 IF THE GU1ER WALL IS SOLID STEEL, 0 FOR CONCREIE 88NU1E THAT CORIUM IN B IS ASSUMED TO SEE ONLY UNE FACE OF THE INFERIOR 88 WALL FOR RADIATION CALCULATIONS 17   16290.                                                HALF SURFACE AREA 0F INTERIOR WALL 18    0                                                     INTERIOR WALL LINtR THICKNESS 19   0                                                     GAP RESISTANCE OF BUILDING INTtRIOR WALL LINER 20    4.0                                                   THICKNESS OF IN1ENIOR WALLS 21    0.92                                                  IHtRMAL CONDUCTIVITY OF INTERIOR WALLS 22    0.157                                                 SPECIFIC HEAT OF INIERIUR WALLS 23    144.

DtNSITY OF IN1ERIOR WALLS 24 6751. AREA 0F FLOOR (USE WATEM POOL AREA IF LESS) 25 0. FLOOR LINER IHICKNESS 26 0. GAP RESISTANCE OF FLOOR LINER 27 4.0 fMICKNESS OF FLOOR 28 0.92 THERMAL CONDUCTIVITY OF FLOOR 29 0.137 SPECIFIC HEAT OF FLOOR 30 144. DENSITY OF FLOOR 31 1.26ES MASS OF EQUIPMENT--THIS REFERS TO EUPT IN1ENNAL TU THIS

 **                                                          REGIONI 88                                                          THE PRIMARY SYSTEM MASS SHOULD N01 SE INCLUDED SINCE IT 88                                                          HAS A SPECIFIC IREAIMENT tLSEWHtRE 32    21422.                                                HEAT TRANSFER AREA 0F EQPT 8800ANTITY 43 IS USED FOR ALL EXTERNAL WALLS 33    8.8                                                   HEAT TRANSFEk COEFFICIENT TO BE USED ON THE OU1ER SURFALE
 **                                                          OF IHL CONTAINMENT UU1ER WALLS (EG IN A AND D) 8834 NOT USED

l A-6 l 35 0.00 FRACTIONAL AREA AVAILABLE FOR REVERSE FLOW ON B-1 FLOWPATH st COMPARLD TO THE FORWAND DIRECTION (EG DOE TO ICE l

   *s              CONDENSER D00R(S) SHUTTING)--THIS NU. MUST
   **               BE NONZER0 AND POSITIVE IN ICE CONDtNSER PLANTS--IGNORED IN              i as              LARGEe DRY CONIMTS 36       1.00   FRACTIONAL AREA AVAILABLE FOR REVERSE FLOW ON A-D FLOWPATH
   **               (EU AIR REIURN FAN FLOW DAMPERS IN ICE CONDEN$tRS) as              ENTER 1 IF NO DAMPER 37      384. FLOW AREA FROM B INTO D 38       1092. FLOW AREA FkOM B TO A 39     0.        NUMBER OF IUNITERS/ IGNITION SOURCES IN 3 40      0.       AVG DISTANCE OF THESE FROM THk CEILING OF B 41     27.       HeIUMT OF FLOOR OF 3 A80VE FLOOR OF C 42      4500.. SEDIMENTATION AREA st asssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssst 4 CAVITY (CCOMPT)                                                                 .

sassassssssssssssssssssssssssssssssss**ss**ssssssssssssssssss***stssets: salHE CAVITY INCLUDES ALL THE VULUME DELOW THE RtATOR N04ZLES INSIDE , asTHE BIOLOGICAL SHIELD AND ALL THE VOL OUT TO WHkRE 1HL TUNNEL SLOPES UP ss 4: NOTE THAT THE CAVITY HAS TWO FLOWPATHS-

• TUNNEL' REFERS TO A WATER s AND PERHAPS CORIUM FLOW PA1H THAT ENIERS NEAR INE BASE OF INE CAVIiYI

  *** BYPASS' REFERS TO A FLOWPATH HIGHER IN THE CAVITYI THIS CDULD BE THE stANEA AROUND THE RV N0lZLESe OR IN THE CASE OF SOME PLANTSe BLOWOUT PANELS ssHIGHER IN THE CAVITY--THE BYFASS AREA IS ASSUMtD TO EMPTY INTO B sa asIN SOME PLANTS WATER CAN FLOW DOWN FROM THL REFUELING FOOL TO THk CAVITY               -

asAND IN SUMEe CURIUM CAN BE EN1 RAINED UP TO IHE UPPER CUMPARTMENT AROUND ssTHE RV ANNULUS--AT PRESENT GAS IS NOT EXCHANGED SETWEEN C AND A HOWEVER ss

  **IN MANY SEQUkNCES: NAT. CIRC. IS SET UP WHEREBY COLD GAS ENTERS THE ssCAVITY IHROUGH THE IUMNEle IS HEA1ED BY PASSING OVER CORIUMe AND LEAVES ssTHROUGH THE. BYPASS AREA 01    7.74       RYPASS (NON-TUNhEL) FLOW AREA COUPLING CAVITY TO LOWER /UFFEk
  **               COMPARTMENTSI IHIS SHOULD BE 1HE LIMITING FLOW ARE4e EG 38               THE AREA AROUNU THE N0ZZLES AS THEY FENETkATE THE 810 LOGICAL ss               SHIELD OR THE ANNULAR FLOW AREA BETWEEN IHE RV AND THE SHIELD 02    476.4      AREA 0F CAVITY POOL--THIS INCLUDES KEYWAY EIC WHERE APFLIL 03    233.0      CHARAC. CROSS-SEC AREA 0F COMPT FOR BURN TIME CALCULATION 04    17.        HEIGHT OF VLSSEL AD0VE BOTTOM OF CAVITY 05     158.2      FUNNEL CRUSS-SECTNL AREA 06    253        LARGEST CHARAC CROSS-SECTNL AREA 1 HAT CORIUM MUST ss               TRAVERSED UN IIS WAY TO THE OPENING WHERE IT MAY BE
  **               EN1RAINtB OR FLOODED TO COMPTS A OR B--IN FLAN 15 WITH
  **               bO1 TOM HEAD PENtlRATIONSe IMIS WILL TYPICALLY BE THE
  **               'KLYWAY' AREA (THIS IS USED TO CALCULATE THE MINIMUM
  **               VELOCITY WHICH CAN kNIRAIN IHE CORIUM AND WAftR) 07    16935      CAVITY FRET VOLUMk 08    22.        HEIGHT OF TOP 0F TUNNEL AB0Vh CAVITY FLOOR (MtASUkED AT ss               CAVITY tND OF THE TUNNEL IF IT SLOPES) 09    2670.      AREA 0F CAVITY OUTEk WALLS 10    0.0        LINER THICKNESS 11    0.0        LINER GAP RESISTANCE 12    10.        THICKNESS OF WALL (UR ptP1H TO BE MUDtLLED FOR HEAT 88               TRANSFER IF IT IS VERY DEEP) 13    0.92       fHERMAL CONDUCTIVITY OF WALL 14    0.157      SPECIFIC HEAT OF WALL 15    144.       DiNSITY OF WALL 16      0.       NUMBER OF IGNITION SOURCES IN C                                        -

l

                                                                -------s

A-7 17 0. AVG DISTANCE OF THESE FROM IHE CEILING 18 794. SEDIMEN1ATION AREA 19 43.1 MINIMUM FLOW AREA WHICH CONNECTS CAVITY TO LOWEk L0e r a THROUGH TUNNEL ss

• CONCRETE AND CONTAINMENI SHELL

           *SI asFIRST 12 OUANTITIES ARE USED FOR ALL CONCRElE DECOMF0SITION CALCS.

• SUNLESS OTHERWISE STAIEDe CONCRtTE PROPERTIES ARE FOR ' PURE' 0 AGE S CIFTC HEAT OF CONCRE1E (UP TO MELT POINT) 02 1M3. MELTING TEMPERATURE UF CONCRElE (BETk EN SOLIDUS AND LIQUIDUS)

           $sst*SNEW: DEFINITIONS OF FOLLOWING HAVE CHANGEDassass8
           *sALL INE CONCREIE MASS FRACS SHOULD ADD UP IO ROUGHLY 0.9 TO 1.1                *
           *s(THE DIFFERENCE SETWEEN TNE SUM AND 1 IS DUE TU N0Y ACCOUNIING FOR stSMALL PERCENTAGES UF RELATIVELY INtRT MATERIALS: EG AL203 AND MGO) 03      .029      MASS FRACTION OF CONCRETE THAT IS FREE WATER 05 ITE
                  .015 SHOULD            L        0 'hA M. 'TTE90NEF " "

Mass PRACTION OF CONLNEfE lHAT IS CO2 06 2.74ES ENERGT ASSOR8ED IN ENDOTHENMIC CHEMICAL REACTIONS

           **               DURING CUNCRETE DECOMPOSITION 07      5.5E5     LATENT HEAT OF MELTING
.         08      1.8E-2    MASS FRACTION OF CONCRElE THAT IS NA20 l          09      5.4E-2      SAME FOR K20 i          10      0.55        SAME FOR SIO2 11     0.30         SAME FOR CAO + OTHER LESS VOLATILE CONCREfE COMPONENfS j

assasstasatssssastsNEW88888888888888844 12 183. RESAR DENSITY ==DCSRCN (MASS OF REBAR FEk UNIT VULUME OF

          **                  REINFORCED CONCRETE) = KG SIELL / M883 S1 EEL + CONCRETE
          **                 RELATED T0 (KG STEEL / KG CONCktTE)==R SY:

sa DCSRCN= RSDCN0/(1 + RSDCN0/DCS) WHERE DCN0 IS THE l ** VIRGIN CONCRETE DENSITT==2300 KU/M 33 AND DC5== VIRGIN

          **                  SIEEL OtNS!1Y==8000 KG/M*s3 -- CONSIDER THOSE VALUES
          **                 HARD-WIRED BECAUSE THEY WILL BE USED BY MAAP 1NIEkNALLY.

( ssREMAINDER OF THE QUANTITIES ANE USED IN IHE CONTAINMENT FAILURE MODEL l *8AND NEED NOT BE SUPPLIED IF THE ' SIMPLE

• MODEL IS USED (SEE GENERAL

          **MOIES SECTION) 4 NOTE: FOR FREE-STANDING STEEL CONTAINMENIS, YOU NEED SUFPLY ONLY THE
          **' LINER' PROPERTIES (WHICH ARE TAKtN TO DESCRISE IHE STEEL SHELL)
         **AND THE STEEL THICKNESS (STEEL THICKNtSS IS INPUI AS 38' LINER' THILKNESS IN THE UPPER AND ANNULAR COMPARTMENT SECTIONS)--

88SEE GENERAL NOTES SECTION 13 3.E11 ELASTIC YOUNOS MODULUS FOR TENDONS 14 1.99E11 ELASTIC YOUNGS MODULUS FOR REWAR 15 3.9/E9 PLASTIC YOUNOS MODULUS FOR TENDONS 16 1.4E9 PLASTIC YOUN6S MODULUS FOR REBAR

17 9.7E8 PRESIRESS ON HOOP IENDONS 18 1.01E9 PRESTRESS ON AXIAL TENDONS 19 1.53E9 TENDON Y!tLD STRESS 20 4 117[8 RESAR YIELD STRESS 21 1.6 m TENDON ULTIMATE STRESS 22 6 2E8 RESAR ULTIMATE STRESS 23 1.99E11 ELASTIC YOUNGS MODULUS FOR LINER 24 1.4E9 FLASTIC YOUNGS MODULUS FOR LINER l 23 4 137E8 LINER YIELD SIRESS 26 6.2E8 LINER FAILukt STRESS 84 85

1 A-8  ; I

                                                        *BR sssssssssssssssssssssssssssssssssssssssssssssssssssanssacessassassassses SCONTROL CARDS lieli"UNililli'lilli'fittii'ii'lif'lill' 01              1       ENTEN A 0 TO USE FAST STEAM TABLES IN PRI SYS WHEN FUSSIBLE 4tilli"Ji'litti' tiff!!'t1 l' 02              1       ENTER A 0 TO USE FAST STE M TABLES IN CONiMT WHkN POSSIBLE 03              1       INTLORATION METHOD: RUN6E-KUTTA ORDER (1 OR 2)3 1 IS
                                                        **                     RECOMMENDED 04             29       UNIT NUMBER (' TAPE' NO. IN CDC JAkGON) sa                      fu W1st RESTART FILES FOR MAIN PROGRAM FROM THIS RUN 2

06 30 UNIT NUMBER TO WRITE RESTART FILES FOR HEATUP FNOM THIS RUN ss07 NOT USED 08 10 UNIT NUMMR TO PUI FRI SYSTEM OUTPUf ON 09 10 UNIT NUMBER TO PUT CONTAINMENT GUTPUT ON (MOST USERS PUT sa IN THE SAME No. WHICH APPENUS THE TWO FILES)

• 10 31 UNIT NUMBER FOR THE FIRST PLOT FILE (UfMERS SEQUEV 4AL) 11 39 UNIT NUMBER FOR SCENARIO FILE ssNEXT 3 OUANTITIES LUNTROL INE PLOT POINT STORAGE FREQUENCY ',EE V0L 1 0F -

stOSER'S MANUAL) 12 250 NON-SPIKE NUMutR OF POINTS (AVENA 6E BEHAV!1P 510 RED 13 15 NUMBER OF POINTS STORED DURING A SP!KE (TO WESOLVE FAST as TRANSIENTS) 14 800 MAXIMUM NUMBER OF PLOT POINTS ALLOWED PER PLOT FILE i' s*SEE ESF LINEUP MENU IN SUBROUTINE ENGSAF WRI1E-UP IN

                                                       **VOL 2 0F USER'S r.ANUAL FOR NEXT TWO ENTRIES 15             2       ESF PUMP LINEUP IN RECIAC (1 FOR ZIONs 2 FOR SE000YAN)

N 0 E UfBAhh AN O fME W NO H MAAP at WAS WRITTEN FOR B AND W PLANTS WHOSE OTSG LOWLR TUDESNEEIS

                                                       **                      LIE BELOW THE LEVEL OF INE PRIMARY SYSTtM NUZILES--NOT
                                                       $8 13      bE          kR    WRI) AUX                                                                        FDP    h     ALONE AUX RUNS (OR 0 NOT TO WRITE DA1A) i sassassssastsattssEWassas**stsasssssssal 19             0        FILE NO. TO READ AUX DATA FNOM (IF THIS NUMBER IS NONZER0s ONLY

, sa THE AUX BUILDING MODELS ARE RUNE l'HE INPUT T/H DATA FROM THE

                                                       **                      CONTAINMENT HAVING BEEN RECORDED FROM A PREVIOUS RUN) 20              4       NUMBER OF NODES IN INE AUX BUILDING (MAX =$i!F 0. INE AUX BLDNG
                                                       **                      MODELS ARE NOT RUNE BUI A FILE MAY S1ILL BE CREATED FOR s*                      SUBSEUUENT STAND-ALONE AUX BUILDING ANALYSES BY SUPPLYING ss                      A NONZERO NO. FOR ITEM 18) sassss**ssssssssssss*****sssssss**********stsas***s**sssssssssssssssssas

• CORE sasssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssss:

01 .031167 FUEL PIN UUTER DIAMETER 02 46920 INTIAL ZIRCALLOY MASS 03 50952. NUMBER OF FUEL PINS 04 222739. TOTAL 002 MASS

                                                       ** N H               MA    S     EM   C 05            10.132     ELEVATION OF BOTTOM OF ACTIVE FUEL ABOVL BOTTOM 0F VESSEL 04           22.132     ELEVATION OF TOP OF ACTIVE FUEL AB0VE 80TIOM OF VESSEL 07            8766.      TIME OF IRRADIATION 08           1.16E10    FULL POWER
                                                       *sTHE CORE NODALIZATION ADMITS UP TO 70 N0DESI IN ADDITION, NO MORE THAN ss20 RUWS MAY BE USED AND NO MORE 1HAN 7 RINGS OR COLUMNS

WHATEVLR NODALIZATION IF. USEDs INSERT PEAKING F ACTORS INio AFFNOFRIATE 1

I

A-9

                               $$kN1RY NUM8ERS (EG SECOND RING FROM INSIDE RADIAL PEAKING FACTOR IS
                               $8ALWAYS ITEM 32 NO MATTER HOW MANY AXI R N0 DES) 88f0P N00E !$ UNFUELED (FISSION GAS PLENUM ETC) AND MUST HAVE ZERO FEAKING
                               $8 FACTOR 09             7                    NUMBER OF RINGS 10             10                  NUMBER OF RUWS 88 TOP R0W IS STRUCTURE (UPPEk PLENA ETC)e 50 PEAKsZENO. THE FULLOWING 88 9-ROW VRUES HAVE SEEN QSIAINtB BY AVLRAGING 10-ROW VALUES.

88 THIS MEANS THAT R0W 5 HAS A SLIGNILY LOWER FEAK THAN OTHENWISE-- 88 RkEVALUATE THESE PEAKING FACTORS IF DESIRtDe BUT DIFFERtNCES ARE SMALL. 11 0.498 AXIR PEAKING FACTOR 80 Tium 12 0./34 AXIAL PEAKING FACTOR 13 1 126 AXIAL PEAKING FACTOR 14 1.402 AXIAL PEAK!NG FACTOR 15 1.480 AXIAL PEAKING FACTOR 16 1.402 AXIAL PEAKING FACTOR 17 1.126 AXIAL PEAKING FACTOR 18 0./34 AXIAL PEAKING FACTOR ' 19 0.498 AXIAL PEAKING FACTOR 8888888888888888888MkW88888888888888888 20 0.000 AXIAL PEAKING FACTOR TOP

                               $8 ENTRIES 21-30 AXIAL PEAKING FACTORS NOT USED IN 1HIS NODRIZATION 31            1.09                 RADIAL PEAKING FACTOR                                                         INSIDE

32 1.11 RADIR PEAKING FACTOR 33 1.10 RADIR PEAKING FACTOR 34 1.115 RADIR PEAKING FACTOR 35 1.096 RADIAL PEAKING FACTOR 36 1.01 RADIAL PEAKING FACTOR ,

37 0.75 RADIAL PEAKING FACTOR OUISIDE  ! 38 0.047 AREA OR VOLUME FRACTIONS INSIDE ' 39 0.042 AREA OR VOLUME FRACTIONS 40 0.145 AREA OR VOLUME FRACTIONS 41 0.124 AREA OR VOLUME FRACTIONS 42 0.207 AREA OR VOLUME FRACTIONS 43 0.166 AREA OR VOLunt FRACTIONS 44 0.249 AREA OR VOLW9E FRACTIONS OUTSIDE 88FOLLOWING QUANTITIES CONTROL ANSI DECAY HLAT CALCULATION 45 32000 FUEL EXPOSURE AT SCRAM (ALWAYS IN 88 MEGAWATT-DAYS /METkIC TON NO MATIEN WHAT UNITS SELECTED) 46 .39 FUEL

• AlfMA' AT SHUIDOWN (FISSILE ISOTOPE 88 CAPTUkES/ FISSION) 47 .032 INITIAL kNNICHMENT OF FUEL IN ATOM FRACTION 48 .642 CONVER$10N RATIO (PRODUCTION RATE OF U-239/ ABSORPTION RAT

                               $8                                  IN FISSILE ISOTOPES) AT SMuTDOWN 49             .487                FRACTION OF FISSION POWER MADE DUt TO FISSIONS IN U-235
                               $8                                 AND PU-241 AT SHUTDOWN 50             .443                SAME AS 49 FOR PU '39 51              .049                SAME AS 49 FOR U-2h (FAST FISSIONS) 52            6.5E-4               FRACTIONAL ZR02 MASS (COMPARED TO ZK MASS) AT TIME O 8888888888888ALL THE REMAINDER IN lHIS SECTION ARE NEW8888 53            .01344               FUEL PELLET RADIUS 54            3.185                CORE FLOW AREA IN THE BYFASS AREA BE1 WEEN THE CUKE BAFFLE
                              $8                                  AND THE CONE SARRkt (ENSURE fHIS IS CONSISfENT WIlH FRI

. 88 SYSTEM CORE FLOW AREA PARAMtTER NO. 5) 88 PARAMETERS S5-60 ARE USED FOR CALCULATING BALLOONING (DATA SHOWN IS 88MOSTLT FROM TMI REPORTS) 55 1.87bE-3 CLAD THICKNESS 56 GAS VOLUME FER FUEL PIN b7 450. AS-9UILT ROOM TEMP FUEL FIN FILL GAS FRESSURE 58 CURE SUPPORT PLA1E MASS--lHIS PLAIE IS MELTkD BY IHE DEBRIS 1 4

  -w -
       , - - -n- - - - - - - . , - . - -             -..,-..
                                                             - -- - ,,.. - - . - - , , - - - , - - - - , - , - - , - - , ..                                       ~ , _ , _ _ - - , , - - - - - , , - - - - - - - , .         --+

1 l l A-10 as AS If LEAVES THE ORIGINAL CORE B0UNDARY 59 FRACTION OF THE TOTAL FUEL PIN GAS VOLUME WHICH IS

   **                CONTAINED IN THE LOWER GAS PLENUM OF THE PIN N                   A U 0     ORYhb          P THIS ENIRT SHOULD DE THL
   **                CIRCUMtERENCE OF INE BAFFLE) 62        0.      ' FLOW AREA PEN R0W' IN CORE BAFFLE (IMPOR1ANI ONLY IF
   **                IN-VESSEL NAIWAL CIRCEATION RETURN LEG IS IN SAFFLE-CORE 83                BARREL ANNULUS--SEE sMKL)--THIS REPREMNIS THk APPROXIMATE as                FLOW AREA AVAILABLE AS INE PLOW (URNS SIDEWAYS AND PENETRATES sa                THE CORE 63        0.      FOR TMI-TYPE GEOMEIRIES THE FLOW AREA THROUGH EACH CORE 38                FuRMER PLATE IN AXIAL DIRECTION                               ,

64 0. FOR TMI-TYFE CORESeNUMbER OF CORE FORMLR PLATLS IN THE

   **                BAFFLE-CORE ANNULUS as
   **sssssssssssas**sssssssssssssssssssss**ssssssssssssssssssssssssssssssas s!CE CONDENSER (*I' COMPARTMENT) asssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssstas 01     0.E0       TOTAL VOLUME INCLUDING 1HE ICE 302 EXIT GAS TEMPERATURE--THIS IS THE TLMPENATUNE OF GAS LEAVING THE as     ICE BOX (SEE WRIIE-UP FUR SufROUTINE HTICE IN VOL 2 0F USER'S MAN) 8803 INITIAL TLMFERATURE OF THE ICE
   **04 SPECIFIC VOLUME OF ICE--NOTE THE TOTAL VOLUME MINUS IHE ICE MASS as     TIMES THE SPEC VOL SHOULD BC THE FREE VOLUML 8:05 FLOOR AREA 0F WATLR SUMP IN 80TTOM OF ICE CONDENSER 306 HEIGHB 0F SUMP (IE CUNS OVER WHICH WATER DRAINS IN10 B) 3:07 VERTICAL HEIGHT OF ICE BOX
   *s00 FLOW AREA BETWEEN LOWLR COMFARTMENT AND *lHE ICE CONDENSEN
   **09 SEDIMENTATION AREA as sessas**ssssssssssssssssssssssssssssssssssssssssssss**sassassssssssssss
   *UPLENUM (UFPER PLENUM UF ICE CUNDtNSER- 'u' COMFARTMENT) 1   asssssssstassss**sesssssssssssssssssssssssssssssssssssssssssssssssssssas 01     0.         VULUME--ENBER 0 VULUME FOR LARGEe DRY CUNTAINMENTS ss02 CHARACTERISTIC CROSS-SEC AREA 0F THE COMPT FOR BURNS

03 HEIGHT OF UPPER PLENUM i a:04 LIMITING FLOW AREA WHICH CDUFLCS THE ICE CONDtNSER TO THE UFFER

COMPARIMENI--IE USE 1HE LESStR OF 1HE UPPER PLEN TO UFPfR COMPT sa FLOW AREA OR THAT C00 FLING THE UFFER FLEN TO THt ICE CONJ.

3:05 NUMBER OF IGNIfERS IN U 3806 AVERAGE DISTANCE OF IGNITERS BELOW THE CEILING OF U

   *s07 AVO DISTANCE FROM THE TOP OF UP PLEN TO IHE PORTION OF THE
   **      CEILING OF THE UFFEN COMPT WHICH IS JUST OVtR THk EXIT 001 0F Ui as      THIS IS USED TO CRC 1 A1E LOCAL BURNING IN THE UPPER COMPT
   *s      INf!ATED BY FLAME PRW AGATION OUT OF U
   **08 SEDIMENIATION AREA IN U sa ss assassassassssss*8ssssssss**sas****ssssssssssssssssssssssssssst*s888888

• ANNULAR COMPARIMENT ('D' COMPARIMtNT) sattssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssas as ssIN LARGE DRY CONTAINMENTS:

asTHIS COMPARTMENT REFRESENTS THE VOLUME BEfWEEN THL CRANE WALL (IF ANY) asAND 1HE CONIMT WALLS AND PEIWEEN THE DECK AND lHE LOWER COMPT FLUOR--

   **!F NO CLEAR DISTINCTIONe ARBITRARILY DIVIDE THE SFACE BELOW THE UFFER l

I k

A-ll asCOMPT AND USE LANGE FLOW ANEAS TO KtEP THE GAS WELL MIXtD--AT PRESENT:

              **CORIUM sa                      IS ASSUMED NOT TO GE1 INf0 THIS C0WARTMEN) g!N {                       DEgSQgVOLUMEREFLECTSTHE' DEAD-END'COMPARTkNTS 02               4888.          AREA 0F WATER POOL 03                  0.         DISTANCE T"                               F     OF D     A 81!Hb.Af3"#J"lHILsat"d* g "80VJ                                                             "'    'ig*"""c"*""

• FLOOR OF B i 04 0.0312S WALL LINER THICKNESS 07 0.28 GAP RESISTANCE OF WALL LINER 08 4.5 THICKn SS OF WALL 09 0.92 fHtRMAL CONDUCTIV!iY OF WALL 10 0.157 SPECIFIC HEAT OF WALL 11 144. DENSILY OF WALL 12 0.

            **                              ENTER IS MADE OF STELL A 1 IF Th OUTER WALL (CONfMT OUIEN BOUNDARY) 13                     0.

14 0. HEIGHI 0F CUNS SEPERATING D ANo 3 MEAsukED FNOM B'S FLOOR - 15 0. NUMBER OF IGNITLNS OR IGNITION SOURCES IN D 16 9774. AVG DISTANCE OF THESE FRGM TR CEILING SEDIMENTATION AREA

          **stsas****ssALL TNk REMAINDER IN THIS SECTION ARE NEWassa
          **1HE NtXT
         *sINERTIAL                      THREE PARAMtTEMS ARE USED TO DEFINE INE EFFICIENCY OF IMPACTION ssIN LARGE: DRY'S THESE PARAMETERS SHOULD CHARACTLkIZE
        **ssGRATES WHICH ARE ASSUMtB TO BE IN INE ANNULAR COMPARTMENT
        **!N ICE CONDENSER PLANTS THESE PARA 9ETERS (EVEN THOUGH LOCATkD             -

asIN THE ssFLOW ANNULAR AREAS AND STRAFCOMPARTMENT DAIA SCCTION) SHOULD REFLECT IMPACTION AND 17 733.2 WIDTHS IN THk TCE BOX--SEE EG POSTMA 18 0.010 IMPACT 10N AREA (ANtA 0F PANS IN GRAIES INAT INfkRCEPT FLOW) WIDTH OF GRATE PARS 19 4154.8 i

       **                                   FLOW     AREAGRATES)

TRAVtRSINO THROUGH GRATES (DEFINES FLOW VtLOCITY OF AENOSOLS ssNOTE: IF MORE THAN ONE LEVkL OF GRATES EXISTS: USE THE TOTAL IMFACTION AREA gGFALLTHEGRAIE$sANDTHEMAXIMUMFLOWAKEAATANYOFINEGRA1EELEVATIONS

      *SI asDETAILED sauSED                               CONTAINMLNr FAILURE MODEL INPUIS--IGNORE IF SIMrLE MODEL 20                        130.

• NUMBER OF TENDONS IN HOOP DIRECTION IN THE FART OF THE WALL 21 216. WHOSE AREA IS GIVEN IN ITtM S ABOVE 22 .01399 NUMSER OF TENDONS WHICH RUN IN THE AXIAL (VERTICAL) DIRECTION as UOLUME OF RtBAR PER UNIT AREA 0F OUTER WALL (EQUIV THICKNESS)

RUNNING IN THE H0OP DIRECTION 23 .06 DIAMtitR OF HOOP TkNDONS 24 .06E0 DIAMETER OF THk AXIAL TENDONS 26 .3 bNNG N HE A EC

   **                                      DISPLACEMENT IN AXIAL DIRECTION WHICH IS SUFFICIENT TO TEAR 27                         .3           THE CONTMT WALL (EG AT A FENEIRATION) sa                                     SAME AS 24 FOR THE RADIAL DIRECTION SBR
   *sssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssas

• ENGINEERED SAFEGUARDS

   *stsassasssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssses
   **IN BRITISH UNITS 4sFLOWNATES                                  SPECIFIED TO BE VOLUMETRIC SHOULD SE Mas 3/SECl OTHtR FLOWNATES s*IE ALL THOSE NOT EXPLICI(LY STAitD TO DE VOLUMETRIC

A-12

    **SNOULD E KG/SECl HEADS SHOULD BE IN MI PRESSUkES IN pal IN ENGLISH THL
    ** UNITS ARE RESPECTIVELY GPMsLBM/HReeTe PSIA--

asNOTE es TO MAAP/8WR USERS--GPM IS USED IN MAAP/PW INSTEAD OF Fiss3/HN

    **IN THE FOLLOWINGe* FANS' REFER TO FAN C00 LENS--(AIR RETUNN FANS IN asCONDENSER PLANTS) as asFOR KTTER ACCUNACYe YOU MAY ELECT TO INPUI ' SYSTEM' PUMP HEAD CUNVES WHICH                       )

as!NCLUDE IHE EFFECTS UF FNICTION IN IHE INLLT AND OUTLLT PIPING (WHICH IS <

    ** IGNORED IN MAAP)I IF YOU DO $0e BE SURE THE ASSUMPTIONS ON STATIL HLAD                           I a WHICH ARE USED IN tHEIR CALCULATION ARE CONSISLENT WItH 1HE PUMP ELEVATIONS stETC. WHICH ARE INPUI BELOW--THIS IS GENERALLY A FACTOR ONLY IN CRITICAL
    *sAPPLICATIONS SUCH AS FLED AND BLELD WHERE 1HE CHARGING PUMP FLOW IS
    ** BARELY (OR NOT) ADEQUATE TO MATCH DECAY HEAT 01       0 833     ACCUMULATUN PIPE DIAMETER 02     195.       FRESSURE SETPOINT FOR LPI

• 03 1336. PRLSSURE stTPOINT FOR HPI 04 615. INITIAL FRESSUNE OF ACCUMULATORS 05 84. 1EMPERA1URE OF REFULLING WATER STORAGE TANK (RWST)--IE

   **                 THE TANK FROM WHICH THE CHARGINGs HPI, LPle AND SFRAYS sa                DRAW THEIR WATER DURING INE INJECTION PHASE 06        100. TEMPERATUNE OF ACCUMULATORS 07     2.91E6       INITIAL MASS IN RWST 08     5.28E4      INITIAL MASS PER COLD LEG ACCUMULATOR 09      1319.5     AREA UF BASE OF RWST 10        18.3     LENGTH OF AN ACCUMULATOR PIPE 11     45.1        PRESSURE SETPOINT OF BLDG SPRAYS                                                '

12 300. PRESSUNE SETPOINT OF BLDG FANS 13 0 NUMBER OF OPERATING FAN COOLERS OR FANS 14 0.0 VOLUMETRIC FLOW THROUGH ONE FAN COOLEN OR FAN 15 2.165E-3 NUMINAL DIAMETER OF CONTAINMr.NT SPRAY DROPLETS AS THEY N 17 1350. 4 UME OL G ACCUMULATOR NUMsER OF OPERATIONAL COLD LEG ACLUMULATORS 18 1 NUM8tR OF OPERATIONAL HPI PUMPS 19 1 NUMBER OF OPERATIONAL LPI PUMPS 20 5 NUMBER OF ENTRIES USED IN HPI PUMP-HD CURVE TABLE (5 MAX) 21 3600. HIGHEST HEAD IN TABLE (UNITS ARE MEIENS) 2 3'00. 23' 3900. NtXT HIGHtST HE(AD IN HPI PUMP-MLAD CURVE TABLENEXT HIG 24 1650. NEXT HIGHEST HEAD IN HPI PUMP-HEAD CURVE TABLE 25 0.0 LOWEST HEAD IN HPI PUMP-HEAD CUhVE TABLE 26 0.0 VOLUME 1RIC FLOWRATE CORESFONDING TO FIRST ENTRY IN as THE PRESSUNE TABLE 27 325. NEXT VOL. FLOWRATE 28 425. NEXI VOL. FLOWATE 29 650. NtXT VOL. FLOWRATE 30 650. NEX1 VOL. FLOWATE 31 5 NUMBER OF ENTRIES USED IN LPI TABLE 32 470. HIGHEST HEAD IN LPI TABLE 33 425. NEXT HEAD 34 390. NEX1 HEAD 35 325. NtXT HEAD 36 0.0 NEXT HEAD 37 0.0 FIRST VOLUMETRIC FLOWRATE IN TABLE 38 2000. NEXT VOL. FLOWWATE 39 3000. MEXT VOL. FLOWRATE 40 4500. NEX VOL. FLOWRATE 41 4500. NtXT VOL. FLOWRATE 42 2687. CHARGING PUMP FRESSURE SEfPOIN1

l A-13 i l l 43 1.0 NUMBER OF WORKING CHARGING PUMPS 44 5 NUMpER OF ENIRIES IN CHARGING FUMP HEAD CUhVE TABLE 45 6000. FIRST HEAD 46 5800. NEXI HEAD 47 4800. NEXT HEAD 48 2000. NEXf HLAD 49 0.0 NEXT HEAD 50 0.0 FIRST VOL. FLOWkAIE 51 150. NkXT VOL. FLOWRATE 52 300. NEXY VOL. FLOWRATE 53 b50. NkXT VOL. FLOWRATE 54 550. NEXT VOL. FLOWNATE 55 160. AREA 0F BASE OF CONIMT SUMP 56 8 DEPTH OF CONTM1 SUMP

            **N0fE, IF DESIRED YOU CAN SUFPLY ONE NUMBhR--IF DO SO GIVE IT A LARGE      '
            **HEADe THkN A CONSTANT FLOW MODEL WILL BE USED 57         1      NUMBER OF USED LNIRIES IN SPRAY PUMP HEAD CURVES (5 MAX) 58      1000      FIRST ENTRY IN SFRAY PUMP HEAD TABLE
            ** HEADS 59-62 NOT USED IN 1HIS SCHEME 63     1 450-1    FIRST VOLUMkTRIC FLOW ENIRY IN SPRAY PUMP TABLE
            ** VOLUMETRIC FLOW VALUES 64-67 NOT USED IN THIS SCHEME
            ** FOR NPSH TABLES THk SAML FLOWS AS WERE GIVEN FOR HEAU CUNVES ARE
            ** ASSUMtD TO CORRESPOND T0 IHE NPSH HEADS GIVEN 68         28. NFSH (UNITS OF LENGTH) REQ'D FOR CHARGING PUMP
            **                 AT FIRST FLOW IN TABLE 69         28. NEXT NPSH ENTRY FOR CHARGING FUMPS 70         28. NEXT NPSH LNTRY FOR CHARGING PUMPS 71        28.      NEXT NFSH ENTRY FOR CHARGING FUMFS 72         28. NEXT NPSH ENTRY FOR CHARGING PUMPS 73         13.5    FIRST NPSH ENTRY FOR LPI 74         13.5    NEXT EN1RY FOR LPI 75        13.5     NEXf ENTRY FOR LFI 76         13.5    NtXT tNTRY FOR LFI 77        13.5     NEXT ENTRY FOR LFI 78        25.      FIRST NP5H ENTRY FOR HPI 79        25.      NEXf ENIRY FOR HPI 80        25.      NEXT LNTRY FOR HPI S1         25.      NEX ENTRY FOR HPI 82        25.      NEXT LNTRY FOR HPI 83         19.      FIRST NPSH EN1RY FOR SFRAY FUMPS 84         3.05     NEXT EN1RY FOR SPRAY FUMPS 85         3.05     NEXT ENTRY FOR SFRAY FUNFS 86         3.05     NEXT ENTRY FOR SPRAY PUMPS 87         3.05     NEX1 ENTRY FOR SPRAY FUMPS 88         1      NUMBtR OF OPERATING SPRAY FUMPS FOR UFFtR COMFARTMENT 89         0      NUMBER OF OPERATING SPRAY FUMPS FON LOWEN COMPAATMtNi 90         77.5   HEIGHT OF B01 TOM OF RWST ABOVE INE ENG SAFE FUMFS 91         22.3   HEISHT OF BOTTOM OF CONTAIN SUMP ABOVL THE EN6 SAFE FUMrS 92         38. ELEVATION OF INE RV INJECTIUN N0ZiLES ABOVE THE SI PUMPS 93    7577.       FLOW TMN00GH Out SFRAY FUMP WHEN ITEM 94 IS MEASUKEu 94    40.         DIFFERENTIAL PRESSURE ACROSS IHE SFRAY N04ZLES 95      0.0       MASS FLOWNATE OF EXIERNAL RWST REFLACEMtNr WATEN, IF ANY 96     .00278     TIME DELAY FOR HPI (IE TIME BtTWEEN THE ACTUATION AND WHEN
           **                ACTUAL OPERATION BEGINS) 97      .002778 TIME DELAY FOR LPI 98      .002778 TIME DELAf FOR CHARGING FUMPS 99      .00833    TIME DELAY FOR UPPER COMPAR1 MENT SFRAYS l           100     .00833    TIME DELAY FOR LOWER COMPARTMENI SFRAYS 101        5.0    TIME DELAY FOR FAN COOLERS 102               NUMBER OF TUBES IN A FAN COOLER
                                , , ,                         =-v-
                                        -ww,   ---------  ww-           -'        ""      '

A-14 103 OUTSIDE AREA 0F ALL IUBES IN A FAN COOLER 104 AREA 0F ALL FINS IN A FAN C00 lek 105 FAN C00LtR FIN EFFICIENCY 106 FAN COOLER INSIDE FOULING FACTOR 107 FAN COOLER FIN DIAMETER 108 FAN COOLER TUBE THICKNtSS 109 FAN C00LtR IUBE THERMAL CONDUCTIVITY 110 MINIMUM FLOW AREA THROUGH FAN COOLEN 111 FAN COOLER 10BE ID 112 5 NUMWER OF NODES USED TO MODEL FAN COOLER (5 MAX) 113 INLtf COOLING WATER fEMP TO FAN COOLER--NOTE IHIS IS

  **                                    ALSO USED AS THt COOLING WATEN TLMr FOR ALL OTHtR
   **'                                  SAFEGUARDS HEAT EXCHANGERS 114                                   INLET COOLING WATER FLOW TO A FAN COOLEN 115              1 NUMBtR OF LFI FUMFS UStD FOR RHR SPRAYS WHEN VALVE OPEN 116               0                   ENTER A 1 IF FANS /C00 LENS DISCHAR6E TO 880 TO D
   *s                                                                                           .
  **ESF HX'S
  ** CALCULATIONS CONTROLLED BY HEAT EXCHANGER TYPE
  ** HEAT EXCHANGER TYPE:
  **             -1          SET OUTLET TEMP OF HX TO RWST TEMFERATURE
  **              0          IS NO HX--0UTLET TEMP IS CONTMT SUMP TEMP
  **              1          STRAIGHT TU8E HX
  **              2          U-1UBE HX
  **!MPORTANT NOTE:
  **FOR HX TYPES 1 AND 2 EITHER SUPPLY ALL GEDMLIRIC FARAnt1ERS
  **0R fHE N1U (NUMBtR OF TRANSFER UNITS) FER HX--ALL KNOWN USERS D0
  **THE LATTER--NTUS ARE AVAILABLE BY CONSULTING NAMLFLATE DATA AND
  **USING GRAPHS INe FUR EXAMPLE HOLMANs HEAT TRANSPER
  **ALL FARAMETERS ARE ON A FER HX BASIS ff7           2.D0                 TYRE OF HX FOR SFRAY 118            0.D0                NUMBER OF 10BES IN SFRAY HXS 119           0.00               NUMBER OF SHELL SIDE BAFFLES IN SFRAY HXS 120            0.D0                SPRAY HX IUBE ID 121            0.00                SFRAY HX tub 5 THICKNESS 122           0.D0                   IUBE 70 IUBE SEPARATION IN SFRAY HX 123            0.00                SHtLL LENGTH IN SFRAY HX 124            0.D0                   fHtRMAL CONDUCTIVITY OF SPRAY HX TUBES 125            0.00              LARGEST FERP DISTANCE FROM SHELL 10 BAFFLE (' BAFFLE CUf')

126 0.00 SHELL f0 lube CLEARANCE AT OUTSIDE OF SPNAY HX IUBE BDL 127 2.35E6 SFRAY HX COOLING WATER MASS FLOWNA1E

 **128 NOT USED l 129            2                  TYFE OF HX FOR RHN 130            0.D0               NUMBtR OF 10BES IN RHR HXS 131            0.DO              NUMWER OF BAFFLES IN RHk HXS 132            0.D0               TUBt ID IN RHR HXS 133            0.D0              TUBE THICKNLSS IN RHR HXS 134            0.D0                 IUBE TO IUBE SEPERATION IN RHR HXS 135            0.D0             SHELL LENGTH IN RHN HXS 136            0.00                fuBE 1HtRMAL CONDUCTIVIlY IN RHR HXS 137            0.D0             BAFFLE CUI DISTANCE IN RHW HXS (SEL 125) 138            0.D0              SHLLL TO IUBE CLEARANCE AT OUTSIDE OF RHR HX IUBE BUNDLE 139            2.475E6 RHR HX COOLING WAIER MASS FLOWRATE 140            0.996             SFRAY HX NTU 141            1.416           RHR HX NTU 142            0.D0             SHELL ID OF SFRAY RECIRC HX 143            0.00            SHELL ID OF RHR RECIRC HX
 **tNTER 4ERO VOLUME FOR ITEM 148 IF NO UHI SYSTEM I

l

A-15 38144 INITIAL MASS IN THL UNI WATER ACCUMULATOR 4:143 LkNG1H OF 1HE UNI P!PE TO INE RV satte DIAMETER OF THE UNI PIPE as147 INTIAL PRES $URE OF 1HE UN! ACCUMULATOR 140 0.00 TOTAL (WATER + gab) VOLUME IN THE UNI ACCUMULATORh as149 FAILURE DIFFERENTIAL PRESSURE OF 1HE UNI PIPE RUP1URE DISK ssTHE ' CAVITY INJECTION SYSTEM' IS (RARELY) USED TO SIMULATE A

**PROPOSkD DEDICA1ED LSF WHICH MtRELY DUMPS WA1ER INTO INE CAVITY 150     0.00      TOTAL MASS IN THE CAVITY INJLCTION SYS1EM TANK 151     0.D0      MASS PLOWR4st OF INE CAU INJ SYS1EM WHEN ACTIVATED ss      USER HAS THE OPTION TO THROTTLE ESF SYSTEMS AT LESS THAN sa      lHEIR FULL FLOW GIVtN 1HE CONDITIONS EXISTING--TO 00 THIS,
**      ENIER FOR THE APPROPRIATE SYSTEM (AND FOR THE AFW IN THE STM 88      GENERATOR SECTION) A TOTAL FLOWRATE DESIREDI INE CODE WILL USE as      THE MINIMUM OF THIS FLOW AND THAT CALCULAft0 FNOM THE HEAD CUNVtS st      AND 1HE NO. OF OPERATIONAL PUMPSIIF OPERATOR !$N'T 1HR01TLINGe     -
**      ENTER A LARGE No.11F HE CHANGES THE DEGREE OF THNOTTLINGs ENTEN 48      PARAMETLR CHANGES USING INTERVENTION NO. 1000 IN CONTROL CARDS 152     7.934E9    THROTTLED FLOW FOR LPI SYSTEM (TOTAL) 153     7.934E9    SAME FOR HPI 154     7.936E9    SAME FOR CHARGING PUM S                                   -

155 7.936E9 SAME FOR UPPER COMPT NORMAL SPRAYS 15/ 7.934E9 SAnt FOR UFFER COMPT RNR SPRAYS (WHkN AC1!VAIED) { 157 7.936E9 SAME FOR LOWkR COMPT SPRAYS as as ssssssssssssssssssssssssssssssssss33333333333333333333333333333333333333 s!NITIAL CONDITIONS sessssss***ssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssa 01 591.8 NOMINAL FULL POWER PRIMARY SYS1EM WATER 1EMPERAl'URE 02 2250. NOMINAL FULL POWLR FRIMARY SYSTEM PRES $ uke 03 28.05 PRESSUNIZER WATER LEVEL (AB0VE BOTTOM 0F PZR HEAD) 04 14.7 CONIAINMENT BUILDING PRESSuht 05 120 LUWtR CONTAINMENT BUILDING COMPART E NTS (ALL DUT 48 UPPER COMPT ANU ICE CONUENSEN) TEMPERATUNE 06 0. ILE CONDENSER GAS ILMPERAluREe WHERE APPLICABLE 07 1. LOWER CONI AINMkNT BUILDING COMPARTMNIS REL. HUMIDITY (0-08 0 INITIAL ICE MASS 09 99253. INITIAL MASS OF WATER ON SECONDARY SIDE OF EACH S/G

• • VALUE TAREN FROM M00tL F SG TSH DATA FOR MILLSTONE 10 120. INITIAL TEMPERA 1UNE OF CONIAINMENT CONCRETE ANU
  • METAL $TRUCfuRES 11 944. INITIAL PRES $URE ON SEC SIDE OF S/G'S sauPPER COMPT CONDITIONS COULD DE DIFFERtNT IN ICE CONDENSERS 12 1.00 UPPER COMPARTMENT REL HUMIDITY (0-1) 13 100. UPPER COMPARTMENT 1EMPERATURE 14 591 8 INITIAL PNIMARY SYSTEM WATER TEMeENATUNE FOR THIS RUN 15 2250 INIIIAL PRIMARY SYSlEM PRESSURE FOR 1HIS RUN 16 0.0 AMOUNI OF SUPENHEAT AT EXIT N0Z OF AN OTSGI IGN0hE0 FOR
  • U-TUBE STEAM UEMERATORS ss sessssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssas SPRIMARY SYSIEM
• sssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssas
  • UNLESS 01HERWISE N0iEDe ALL ELEVATIONS IN fMIS SECTION SHOULD BE a: REFERENCED TO THE LOWEST FOINT OF THE INSIDE OF THE RV HEAU ssWHEN A PARAMETER SUCH AS INE VOLUME OF THE DOWNCOMER IS CALLED FOR, ssTHE ACTUAL DOWNCOMkR VOLUMt SHOULDe OF COUNSE BE USED EVEN IHOUGH THE 4sMAAP N0DALIZATION LUMPS UTHER VOLUMES WI1H 1HE DOWMCOMER VOLUME (IHE l

_ - - _ - _ - - - - l

A-16 asLUMPING IS DONE INTERNALLY IN THE CODE) 01 4 NUMBER UF COLD LEGS 02 2.42 INNtR DIAMETER OF A HOT LEG PIFE sessssssssssssssssssNtWassssssssssssssss: 03 INSIDE RADIUS OF THE CYLINDRICAL FANT OF THi REACTOR VEhSEL sassasssssssssssssssMEWasassassassesssss: 04 VOLUME WICH IS INSI T6E CORE BARREL AND LIES BETWitN s: 1HE pulT0M OF IHg C AND 1HE LINE WHICH DENOTES INE TOP ss OF THE RV HEAD (IE BOTTOM OF THE RV CYLINDRICAL SECTION) 0S 51.325 PLOW AREA 0F CORE PLUS CUME BYPASS AREA ' 06 86.01 VOLUME OF HORIZONIAL RUN OF PIPE IN ONk COLD LEG FKUM as 1HE REALIOR VESSEL QUT TO 1HE MAIN COOLANT PUMP l 07 .0623 RADIUS OF VESSEL PENtlRATION--IF NO VESSEL PENElRATION l ss (tG SOME CE PLANTS) UsE 1HE ASSUMED INITIAL RADIUS OF sa FAILUNE WEN THE RV HLAD FAILS DUL TO CORIUM ATTACK ANU

              **               SUPPLY 1 FOR 1HE NO. OF FAILED PkMtTRATIONS IN SMODtl 08    1.2E7       ENERGY INFUf FROM ONt PRIMARY SYSTEM PUMP (WHtN RUNNING) 09    0.0          IUTAL MAKEUP FLOW TO 1HE PRIMARY SYS1LM--UN0kR NORMAL                        ,

as OFERATION SHOULD EQUAL LEfDO W FLCW BELOWITHIS IS USED ss MAINLY IN thE TMI SCENARIO AND MOST USkRS WILL INPUT ZER05

             **                THIS WATER !$ NOT SubTRACTLD FROM THL RWST ANU CONTINutS ss                (IF POWER IS AVAILABLE) UNTIL MANUALLY SHUT OFF 10    S40.        TEMPERATURE OF MAKEur WATENe IF ANie GIVtM IN 09 11    2.2917      INNtR DIAMEfER OF A COLD LEG P!PE 12    28.3S      ELEVATION OF THE N0ZZLE WICH ATTACHkS THE SUNGE LINE ss                TU 1HE H0f LEG--fMIS MUST SE GREAftR IHAN IIEM 47

NOTE: IT IS HELPFUL IN LOCAS (ESP SMALL 8REAKS) TO AVOID i a: PUTTING IHE BNtAK ELEVATION IN THE VICIM11Y OF THE SURGE LINEl 48 ARTIFICIALLY INCREASING THE

            ** ELEVATION OF THE SURGE LINE 0.S-1 METER OR SU AB0VE THE DREAK IS SUGGESTED satuR1HER IT IS HELPPUL ID AVOID PUITING BREAKS NEAR 1HE ELtVATION OF INE ssTUBESHtET IN U-TUBE TYPE S/G PRIMARY SYSTLMS--80fH OF THESE MEASUNES s HELP AVOID WATER SLOSHING INTO AND 00T OF NODES (WHICH CRANKS IHE TIME STEP 4DOWN) ANU WILL GREATLY DECREASE RUNNING TIME Af NEbLIGIBLE LUSS OF ACCUNACY
            ****ssssssssssssssssNEWassssssssssssssss:

13 6 ENIER BROKLN LOOP BREAK LOCATION KEY (NODE NO.)I

             **                3--BR0 KEN HOT LEG NODE
            **                4--BR0ktN HOT LEG 'TUWE' NODE (8 AND W ONLY) l             ss                6--8R0 KEN INIERMEDIATE LEO NODE (BkfWtEN PUMP AND COLD SIDE OF
            **                     S/G) es                7--BROKEN COLD LEG NODE (HORIZ PART OF COLD LEG)
            **                8--DOWCOMER NODE (IE DOW COMER PLUS LOWER HEAD) 14    0.          BR0RtN LOOP BREAK AkEA (Ffss2)
           -15     27.347     BROKEN LOOP BREAK ELEVATION--SEE NOTES ABOVE

16 VOLUMt IN A COOLANT LOOP (BOTH COLD LFGS FON FLANIS WITH ss IWO COLD LEGS PER HOT LEG) WHICH IS UNutR A HORI20NTAL 4

LINT DRAW THWOUGH THE BOTTOM 0F A COLD LEG N0ZiLL 17 MAX VOLUME OF WATtR IN ONE COLD LEG WHICH WILL STILL ALLOW sa GAS TRANSFER TO OCCUR FAST THE LOWEST FART OF The CULD LEG i 18 TOTAL VOLUME OF ONE COLD LEG 19 TOTAL VOLUMt 0F ONE HOT LEG i 20 TOTAL PLUID VOLUME OF (HE RX VESSEL IE IHE VOLUME NOT as INCLUDING THE CORE ITSELF OR INftRNAL STRUCTUNES ss21 GAS FLOW A1E OF REACTOR HIGH POINT VENT (S)eIF ANY, AT ss NOMINAL SYSTEM PRESSUkE i 4:DOWNCOMER IS M00tLLED AS kNDING AT IHE POINT WHERE IHE LOWER HEAD

            **0F THE RV MEETS THE CYLINURICAL SECTION--NOTE THE CORE 6ARREL IS
            **ALSO ASSUMtD TO STOP AT IHIS FOINT j            22                TOTAL VOLUME OF DUWNCOMtR 23                PORTION OF DOWNCOMER VOLUME WHICH IS BELOW IHE

__ ._ _ . _ . _ .~ - ._-. _ _ __. . . __._ __ _ _ __ _ __ _.___. A-17 ss ELEVATION OF THE 80TTOM OF THk COLD LEG N0ZZLES 24 3 sa kNTkR A 3 FOR PZR TO DE IN BROKEN LOOP 3 9 TO BE IN UNBR0 KEN

       **                 LOOP FOR U-TUNE GEONETRIESI USE 4 AND 10 REbPECTIVELY Fuk 25       4          8 AND W PLANTS (NODE NO. OF PRIMARY SYSTEM SURGE LINE N0Z)

NUMBER OF HOT LEGS 26 0 10 VOID FRACTION AT WHICH REACTOR C00LANf PUMPS TN1P OR Fall asSCRAM SEIPOINTS1 IF A GIVEN IRIP DOES NOT EXISie INPUT A VALUE WHICH THE asCODE WILL NkVER CROSS t 27 1900. LOW PRESSURIZER PRESSURE TRIP POINI 28- 12400. HIGH PRESSURIZER PRESSURE TRIP POINT 29 -381.7 HIGH LOOP DELTA-T SCRAM SETPOINI MINUS 459

       **                  SO 1 HAT MAAP GETS 78 F AFTER CONVER,10N 30    -100.        LOW PRESSURIZER LEVEL TRIP----THENE Ib NOME HEME 31    48 2         HIGH PRESSURIZER LEWL TRIP 32    5.556E-4 REACTOR TNIP DELAY TIMk

• 33 35.83 LOW S/G WATtR Lkvtl SCRAM SETPOINT 34 5 NUMsER OF P0!NIS IN MAIN C00LANI PUMP COAST-DOW LUNVE 88 (5 MAX) 35 3.55E7 N 3.23E7 hEONDFIRST L T MASS FLOWATg IN MCP COAST g0W CUNVk(M 37 2.48E7 NEXT FLOWRATE 38 1.77E7 NEXT FLOURATE 39 1 10E7 NEXT PLOWRATE 40 0.D0 FIRST TIMt IN COAST-DO W CURYL--MUST SL 0 -

41 2.778E-4 NEXT TIME b*. - 44 5.833E-3 XT TIME 45 33.13 ELEVATION OF 30TTOM OF S/G TUsESHkET A80VL BOTTOM OF RV

      **                  (IGNORED IM 3 AND W PLANTS) 46    .50          THICKNESS OF RV MkAD 47    26 20        ELEVATION OF INE SASE OF INE COOLANT LOOP N0Z4LES
      **                 (DISTANCE FROM 30TTOM 0F N0ZZLES TO BOTTOM OF RV NkAD) l 48      6.4        VERTICAL DISTANCE FROM LOWEST POINT OF A COLD LEG TO THE as                 ELEVATION OF THk 8ASE OF THE COLD LEG N0ZZLL ON THE RV 4?                 VULUME OF INE HORIZONIAL RUN OF A HOT LEG PIPE 50    0.0          TOTAL LETDO M FLOW--SEE NOTE MEAR MAKhur FLOW ENIRY ABOVE 51    35.5         NORMAL DIFFERoNTIAL PRESSUNE FROM CORE INLET TO HOT LtG as I                         SIDE OF OUTLEi N0ZZLES WHEN MAIN C00LANf PUMPS ARL ON sasassasssastALL THE NkMAIMDhR IN 1HIS SECTION ARE NkWassa asM0ST USERS WILL USE Tit 'UNDROKkN' LOOP SkEAK UNLY FOR PUMP SEAL LOCAS
     **IN IMLB SEQUENCE $l IT CAN ALSO BE USED FOR SPECIAL PURPOSES (EG LOFT FP/2

SIMULATION) asTHIS BREAKe ALONG WITH THt: BROKEN LOOP SREAK IS CONTROLLED Br EVtNT Col'E ss20YI ONE CAN TURN INE BNEAKS ON AND UrF SEPERAlhLY BY USING A PARAMETER .

ssCHANGE-TYPE INTERVENTION (CODE 1000---SEE V0L 1 0F USER'S MANUAL) 52 12 LOCATION KEY FOR UNSW0 KEN LOOP BREAK, IF ANY ss 9 --UNaR0 KEN HOT LEG N0DE

     **                  10--UNDROKkN HOT LEG 'luBC' NODE (3 AND W ONLY) l
     **                  12--UW8ROKEN INIENMkDIATE LEG N0DE--

a D C RN A HE P S 53 0. AREA 0F UNBROKEN LEG BRtAK--PUT IN ZERO IF NONE 3:54 35 ELEVATION OF UNBROKEN LOOP 3REAK (SEE NOTES FERTAINING

     **                        TO BREAK ELEVATION AD0VE                                                                                                                            ~

ssTHE ' DOME' REFERS TO THE REGION A90VL THk UFFEN PLENUM

     *siHE '00ME PLATE' IS THE PERF0NAIED PLATE fMAT DIVIDES IHE UPPER PLENUM ssFROM 55        THE DOMk--SEE DRAWINGS IN THE PRISYS SECTION OF THk USEN'S MANUAL                                                                                                     ~

ELkVATION OF THE RV DOME PLATE

                                                                                                                                                                                   +
                                                                                                       - , , - - - - - - -  w    ----     -
                                                                                                                                                           --- - - -- + . - --si.r
                                           -~,,w,m,,            -
                                                                         ----,--,------,------r-

l A-18 54 ELEVATION OF THE INSIDE OF THE RV HLAD 57 ELEVATION OF INE RV FLANGE (CLOSURE STUDS) as (NOTE THAT 'THIS EgVATION Ib 58 59 OU D AffA OY hME b Xft 0 F MASS OF THE CORE BARREL BELOW 1HE ELEVATION OF 1HE TOP OF 38 THE CORE (' LOWER CORE BARREL') FROM INr DATA + MAAP2 FILE E RE -- 2Z0 CN S 41 MASS OF UPPER PLENUM INTERNALS -- MAAP2 ZION CALC NOTES 62 MASS OF THE RV Dom PLATE -- MAAP2 ZIUN CALC NOTES 63 MASS OF 1HE WALL FORMING 1HE ExTERICR OF IHE DOME (IE as INCLUDES THE RV CLOSUNE NtAD) FNOM MAP 2 FILE 44 TOTAL MASS OF OhE HOT LE0tHOT INLET PLENUM WALL OF THE S/G+ sa THE TUl(SHkET MASS ASSOC WITH THk INLET PLENUM

• 65 TOTAL MASS OF ONE COLD LEO + COLD OUTLET PLENUM OF INE S/G ss FLUS THr. TUsESHtET MA5S ASSOCIAlkD WITH THE OUILEf PLENUM

       **                   NOTE: FOR PLANTS WI1H IWO COLD LEGS PtR OUILET PLENUM, as                   ADD ONLY HALF THE OUTLET PLENUM MSS--TML OlHkR HALF IS sa                    IHEN ASSOCIAftD WilH lHE UlHER COLD LES IN 1 HAT LUOP

., 66 MASS OF THL RV WALL (BELOW THE RV FLANGEi THE Dum WALL

       **                   ENitRED ABOVE STARTS AT IHE PLANGE) FROM MAP 2 FILE 67                   WATER LINE AREA IN THE UrPER PLENUM (ABOVE THE CQkE AND BELOW
       **                   1HE DOME PLA1E) -- ESTIMTE WIIM D=12Fre 1/2 WATER i       68                   HYDRAULIC DIAMETER IN THE UPPfR PLENUM -- INP
                                                                                      ~

69 TOTAL HEAT TRANSFtR AREA 0F INE UPPER PLENUM INTERNALS 70 CONVECTIVE (NON-RADIA1IVE) HkAT LOSSES UNDER NOM CONDITIONS , as FNOM S1EAM GENERATORS, PRESSURIZERe AND NEST OF PRIM. SYS. I as NOTE: DETAILED CALCULATIONS INDICATE THAT UNL(R NORML

       **                   OPtRATIONe IHE PRIMARY SYSTEM HEAT LOSS IS DUE VIRTUALLY ss                    ENTIRELY TO UNINSULATED PARTS OF THk SYSTEM (LOSS THkUUGH s

A Rb H T NAL PRIMRY SYSTEM HEAT LUSS (SEE IDCOR WEPORT 85-2 FOR DISCUSSION) 71 NO. OF PLATES IN PRIMARY SYSTEM REFLECTIVE INSULATION OR:

      **                    EN1ER 0 FUR CALCIUM SILICATE BULK INSULATION OR as                    ENTER -1 FOR ROCK WOOL INSULATION--IF YOU HAVL A
      **                    DIFFERENT lYPE OF INSULATION YOU SHOULD CONSIDER MODIFYING i

FUNCTION THCSUL WHICH SUPPLIES THE THtRMAL CONuuCTIVITY 72 TOTAL 1HICKNESS OF INSULATION 73 ELEVATION OF THE BASE OF THE CYLINDRICAL FANT OF THE RV 74 VOLUME OF IHE LOWER HEAD OF THE RV 75 TOTAL HEAT TRANSFER AREA 0F LOWkR CORE BARREL / THERMAL as SHILLUS (IE 1 HAT POR110N BLLOW INE TOP OF INE CUKE) l 76 TOTAL HtAT TRANSFER AREA 0F UrPER CUME BARREL , sa i as assstsatssssssssssssssssssssssssssssssssssssssssses**ssssssssssas***ss 8

• PRESSURIZER 488ssssssssssss**ssssssssssssssssssssssssss** sass ******s ssasssas***ss**

t L S SECTIONAL AREA 03 2235. PRESSURIZtR HEATER PRESSURE SETPOINT 04 2325. FRESSURIZER SPRAY PRESSUkE SEfPOINI 05 7.8 WATER LtVEL BELOW WHICH PZR HEATERS TRIP 06 6.14E6 PRESSUNIZLR MtATER TOTAL OUTFUT--IN MAAP THE HEATER:i as ARE EIIHER ALL ON OR ALL OFF 07 3.47E5 SPRAY SYSTEM FLOW RATE 08 4.2E5 FLOW RA1E OF SAFE 1Y VALVE AT ITS SETPOINT

A-19 09 2500. LOWkST SETF0INI 0F A SAFETY VALVE (0FENING FhES5UNE) 10 2$00. HIGHEST SCIPOINT OF A SAFE 1Y VALVE (OPtNING PRESSURE) 11 0.93 DIAMtTER OF THE SURGE LINE 12 46.25 ELEVATION OF SPRAY HEAD AB0VE BOTTOM OF FZR 13 64. LENGTH OF THE SURGE LINE

14. 3 NUMBER OF SAFELY VALVES 15 3.281E-3 NOMINAL FZR SFRAY DROPLET 16 2350. LOWEST SET FO!NT OF PORV (0PtNING PRESSURE) 17 2350. HIGHEST SET FOINT OF PORV (OFENING FREShuME) 18 2 NUMBER OF PORUS 19 2.1E5 NOMINAL FLOWWATE OF A PORV AT ITS SETF0INI 20 1 456E5 EMPlY MASS OF FZR STEEL i 21 0 ENTER A 1 IF THE SURGE LINT HAS A LOOP SEAL (EG TMI)I

                       **                        1HIS FREVENTS COUNTER-CURRENT DRAINING OF PNESSURIZER
                      **                        THN00GH SUNGE LINT WHEN THt FRIMANY C00LANI LOOP SIDE
                       **                        IS VOIDED (SEE WRilEUP FOR SUBWOUTINE DRAIN) 22           38.5          SEDIMENIATION ARLA                                                                          '
                      **sssssssssas4LL INE RtMAINDER IN IHIS SECTION ARE NEWasas asPRESSUNIZER RELIEFS ARE ASSUMED TO CLOSE AT FRES5UNE FSET-FDEAD WNEME EsPSET IS 1HE OPENING PNESSURE DEFINED ABOVE AND PDEAD IS GIVkN BELOW 23           100          DEADBAND ON FRESSUNIZER SAFETY VALVES 24          100           DEADBAND ON FRESSURIZF.R PORVS sa ss assassesss****sassas******sssssssssssssssss**sssssssssssssssssssssssssas

• STEAM GtNERATOR (VALUES NEFER TO ONE UNIT) assasssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssssa 01 5900. TOTAL SECONDARY SIDE FREE VOLUMEe EG OUT TO THE MSIV'S 02 DOWNCOMER CROSS-SECTIONAL FLOW AREA 03 fuse BUNDLE (SECONDARY SIDE) PLOW AREA 04 0 00 B AND W QNLY--ELEVATION OF AUX FEED SPRAY HEAD ABOVt

                     **                         1HE TOP OF 1HE LOWER IUBESHEET 05          1 158E6       INITIAL MASS IN CONDENSATE STORAGE TANK--OR A LARbE
                     **                        NU. IF NO LIMIT ON AFW SUPPLY 06                        2-FHASE WATER LEVtL IN TUsE BUNDLE AT THE SEC SIDE sa                         INVtNTORY SUFPLitD IN IHE

• INITIAL CONDITIONS SECTIONI Es THIS IS USED TO ADJUST THE VOID FNACTION DISTRIBUf!ON

                     **                         IN lHE IUBE BUNDLE SO AS TO APPROXIMATELY MAKE UP FOR s*                       SIMPLIFICATIONS IN THE MAAF MODEL I THE CDRRECTION 4:                        SHOULD MOST IMPACT LOSS OF FEED SEQUENCES 07           440.         MAIN FEEDWATER TEMPERATURE 08          1199.7        LOWEST SETPOINT OF SELONDARY SAFE 1Y VALVES 11           9.19E5 0   b Al b WR            G NOMINAL FLOWNATE OF A SAFETY VALVE AT THE SETF01NI 12          1135.        SETPOINT OF SEC RELIEF VLV (ASSUMtD SAME FOR ALL RELIEFS) as!F NO ' RELIEF VALVES *--SUPPLY A SET F01NI PRESSURE HIGHtN THAN THt s SAFETIES AND USE THE RELIEFS AS MANUALLY CONTMOLLED STEAM DUMPS 13           1            NUMBER OF RELIEF VALVES FER S/G 14           4 0E5        NOMINAL FLOWRATE OF A RELIEF VALVE 15           3.97E6      MAX FEEDWATER FLOWWATE FER S/G
                    ****sssssssssDEFINITION CHANGED FOR OTSG'Sasssstastasi

INCLUDE THAT FORTION OF THE TUkE VOLUME WHICH IS NOT COOLED (IE IS INSIDL salHE IUBESHEET(S)) IN IIEM 16 16 318 8 TOTAL (BOTH PLENA FOR OTSG'S) PRIMARY HtAD(5) VOLUMt--

asMAIN SIEAM ISOLATIUN VALVE (MSIV) CLOSURES MAIN FEEDWATER SHUT 0FFe asAND AUX FEEDWATER ACTUATION ARE ASSUMtD TO OCCUR AT REACTOR SCRAM sauMLESS DEFEAIED W!lH APPROPRIAIE kVtNT CUDES 17 .0028 TIMt DELAY FOR ACTIVATION OF AUX FELD AFTER SCRAM 18 .00138 TIME kEOUIRED FOR MSIVS TO LINEARLY RAMP FROM OFEN 1

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

A-20 l 88 TO CLOSED 19 944.1 TOTAL PRIMARY $!DE VOLUME OF ONE STEAM GENERATOR ' 20 1.78E5 MXIMUM AUX FEED FLOWRATE PER S/G l 21 100. AUX FEED T(MP(RATURE ' 22 5424 NUM8ER OF IUBES IN A STEM GENERATOR 23 0.0033 THICKNESS OF STEM GENEMATOR TusES 24 0.05067 ID OF STEAM UkNERATOR TURES 25 f.35 THERMAL CONouCTIVITY OF STEM GENtRATOR TusES 26 7.934E9 lHROTfLED FL0d PER STEM GENERATOR FOR AFW SYSTEM OR LARGE sa NUM8ER IF FLOW NOT THNOTTLED (SEE DISCUSSION AFIER ENGIN.

**                 SAFEGUARDS IfLM 151) 27       1.       FRACTIONAL AREA USED FOR STEAM DUMPS IN BROKEN LOOP S/G 28       1.        FRACTIONAL ARkA U5ED FOR STEM DUMPS IN UNSKM LOOPS S/GS                                     .

s* STEAM GENERATOR WATER LEVtL CONTROL (SbWLC) SYSTEM PARAMETERS! 29 34.8 00WCOMER PROGRAM WTR LEVEL FOR STEM 6tNERATOR WATER as LEVtL CONTNOL SYSTEM IN BROKkN LOUP S/G 30 34.8 DOWCMR PROG WTR LVL FOR S0WLC SYSTkN IN UNDEN LOOP S/GS 31 STEAM GENENATOR TUDESHEET DIMtTER ssFOR 34.8 ACCIDENT SIMULATION IT WAS MECESSARY TO INCORPORATE A BANG-BANG ssMODE OF S/G WATER LEVEL CONfROL--IE OPERATOR CONfkOLS THk WATER LEVEL 3 SIN A34.8CILLATORY WAY WIfMIN A DEADBANDI MOST USERS WILL NOT WISH

*sTO USE THIS MODE AND SHOULD LEAYt THE NEXl 1HNEL EMIRIES EQUAL TO 0 32     0.0           3-LOOP S0WLC DEADBAND 33     0.0           A-LOOP SGWLC DEADBAND (NONLERO VALUL ACTUATES ss                   BATCH FEED MODE) 34      0.0           FOR BANG-SANG MODE THE MINIMUM AFW FLOWATL PER S/G T0 as                   BE USED ON THE DECNEASING CYCLE sas***sts8ss:ALL THE REMINDER IN THIS SECTION Akt NkW8888 35     0             MAIN STEM LINE BREAKS CAN BE $1MULATED: ENTER 0 FOR NO

• • MAIN STEM LIk BREAKI 1 DIRECTS STEAM FNUM DNOKEN as LOOP S/G TO CONTMTl 2 DIRECTS STEM FROM ALL S/GS TO

!! Oit / AM S) A NO O ss THIS PARAMETER 36 TOTAL HEIGHT OF S/G SHELL AB0VE TUBESHEET 37 MASS OF $/G SHELL--DON'T INCLUDE MASSES ASSOCIATED WITH PRIMARY sa HLADS OR TUBESHEETS WHICH ARE LUMPED WIIM INE ASSOCIATED COLD as AND HOT LES MASSES IN THE PRIMARY SYSTEM SECTION 38 12 NUMBER OF PLATES IN REPLECTIVE INSULATION ON S/G SHkLLS OR

• • CODE INDICATING OTHER INSULATION TYPE (SEE FR1 MARY SYSitM

INPUT N0. 71) ss sassass**ssssssssssssssss88ss88 sas8sssssssssssssssssssssssss8 **stssssa i sTIMING DATA assassstas***8stsassass*8888838888888888388 333838388888838 338 statsssa 3:01 NOT USED 8:02 NOT USED 03 20.0 MAX TIMk STEP (ALWAYS INPUI IN SECONDS) 04 .005 MINIMUM TIME STEP (ALWAYS INPUT IN SECONDS) asTIME SELECTION ALGORITHMS ARE EXrLAINED IN THE WITL-UPS FOR SUBROUf!NtS f

• s!NTGRT (T/M MODELS) AND INTGFP (FISSION PRODUCT MODELS) 05 .05 RELAT!YE MASS CHANGE USED TO SELECT TIME STEP 06 .02 MINIMUM INTER-NODE FISSION PROD MASS TRANSPER CONSIDERED WHEN 6 33 PICKING TIME STEP IN FISSION PRODUCT MUDELS f 07 .02 RELATIVE GAS TEMPtRA10RE CHANGE USED TO SELECT TIME STEP 08 .1 REL MASS CHANGE FOR FISSION PRODUCTS USED TO SELECT TIME g STLP IN FISSION PRODUCT ROUTINES

A-21 sessssssssssss**sssssss:sssssssssssssssssssssssssssssssssssssssssssssans 400ENCH TANK ('OT* COMPT) sesssssssssssss**sssssssssssssssssssssssssssss***sssssssssss***stsstsess b 8N9 . ffTALWA l't A 03 105.7 FAILURE DIFFERENTIAL PRESSURE OF RUPTURE DISK 04 12 5 HEIGHI 0F RUPTUNE DISK A80VL SCOMPT FLOON 83 f AC THL 0 NCH A $ M L 88 sSI assssssssssssssssssssas**sssssssssssssssssssssssssssssssssssssssssssstas sMODLL PARAMETERS sassassssssssssssssssssssssssssss338833333333333333333333333338sssassss: ssSEE DISCUSSION IN VOL 1 0F USER'S MANUAL FOR ALLOWA3LE LIMITS ON

           *MODEL PARAMLTER VALUES AND THL DIFFERENI SENSITIVITY ANALYS15 MODkS sa as* SCALE FACTORS' MULTIPLY MODEL PREDICTIONS OF FLOWNATES ETC.-                                 .

s! ss **Ns HEY 0 L N bVE MEN CHMGEDassa 01 .005 CORIUM FRICTION COEFFICitNT FOR VESSEL ABLATION HEAT ss TRANSFER (REYNOLD'S ANALOGY) CALCS i 02 .002 LEAK-BEFORE-BREAK CONIMT LEAKAGE ARtA (IF 1HE CONTMT STRAIN as MODEL IS NOT USED, THIS IS THk AREA USED WHkN THL CUNfMr PRESS N 60. FA L N 1i ACT WITH CM

                                                                                                   ~

N fRYS ATfD AGkOS R E S I

          **                 CRIBER!ON IS REACHED AT PENETRATIONS)I UStB ONLY IF DETAILED sa                CONTAINMNT FAILUKE MODEL ACTIVATED 05     2.0          MULTIPLIER OF NORMAL CLAD SURFACE AREA USED IN OXIDATION
          **                CALCS TO ACCOUNr FOR STEAM INGRESS AFTER sa                CLAD RUP1UNE (MUST SE Bti HEN 1 AND 2) 06     983.D0      CRITICAL FLAML TEMP AT ZERO STEAM N0LE FNACTION ss                USED IF NO IGNITION SOURCESI 1HIS IS MULTIPLIED BY THE
          **                WESTINGHOUSE FLAME TEMPERATURE MULTIPLIER CURRELATION 83                FOR NONZERO STEAM M0LE FRACTIONS 07     1.          SCALE FACTOR FOR FISSION PRUDUCT ANU INERT AERO RkLEASE RATES N

09 300.0 850.

                                  -R D TIVE         hM   Ih. f. b         OE NAT. CIRC. (MCP'S OFF) S/G PRIMARY SIDE FILM RESISTANCE CMThPUOL i          **                WHEN 2- OR 1- PHASE NATUNAL CIRCULATION IS OCCURING 3:                  IN INE COOLANT LOOPS--NOTE IMAT COOLANT YkLOCI1Y AND sa                VOID FRACTION DISTRIBUlION ARE NOT COMrUILD UNDER THESE COND.

l- 10 .5 FRACTION OF S/G TU3ES CARRYING '00T' FLOWS IN 1HE HOT LEG

          **                NATUNAL CIRC MODEL (SEE SUsROU1!h HLNC WNITE-UP)I IF YOU as                WISH TO FORCE THE FLOW 0FFe USE O (1HIS REQUIRES BYPASSING sa                PARAMATER CR CKING BY USING THE SENSITIVITY ANAL OFTION as                 IBATCH=2)I PARAMETER DOES NOT AFFECT B AND W GEOMETRY UNLESS ss                0 IS INPUI SINCE 0756 TUM S DON'T PARTILIPA1E IN FLOW 11        .1      s AND W ONLYi FRACTION OF $/G TU8ES STRUCK BY APW 12      1.D3      HT. TRANSFER COEFF BETWLEN MOLTEN CUNIUM AND A FROZEN CRUSTI as                USED IN DECOMP AND IN CALCULATIONS WIIMIN A MOLTEN POOL IN 8                 THE CORE 13      0         ENTER A 0 FOR ENTRAINMENT FROM C TO Bf 1 FOR C TO A--LATTER 8:                GENERALLY USED ONLY IF CAVI1Y HAS NO INSTRUMENT IUNNEL 14      0.D0      IF 13 IS NONZLR0s FRACTION OF THE ENfRAINtD MASS WHICH I          as                $1RIKES IHE MISSLE SHIELD (BEFORE SIGNIFICANILY INTERACTING ss                WITH THE UPPER COMPARTMtNT 6AS) l l

A-22 15 1.00 DRAG COEFFICIENT OF RISING PLUME DURING BURNS IN UPPER 88 COMPT-- LARGER VALutS RESULT IN A SLOWER AND FATTER PLunt 88 AND IMUS INCREASE 1HE EFFICIENCY OF INE IGNITkRS kh . 18 1.00 SAME FOR D COMPT 19 1.00 SAME FOR U COMPT 20 1.53 CHURN-TUNSULENI CRITIC E VtLOCITY COEFFICIENT 21 3.7 DROPLtf FLOW CRITICAL VkLOC11Y COEFFICIENT 22 1. SPAR 6ED POOL VOID FRACTION COEFFICIENT

                                     .5 5
                                                .9 E RI     N EMIS5IVI1Y OF WATER T     N      OF 26        .8S    EMISSIVITY OF WALLS 27         .85   EMISSIVI1Y OF EQUIPMENT 28        .85    EMISSIVITY OF CORIUM SUNFACE 29         .6    EMISSIVITY OF GAS 30        .3     CORE MfDR0 DYNAMIC LIMIT KUIATELAD4E NO. FOR RLFLOODING M1
                                     **               AND OXIDATION CALCOLATIONS 31     .33D0     NUMBER TO MULTIPLY KulATELADZE CR1TERION ST TO REPRESEN1
                                     **               DIFFICULTY (GT 1.D0) UR EASE (LT 1.D0) FOR DEBRIS TO GET ss               OUT OF CAVITY 32      3.0      t'LOODING CRITICAL VELOCI1Y C0 EFFICIENT 33       .14     FLAT PLATE CHF CRITICAL VtLOCITY COEFFILIENT 1.
                                                                              ~

34 NUMBER OF VESSEL PENEiRATIONS 'lHAT FAIL 35 .75 DISCHARGE COEFFICIENr FOR PRIMANY SYSTEM BREAK (b)

                                     *** SCALE FACTORS' MULTIPLY MODEL PREDICTIONS--lHE BEST-ESTIMATE VALUE
                                     **IS USUALLY 1 36     1.00      SCALE FACTOR FOR BUNN VELOCITY CORRELATION 37     1.00      SCALE FACTOR FOR HEAT TRANSFER COEFFICIENTS TO PASSIVE
                                     **               HEAT SINKS 38      2.500    GAMMA SHAPE FACTOR (TO ACCOUNI FOR NON-SFHERICAL SHAPES IN st               1HE COA 00LATION EDUATION) UstD FOR AEE050LS 39      1.D0     CHI SHAPE FACTOR (TO ACCOUNI FOR NON-5FHtkICAL SHAPES IN
                                     **               STOKES LAW) USED FOR AEROSOLS 40        3.D0   RATIO 0F AIRBORNE AEROSOL MASS TO THE MASS WHICH WOULD LEAVL
                                     **              -YOU IN STEADY-STATE WI1H 1HE CURRtNT SOURCE STRLNGlHilHIS IS
                                     **               USED TO CONIN0L THE SELECTION OF DELAY VS SitADY-STAfE AtKOSOL
                                     **               SET 1 LING CORRELATIONS 41       10.D0    DEC0NIAMINATION FACTOR ASSOCIATED WITH THt FAbSAGE THkOUGH 1
                                     **               MtTER (REFtRENCE LLNG1H USED FOR E!1HER SET OF UNITS) 0F WATERI 48               ASSUME DF IS LINEAR FUNCTION OF DtPTH FOR OTHtR DEPTHS 42       .02     CAPIURE EFFICIhNCY OF CONIMT SFRAY FOR AER0 DOLS--lHIS IS
                                     **               THE FRACTION 0F THE TOTAL VOLunt SWtFT ff FALLING Dr.0FS WHICH
                                     **               IS CLEANSED OF AEROSOLS 43     1.D0      ASSOLufE VALUE OF THE DLSIRED MULTIPLIER 0F Cb! ANU
                                     **               CSON VAPOR PHESSURE--ENTLR A NEGATIVE NUMBER TO StLECT
                                     **               JANAF CSON FUNCTIONI F05 FOR SANo1A CURLLLATION (BEST-EST) 44     .1D0      FRACTION OF CLAD OXIDIZED WHICH CAUSES LOFE TO COLLAFSE ON
                                     **               REFLOOD (GIVES SMALLER KU FOR HLAl' 1RANSFER THAN INIALI
                                     **               MODLL) AND CAUSES CORE GEOMEIRY TO CHANGE 45     0.00      FOR S AND W UNITS ONLYe FRACTION OF FtRFtCT CUNUENSA1 ION
                                     **               OF STEAM LN1ERING DOWNCOMtR IHROUGH PLAFFER VALVES
                                     *SI 46     2100. TEMPERATUkE AT WHICH CLAD FAILS IF IT HASN'T ALREADY RUtiukEDI
                                     **                IMIS HALIS PUR1HER BALLOONING AND ALLOWS fib 5 FROD RtLEASE 47     2.M5      LATENI HEAT OF U-IR-ZR02 EUIECTIC 48       .25     VUID FRACTION OF A COLLAFSED CORE

A-23 49 .3D-6 SEED RADIUS ASSUMED FOR HfGROSCOPIC AEROSOL GROWiH CALC 50 -2 EN1ER A 2 FOR FISSION PRODUCT kELEASE TO BE COMPUTED ss SY THE IDCOR/LPRI STEAM OXIDATION MODELI 1 FOR

      **                    NUREG-0772 MODELI NEGATIVE N05. ACTIVATE INE SAME MODEL
     **                    AS POSITIVE NUMutRS SUT ALSU TUhN ON A BLOCKAGE MODEL
      **                   WHICH REDUCES INE RELEASE OF NONVOLATILE FISSION PRODS 88                    WHEN THE N0DE IS BLOCRED FOR GAS TNANSFURI 31          0         ENTER A 1 IF TELLURIUM IS RELEASED IN-VESStLi 0 IF IT
     **                    IS ASSUMED TO BE TOTALLY BOUWu Ur WITH 2IRCALLOY 88                     (0 IS 8EST-EST) hh                    h!f kb                   A
     **                   CALhSifMIkOkhh0RAXf!MATEDBYFs2.sDP8KH CAN BE ES VALUES ARE FOR NORMAL OPERATION WI1H MCP'S ON) DP.C0KE sa                   PRESSUNE DROFe RHO

• DENSITY OF FRIMARY SYSTEM C00LANie Ga

     **                    CURE AVERAGE MASS FLOW PtR UNIT AREA (IN BRIT UNITS,
     **                    INCLUDE G0 ANU OTHER NtCESSARY CONVtRSIDNS TO MAKt F
     **                    DIMENSIONLES$)--USE GT 100 TO ARTIFICIALLY STOP PLOW
     **                    (REQUIRES USING THE SENSTIVITY OPTION IPATCHs2) 54          0.        INSERT 0 IF IN-VESSEL NAIURAL CIRCULATION PLOW RElURN LEG 88                   IS IN DUTER FUEL ASSEMsLIES (USUAL CASE)flNSERT 1 IF RETUKN
     **                    IS DOWN ' BYPASS' (IE BAFFLE-CUKE BARREL ANNULUS)--lHIS UQULD ss                   BE EXFECTED ONLY IF THERE WAS A LOT OF FLOW AhEA IN THt
     **                   BYPA$$ (EG FERHAPS B AND W PLANTS) 55             .I     A VOID FRACTION: BELOW WHICH A CORE NODE IS AbbOMED BLOCKED
     **                   FOR GAS PLOW OR OXIDATION                                          .

56 10. NO. OF SAMrLES AVtRAGED OVtk IN NC MODEL (SEE UhER'S MANUAL) 57 .25 CROSS-FLOW PRICTION COEF IN NC MODEL (LIBERA 1 UKE SAYS .25 .45) 58 .05 FRACTION OF XENON INVENTORY IN THL FtLLET-CLAD GAP DUE TO 4

     **                   LONG-TERM OPERATION (OFTEN CALLED IHE ' GAP RELEASE *, 1HIS
    **                    IS USED IN CALCULATING THE FREShuRE INSIDE THE FUEL FIN FOR
     **                   8ALLOONING CALCS--NUMEG 0772 SAYS OBSERVED VALUES ARE 0-0.25) 59             .35    VOID FRACTION IN FRIMARY SYSTEM AWOVE WHICH THt FMASES
    **                    SEPARA1E AND 1WO-PHASE NA10RAL CIRCULATION STUFS 60             1060. TEMPERATUNE OF H2 JEl ENIERING NON-INtRTED COMPARTMtNI WHICH
    **                    IS SUFFICIENT TO CAUSE A LOCAL SURN--FRUM HEDL-BME 78-50 ts
    ***sss**********************sastas****sasssas48****ss**ssssssssssassanas

• FISSION PRODUCTS

    ************888**sts*****st*****************************s*ss*******ssss:
    *sts********sas*****NEWas**** stats ***s**:
    ** FISSION PRODUCT GROUPING SCHEME:
    ** GROUP 1: NOBLE GASSES AND 'INLRT' (NON-RADIDACTIVt) AEr.0SOLS s* GROUP 2: CSI

• ** GROUP 3: TELLURIUM (TAKEN TO BE ELEMENfAL TE)

   ** GROUP 4: STRONT!UM (TAKEN TO BE SRO WI1H BARIUM LUMFED IN AS BAO)
   ** GROUP 5: MOLYBDENUM (TAAEN TO BE M003)
   ** GROUP 6: CSOH ss

( ** STRUCTURAL MATERIAL GROUr!NG SCHEMt

   **USED IN CORE N0 DES (TRACKED IN CONfAINMtNT AS LUMrED GROUP 1 AEROSOLS)
   ** GROUP 1: CD
   ** GROUP 2: IN
   ** GROUP 3: AG ssGR00P 4: SN
   ** GROUP 5: MN 88
   *$$$$tttst*Satss8888NEWilstt$sttSasSE***:

1

A-24 01 .0428 FRACTION OF FISSION PRODUCT POWER IN GROU7 1 02 .222 SAME FOR GROUP 2 03 .0467 SAME FOR GROUP 3

           .                      UPh 06       .0415     SAM FOR GROUP 6 07       433. INITIAL MASS OF FISSION FRODUCTS IN GROUP 1 (NOBLES ONLY) 08       23.49    INITIAL MASS IN GROUP 2 09       40.92    GROUP 3 10       181.5    GROUP 4 (TOTAL KG OF SR AN9 BA EXFRESEED AS OXIDES) 11       552.4    GROUP 5 (EXPRESS AS KG OF TRI-0XIDE) 12       250.0    GROUP 6 13       116 8    INITIAL MASS OF CD IN CCRE (STRUC MATERIAL GROUr 1) 14       305.4    INITIAL MASS OF IN IN CORE 15       1868 6 INITIAL MASS OF A6 IN CORE 16       308 6    INITIAL MASS OF SN IN CORE                                -

17 6.68 INITIAL MASS OF MN IN CORE 8818 NOT USED sssssssssssssss***ssss**********ss****sassatsssssssssssssJss************* sAUXILIARY BLUILDING

  **sassassas**********stssssas*********s****s***ssassassassssssssssssssssa
  **CAN MODEL A MAXIMUM OF 5 SERI ALLY-LONNECTED H0 DES stT OSE RECEIVE FLOW FROM THE CONfAINMENI FAILUME AND, IN V-SEGULNCES, asFRm THE PRIMARY SYSTEM BREAK (S) ssYOU NEED SUPFLY INFORMATION ONLY FOR THE NO. OF NODES YOU SELECTED IN
  **THE 8 CONTROL SECTION--NOIE 1 Hale FOR EXAMPLEe 1HE Mass 0F WATER INITIALLY IN NODE 1 ALWAYS GOES IN INFUf NO. 6 NO MATTER HOW MANf 88 MODES YOU ANE USING 01      421 4     VOLRB(I) VOLUME OF NODE 1 02      711.8     NODE 2 03     2638.2     NODE 3 04      1380.2- NODE 4                              3
 **05        0. NODE 5 ssMASS OF WATER CAN BE USED TO REFRESENT, FOR EXAMrLE: REFUELING FOOLS OI     $$        !$RhI) MkSh0FWATERINFDDE1                                     -

07 0.0 NODE 2 08 0.0 NODE 3 09 0.0 NODE 4

 **10 0.0         NODE 5 11     0.0       AWATRB(I) SUNFACE AREA 0F WATER FOOLe IF ANf IN NODE 1 12     0.0       NODE 2 13     0.0       NODE 3 14     0.0       NODE 4
 **15 0.0         NODE 5 asAT FRESENT ONLY ONt EXIERIOR WALL PER NODE IS MODELED IN THE AUX CODEI 881E THE WALL HAS AUX CONDITIONS ON ONE SIDE AND tHE ENVIRONMENT ON 1HE ss0THER--EITHER A STEEL OR CONCRETE WALL CAN BE MODELED BY INFUITING THE 88 APPROPRIATE MATERIAL PROPERTIES 16     211.8     ANSRB(I) ONE-SIDED WALL AREA FOR NODE 1 17     357.8     N0DE 2 18     1266 6    NODE 3 19     710.2     NODE 4
 **20       0. NODE 5

((

 ..     .457{
        .457.

XHSRB(!) THICKNtSS FOR NODE 1 WALL NODE 2 23 .4572 NODE 3 24 .4572 NODE 4 4 as25 0. NODE 5

A-25 26 1.59 KNSR$(I) TNtRMAL CONDUCTIVITY OF WALL IN NODE 1 27 1.59 N0DE 2 28 1.59 NODE 3 29 1.59 NODE 4

 **30      0. NODE 5 31     656.7    CFNSRB(I) SFECIFIC NtAT OF WALL IN NODE 1 32     654.7   NODE 2 33    656.7    NODE 3 34     654.7   NODE 4
  **35 0.        NODE 5 34 2.65        ZNSR8(I) NEIGHi 0F WALL (FOR NC CALCS) FOR NODE 1 37 4.48        NODE 2 38 16 61       NODE 3 39 8.69        NODE 4
  **40    0. NODE 5 41              DNSRB(I) DENSITY OF WALL IN NODE 1 42   3'308.5 308.5   NODE 2                                                                              -

43 2308.5 NODE 3 44 2308.5 NODE 4

 **45     O. NODE 5
 **T,wE VENTILATION (OR 'SGTS') SYSTEM IS NODELED Bf SUFFLTING
 **A FORCED OUT > LOW AND/ux A F0NCED IN FLOW
 **TNIS FLOW IS ON UNf!L tnt FIRE DAMFER SETPUINt(SEE ftLGW) IS
  **REE%D IN A COMPARIMkNT--lHIS SHUTS FLO'J DOWN IN IHAT COMPT
 ** NOTE THE AC POWER EVtNr CODE DOES NOT AFFECT THE SbiS FLOWI TO
 **SNUT 1NE FLOW OFFS SUPPLY O'S NELOW OR A LOW FIRE DAMPER TtMP 44 0.0          WVOR8(I) FORCED VOLUnt RIC (M**3/SEC OR GFM)
  **             VENTILATION FLUW OUT OF NODE 1 47 0.0          NODE 2 48 00           NODE 3 49 22.26       NODE 4
 **50     0.0    NODE 5 51   17.96     WVIRB(I) FORCED VOLUMETRIC VtNIILAllbN FLOW INIO NODL 1 52   0.0        NODE 2 53   4.30      NODE 3 54   0.0       NODE 4
 **55     0.0   NODE 5 54   200.6     ASEDRB(I) AEROSOL SETTLIhG AREA FOR NODE 1 57   200.6     NODE 2 58  200.6      NODE 3 59   200.6     NODE 4
 **60 4900. NODE 5
 ** AEROSOL IMPACTION DATA
 **SEE DISCUSSION OF IMrACTION FARAMETERS IN

• ANNULAR SE T!bN AbOVL

 **IF IMPACTION IS MODELhD IN A NODES INE IMF ACTION AEEA, DIAMETER (EG GRATE
 **TNICKNESS)e AND FLOW AAEA MUST ALL EL GIVth 61 0.          AIMPRB(I) IMPACTION AREA FOR NODE 1 62 0.          MODE 2 63 0.          NODE 3 64 0.          NODE 4
 **45     0. NODE 5 64    0.       XDIMMB(I) IMFACTION DIAMtTER FOR NODE 1 67 0.0         NODE 2 68 0.0         NODE 3 69 0.0         NODF 4
 **70 0.0       N0DE 5 71 0.          AGRARB(I) GRATE FLOW AREA FOR NODE 1 72 0.0         NOUE 2 73 0.0         NODE 3 74 0.0         NODE 4

A-26

                   **75 0.0        NODE 5
                   ** SPRAYS (EG FIRE SPRAYS)--THtSE ARE TURNED ON ANu 0FF USING EVENS
                    ** CODE 240 MANUALLY--NO AU7CMATIC INITIATION j6 O0  g        gggBI) SPI 4r NASS FLOW RATE FOR NODE 1 78 0.0          NODE 3 79 0.0          NODE 4
                   **SO 0.0        NODE 5 61   0.0        XHSPRB(I) SFRAY FALL HEIGHf FOR NODE 1 82   0.0        N0DE 2 83   0.0        NODE 3 84   0.0        NODE 4
                   **85    0.0     NODE 5
                   **INITAL CONDITION DATA 86    305.      INITIAL TEMPERATURE OF AUX BUILDING 87    305.      AUX BLDG SPNAY WATER TLMP                                                             '

88 1.D-3 AUX BLDG SPRAY DkOF DIAntTER 89 .5 INITIAL Rtl HUMID!lY OF AUX BUILDING COMPTS 90 311. ENVIRONMENs TEMr 91 1.D5 ENVIRONMENT / AUX BLDG FRESSURE 92 355. FIRE DAMFER ACTIVATION TEMr (FUf IN 0 TO Shut DUWN SGTS (SEE

                   **              A80VE)I PUT IN A VERY HIGH NO. IF NO FIRE DAMrENS)
                   **IF DESIRE TABULAR OUIPut ANU OFERATOR INiERVtNIIUNS IN BNITIhH UNITS
                   ** INSERT A *BR HERE (EG FOR A PARAMEIER DUMP IN BKITISHe OPERATOR
                   **INTERVEN f!ONS: ANU TABULAR OUfFUI)
                   *SI l
 . _ _ . - - _ . _                  .--        ., - --       -     ,-- - - - - - . . . - - . ~ -

a sur. -- -a. a x .-_ _+- ~,. i B-1 APPENDIX B Steam Generator Tube Integrity Analysis B.1 Introduction In the previous sections, the analyses show that the steam generator tubes can be subjected to temperature and pressure conditions beyond the design basis for certain severe accident sequences. After core uncovery, natural circulation in the primary system can bring hot steam and hydrogen , into the steam generator inlet plenum and the tubes. Heatup of the tubes is retarded by heat transfer to the secondary side, so gas temperatures will exceed the tube temperatures. At this time, primary system pressure could ! be at the pressurizer safety or relief setpoints, while the secondary side . could be at the steam generator safety or relief points. The combination of high pressures and potentially high tube wall temperatures thus raises the possibility of tube failure and containment bypass. The purpose of this section is to provide information related to the strength of the steam generator tubes at these conditions, from which conclusions related to their integrity can be drawn. B.2 Tube Degradation During the course of operation, tubes may experience some degradation, i and the effect of such degradation must be considered in evaluation of tube integrity. On the basis of operating experience, tube degradation may be ! separated into two distinct categories: 1) denting and 2) thinning and/or cracking. The effect of these two types of defects on tube integrity are discussed in the following sections, in light of the predicted severe accident loadings. Numerous studies have reported on the various modes of tube degradation and the burst strength of steam generator tubing. The results discussed below are based on test data reported in Reference (B-1). All of the tests were conducted in a simulated steam generator environment at 600 degrees F i l i i _ _ , _ . . _ . _ _ _ . _ _ . , _ , ~ , _ . . _ _ _ _ _ , _ _ _ _ _ . _ _ _ _ . _ _ _ , . , . _ _ . _ . . . _ _ _ _ _ . _ _ . , _ _ _ _ _ _ , _ _ __ _ _ ._ _. . . . . .

B-2 for 0.875 inch diameter and 0.05 inch wall thickness Inconel-600 mill-annealed tubing. l B.2.1 Tube Properties ' The following sections sumarize results of tests conducted with steam generator tubes in research facilities to determine the burst strength of new and degraded tube at a temperature of 600 degrees F whereas the pre-dicted wall temperature for severe accidents may be substantially higher. In this section, a method of relating these data to the accident case is ' presented. Theories on ductile failure indicate that the flow stress is approximately proportional to the sum of the material yield and ultimate strength (B-2). The strength properties of mill-annealed Inconel-600, at elevated temperatures, taken from Reference (B-3), are shown in Figure B-1 i and are summarized below for a few selected temperatures. Temperature Yield Ultimate Flow Stress Flow Stress (F) (ksi) (ksi) (ksi) Ratio 600 43 94 137 - 1000 41 82 123 0.90 1350 30 47 77 0.56 The correlation of material properties shows that the expected burst strengths at 1000 degrees F would be 90% of those reported in the tests; the expected burst strengths at 1350 degrees F would be 56% of those reported in the tests. The reported data at 600 degrees F can therefore be used and adjusted for the predicted higher severe accident temperatures. B.2.2 Thinning / Cracking Type Defects In the case of thinning the defects were produced by machining uniform-ly on the tube outside diameter (00). Combinations of three different penetration depths and four defect lengths were tested, in addition to the l undefected virgin specimens. Results of these tests are shown in Figures B-2a and B-2b. As indicated by these results for a tube with a 1.5 inch

8-3 100 , , , , , g % ,% ~ } T'l,'% St ~ 90 = "gth - i V 80 - . c 9 l70 = [ l 5 w i i I 60 - ' - l k50 - (Ybh, f' n - , m.. e t) , -- - ----. . . . . .-  !

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l

  - a0                                     -                              ,          -
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30

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                                                           .l -- _.

20 - i - i 10 - i e i I 0 I I I I I I I i 0 200 400 600 800 1000 1200 1400 1600 1800 Temperature. 'F i Figure B-1 High temperature tensile properties of annealed (1600 F/1 hr.) hot-rolled plate (B-1). ,

B-4

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e m g # \ E - 0 oo W. 7

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  .                                        e 53 e
        ,            o UNDEFECTED                                                                                       ,

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2 = O $5 60*.

                     , 7 5 8 0 *'.

1 - 0 ' ' ' ' ' ' - 0 0 25 0 50 0 75 5 00 1 25 t to OEFECT LENGTM IINCMESI Figure B-2a Burst data for 0.875 x 0.050 uniform thinning specimens - defect length variation (8-1). 10 - O  ; e > s - I 7 , m , E [. - e  % 3 ' l' g4 -

                                                                                          =

Ia - o UNDEFECit0 y a 2 - e 3e te* TH NNEO tt%GfM 0 3, e TMIN es t o 6 t 'e G tM N e 3 / 4

• TMia.NI O Li a.G t M t , a 11/2" TMihNEO LENGTM 0 ' ' ' ' ' '

O 10 20 30 40 90 to 70 80 90 100 M AXIMuM DEGR AD Afs0N ia. W ALu Figure B-2b Burst data for 0.875 x 0.0F's uniform thinning specimens - defect depth variation (B-1). l l l l

B-5 long defect, 50 percent of nominal wall thickness is required to withstand the postulated event loadings when the degradation in flow stress at high temperatures is considered. That is, 5 ksi

• 0.56 = 2.8 ksi which exceeds any possible aP across the tubes, given the rather high 1350'F temperature.

Considering a aP of 1250 psi present in the reactor, a good margin of safety exists. For shorter and/or nonunifom defects, the allowable degradation ! would be somewhat larger than 50 percent. To study the effect of tight cracks the defects were simulated mechan-ically by EDM slots. Slot lengths of 0.25 inch. 0.5 inch, and 1.5 inch and slot depths in the range of 25-30 percent, 55-60 percent, and 85-90 percent

• of nominal wall were tested. The results of these tests are shown in Figures B-3a and B-3b. Again, from these results, it is seen that a tube with a 1.5 inch long and 50 percent through-wall crack would be able to withstand the postulated event loading without rupturing.

B.2.3 Tube Denting Because of the inherent mechanism of denting, namely the corrosion of carbon steel tube support plates, the dented region of tubes is confined within the support plate thickness. The denting was therefore simulated by hot-swagging a carbon steel ring onto the tube. The ring thickness was 0.75 inch, simulating the thickness of the support plate. The nominal denting depth of 0.04 and 0.05 inch. To investigate the effects of tube thinning near a dent, tests were also perfomed on specimens with denting plus uniform thinning and denting plus elliptical wastage defects. Results of burst pressure tests with various denting and thinning configurations are summarized in Figure B-4 and Table B-1. These results clearly indicate that denting has no degradation effect on the burst strength of a tube with either the nominal or wasted wall. A similar t conclusion has been derived regarding the strength of a dented tube with superimposed cracks (B-5). Thus, the observations made previously regarding

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            ,4      .
            .Is                         O UNDEFECTED 2  -                    0 1/4' SLOT                                                            ,

c 1/2" SLOT g , a 11/2" SLOT 0 t ' ' ' ' ' ' ' ' 0 to 20 30 40 50 60 10 80 90 twJ MAXIMUM DE GR AD ATION I'e W ALL; Figure B-3a Burst data for 0.875 x 0.050 EDM slot specimens - defect depth variation (8-1). n..... ie .

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0 0 26 0 80 0 76 1 00 1 25 i se of f ECT LENGfM tehCMiss ! Figure B-3b Burst data for 0.875 x 0.050 EDM slot specimens - defect 1&ngth variation (8-1).

B-7 a, <t W

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Figure B-4 Burst pressure data of 0.875 x 0.050 uniform thinning specimens with and without denting (B-1).

B-8 Table B-1 COMPARISON OF BURST PRESSURES OF ELLIPTICALLY WASTED j

                                      .875 OD x .050 WALL TUBING WITH AND WITHOUT DENTING (B-1)                                                                                                                                !

l Radius of Wrap Burst Pressures (psi) Cutter Angle Depth (in.) (degrees) (% Wall) Without Denting With Denting 24 0 25-30 8155, 8150 7905, 8150 . 24 45 25-30 7615, 7940 9200, 8060 24 135 25-30 5200, 7780 7800, 7770 12 0 55-60 5550, 5635 5520, 5605 i 12 45 55-60 5680, 5840 5650, 5675 i 12 135 55-60 5610, 5455 5320, 5230 I .I l l

B-9 the required wall thickness to withstand the event loads are unaffected by  ; the presence of denting. B.3 Creep Rupture The potential for creep rupture must be evaluated given the possible combination of high temperatures and pressures. Pressure differences of 1230 to 1300 psi can exist across the tubes when the primary and secondary , sides are both at the safety setpoint pressures. A pressure difference of 1215 psi can exist if both sides are at the respective PORY setpoints. Given the tube dimensions, this translates to nominal hoop stresses between . 9330 and 9980 psi. Creep rupture times for various tube temperatures can be estimated given these stresses. i Creep data for Inconel is presented in Figure B-5 and sumarized below. At temperatures of Il50*F and 1350*F, corresponding roughly to 900*K and - l 1000'K respectively, rupture times are seen to be on the order of a thousand hours or greater for the stresses considered. Rupture Stress (ksi) r Temperature 1000 hr. 10,000 hr.

                                   'F,(*K)                                      Rupture                                                                                     Rupture 1150 (894)                                       = 15                                                                                               = 12 1350(1006)                                                     9                                                                                                  7 Additional creep rupture data for Inconel-600 from Ref. (B-8) is sumarized in Table B-2. Data are provided here for shorter rupture times, corresponding to higher temperatures or stresses. The 1000 hour rupture data are consistent with that of Figure B-5. According to this table, a temperature of 1500*F (1089'K) is required for a 10 hour period to rupture tubes with a 10 ksi stress. Thus, for the accidents considered, creep                                                                                                                                          '

( rupture is not a credible phenomena. i

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.,       800    6000 1200 1400 i600 1800                              800 1000 1200 4400 1600 iS00 Testing temperature, F                                               Testing temperature, F 40                                                                                                                           i l                                                                                                                      01% Crees 304                       \                                                                                             " '          

type 316 steintess ,' l d 10 i

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                                                                                     /                              3o 446                                                     i l

700 800 900 1000 l100 6200 i300 1400 l*CO #600 .70 0 5: Testing temoerot.re. 8' l Figure B-5 Creep and creep-rupture comparisons (8-4). I e na- - , - - , - - , - _ , , , ,,_--.-,.-.---,--,,-_.,_,,,,y------._m.m,

B-ll Table B-2 CREEP RUPTURE DATA FOR INCONEL-600, HOT ROLLED, AND ANNEALED AT 1600*F Rupture Stress (ksi) Temperature 10 hr. 100 hr. 1000 hr.

                                                      'F,(*K)                 Rupture                             Rupture           Rupture 1100 (866)                           48                               35                22                               ,

1200 (922) 33 22 14 1300 (977) 22 14 8.6 1400 (1033) 13 8.6 5.8 1500(1089) 10 6.5 3.9 4 i t 1 i l l 1

  - ,o,-- ~ - - r..-nn-,v,- - - - - - - - - - ,               ,.e---    ,,y,,        -y,-7  ,- r y,--v-,w,-wn~-                             - , , - - - -war---nnw-w,,,-

l B-12 l l Prediction of the failure time given a temperature and stress, for failure times below 1000 hours, is possible by correlating a parameter combining time and temperature with the stress. The Larson-Miller parameter is such a parameter [Ref. (B-6)]. For 316 stainless steel, it has the fom LMP = T*(20 + log t )

• 10-3, where T is in temperature in degrees Rankine, r

t ris the rupture time in hours, and log represents the base 10 logarithm. A master rupture curve for 316 stainless steel is shown in Figure B-6, as presented in Ref. (B-7). The Inconel data from Figure 8-5 maps onto the j same curve as the ASME N-47 data for stainless steel, and from Figure B-5 it can be seen that these two materials exhibit very similar creep rupture , behavior. Thus, for example, assuming stress of 10 ksi, a Larson-Miller parameter value between 40 and 41.5 is obtained. Given a temperature of ! 1500*F = 1960'R = 1090*K, a rupture time between 2.5 and 14.9 hours is obtained. Checking the validity of this extrapolation, at 10 ksi and thus the same Larson-Miller parameter range, for tp = 1000 hours a temperature i between 1280*F and 1340*F is predicted. This agrees quite well with the i value 1325'F from Figure B-5 and an interpolated value of 1275'F from Table B-2. Therefore, the extrapolation approach is seen to be valid, and temper-atures in excess of 1500*F would be required to fail tubes by creep rupture, ! since the time at elevated temperatures is well under an hour. i i B.4 Sumary i i Based on the material properties of Inconel-600 at elevated tempera-tures and the results of the test program on burst strength of steam genera-l tor tubes, it can be shown that the combined pressures and temperatures at which the tubes can continue to maintain integrity are within the predicted conditions for severe accidents described elsewhere in this report. The results are applicable to conditions of tube degradation up to 50 percent 3 unifom through-wall wastage as well as for crack profiles which represent the limits of operation.

Creep rupture is not considered to be a failure mechanism for steam 3 generator tubes at the pressure, temperature and time conditions predicted

! for severe accidents. This conclusion is based on three independent data i i

Master Rupture Curve for 316 Stainless Steel 10 0 - i i i i i i

                                          ~~ '

n -n.xa4 o ,

                                                        / '-       , 'gpx ASME Code C:-e N-47 i

g m Larson-Miller data / f or 18Cr- 8Hi S.S

• Q'a 0 ~u,,

N j 0 lI 700 K (800 'F) a ' fs X 755 K (900 'F) 's O 0 - 811 K (1000 'Il .'[ L j () - 922 K (1200 'F) A - 1033 K (1400 'F) y V - 1089 K (1500 'F) e vi

               ). _ __ _          i     ...___.!..___        l____....      _1_____...        I             __1_.           _. ...

15 20 25 30 35 40 45 50 Larson-Miller parameter [T(20+1ogtr) x 10-a Figure B-6 Master creep rupture curve for 316 stainless steel, taken from Ref. (B-7). 9

B-14 sources, which indicate that temperatures in excess of 1500*F must be sustained for long periods for creep rupture. B.5 References (B-1) M. Vagins, et al., " Steam Generator Tube Integrity Program - Phase I Report". NUREG/CR-0718, PNL-2937 (September,1979). (B-2) R. J. Eiber, et al., " Investigation of the Initiation and Extent of Ductile Pipe Rupture (Final Report)", BMI-1908 (June,1971). (B-3) " Source Book on Industrial Alloy and Engineering Data", American Society for Metals. ' (B-4-) " Metals Handbook", 8th Edition, American Society for Metals. (B-5) D. L. Harrod and D. A. Kaminski, " Burst Strength of Dented and Cracked 1600 Steam Generator Tubes" Westinghouse R&D Memo 79-102-DEFIN-MI (January, 1977). I (B-6) F. R. Larson and J. Miller, "A Time-Temperature Relationship for Rupture and Creep Stress", Transactions of the ASME, pp. 765-775 (July,1952). (B-7) V. Shah, Presentation to IDCOR/NRC Exchange Meeting on Direct Con-tainment Heating (April 23, 1986). (B-8) "Inconel-600", Technical Bulletin of the International Nickel Company, Inc., 1969, cited in Nuclear Systems Materials Handbook, Volume 1. 1 4

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                                                 ----.%----.-,--gy - - . , - -y ,,-..--, ym - . - _ _ - ,--, ---_ -

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1 l C-1 i l APPENDIX C l Estimation of Steam Generator i Tube Wall Temperatures l MAAP 3 does not contain a detailed model for natural circulation on the secondary side of the steam generators. An overall heat transfer coeffi-cient is calculated in the code which is used to represent the loss of heat from the tubes to the secondary side gas, but this does not consider the detailed behavior in the vicinity of the hottest tube surfaces. This area lies just above the tubesheet on the inlet plenum side on those tubes which . carry fluid from the inlet plenum out to the outlet plenum ("out" tubes). The purpose of this appendix is to determine the relative magnitude of the heat transfer coefficients on the primary side of the inlet part of the out tubes to that on the secondary side. If, as will be shown to be the case, these heat transfer coefficients are comparable, the tubes will achieve a - temperature about half-way between the inlet primary side gas temperature and the average secondary side gas temperature. The procedure to be used is to calculate the natural circulation flows on the primary and secondary sides and to use these flows to calculate the heat transfer coefficients. Since both flows are driven by the high tem-peratures which exist near the inlets of the out tubes compared to the cold temperatures existing elsewhere, scaling arguments can be effectively used to compare the flows and heat transfer coefficients. t As the primary side flow W (see Figure C-1) enters the out tubes, it ~ rapidly cools off from a temperature Tp to a temperature which is essential-ly the mixed average secondary side temperature T . If we assume a roughly ' 3 linear drop in temperature occurs over some length t, the magnitude of the flow is given by equating the total frictional pressure drop required to move fluid a distance 2L on the primary side to the difference in hydro-static head created due to the high temperatures in the zone of length t at the inlet of the out tubes:

C-2 Ts n , i k A 0 3 h

                                                       ,        o                                 "

TP inlet outlet plenum plenum Figure C-1 Natural circulation flows on the inside and outside of a tube carrying fluid from the inlet to the outlet plenum. _ , _ _ . . _ . , , _ . - , - ~ - - ' * " ~ ~ ' '

C-3 2 h AT I" *p g8 t (C-1) [W h([h+II~")2 pDA pt " I where, op= average density of fluid on the primary sides of the tubes, D, = hydraulic diameter on primary side (tube diameter), A = total tube flow area, t a = fraction of tubes carrying out flow; (1 - a) carry the return flow back from the outlet plenum, f = friction factor, s = coefficient of volumetric expansion, g = acceleration of gravity. The temperature decrease in the fluid aT p is given by ATp Wp cp =0 3g (C-2) where, c = specific heat, p Q3g = total heat loss to secondary side. Substituting Q SG tDp At2 W3p

                      , ,2  gg             ]           j                          (C-3) p        c             +

p ft a (1 - a) On the secondary side, we have by the same reasoning

C-4 N aT 2ft s 2a 2g 2D F"Ds 5 98 T t (C-4) s s where, ATs Ns cp =Qg 3 (C-5) or W 3=o 98 2f0go 3 Af (C-6) (U p where sA is the flow area associated with the inlet half of the tube bundle. For simplicity, we have ignored the typically small differences in volu-metric expansion coefficient, specific heat, and friction factor between the primary and secondary sides. If we divide Equation (C-3) by Equation (C-6) we obtain [ j }l/3 Y*1 a l For typical values of a s .2, the last term is very close to 1. 1 The heat transfer coefficients can now be calculated by using the Dittus-Boelter correlation for turbulent flow in passages. Since super-heated steam has a Prandtl number of 1, and if we neglect small differences in thermal conductivity and dynamic viscosity between the primary and ! secondary sides we have:

C-5 g.8 h' (C-8) , 3.s 3.z or s 2'U h (C-9) For Seabrook: At " I*I

• As = 2.95 m 2 Dp = .0155 m Ds = .038 m L = 17 m t s 1-2 m (based on the rate at which the primary fluid loses its heat to the tube wall in the MAAP calculations).

At the time of peak inlet plenum temperature, the primary side pressure is about twice the secondary side value and has an absolute temperature which is approximately 1.3 times as high so that 2h=1.5 Thus 2(1.2)(1.2)(.5)(1)(1.3) s 2 .9

                                                             . . _                                                          =                                    -                 - __   _     . _                          _

C-6 or hsI h p It should be noted that this conclusion is a relatively weak function of the assumptions (e.g., on the value of t), given the small values of the expo-nents in Equation (C-9). As noted previously, if the heat transfer coefficients are equal, in steady-state the tube temperature will be half-way between the primary and - secondary side gas temperatures. Further, by substituting in Equation (C-7), we obtain W3 s5W p so that there will be a relatively small variation AT in s secondary side gas temperature compared to aT p ; thus, the average secondary side temperature can conveniently be used for estimating the tube temperature. A e h l

   . . . ~ , . _      _.___.,r.-,_.,___,,___,___.__,__,.-,m,     _           , _ _ - _ _ . , _ . . _ . , . , , . - . - _ , , - , - _ , _ , . , . , . , . . _ , , _ . _ _ . _ - , _ .      _

_.n - _ , _ - _ _ _ - _ . , _ - , .

RAI 48

Most of the work pertinent to severe accidents has addressed plant behavior at full power, on the assumption that this represents the major contribution to risk. Also, WASH-1400 assumed containment failure was probable following a core melt, making containment bypass sequences relatively _less important. Therefore:

a. Please address the possibility of accidents inside the containment building while in Modes 2-6 (Startup, Hot Standby, Hot Shutdown, Cold Shutdown, and In particular, Refueling) insofar as these accidents could impact upon risk.

consider the ef fect of reduced safety equipment availability and containment integrity requirement permitted by technical specifications while shutdown or refueling.

b. Event V and steam generator tube rupture provide a direct path from the RCS to the environment during severe accidents. Please describe the Seabrook work which identifies any other direct path.

to

c. Please provide f urther information and/or specific ref erences pertinent building release of radioactive material located outside of the containment (e.g. spent f uel pool, radwaste systems) insof ar as the magnitudes are large enough to impact upon the issue under consideration here.

RESPONSE 48 We would like to note that while it is true that WASH-1400 assumed a high probability of containment failure, it is not true that bypass sequences were relatively less important. Most of the containment failure probability in WASH-1400 for PWRS was assigned to the base mat penetration and leakage failure modes and these modes did not result in high contributions to early fatality risk. The V-sequence in f act dominated category PWR-2 which, in turn, dominatedin early fatality risk. Hence, the V-sequence type of bypass was very important the WASH-1400 results.

a. Modes 2-6 The Seabrook Station technical specifications generally require an equivalent level of safety equipment to be available in modes 1, 2, and 3, consistent with the actual mode of operation, for the reactivity control systems, instrumentation, coolant circulation, and ECCS subsystems. The technical specifications require that the accumulators only one train of ECCS subsystems be available for mode 4 and that be isolated. The safety injection pumps are not required to be operable in Mode 4. Technical specifications for containment integrity, containment cooling and support systems such as of f-site power, on-site power, component cooling and Considering service water pumps are generally the same for modes 1, 2, 3, and 4.

that the plant is in modes 2, 3, and 4 for only a small portion of time and the technical specifications generally require an equivalent level of safety equipment to be available, the SSPSA and RMEPS analyses account for the mode l' 2 and 3 , events. The ef fect of reduced safety equipment and reduced containment integrity requirements permitted by the technical specifications while shutdown or refueling are addressed in the response to request for additional information #21.

b. Direct Paths For completeness, all sequences in the SSPSA were considered with regard to their potential for early release. As shown in Figure 1, there are four ways of having an early large release (Release Categories S1, S6 and S7). Each is described below:

Containment Bypass There are two types of containment bypass initiating events in the SSPSA. The interfacing LOCA (V) is quantified in SSPSA Section 6.6 and an enhanced new model is provided in the RMEPS. As shown in Table 1, the V sequences contribute approximately 12% to early release f requency (1.3% of which is in j release category S1 and 10.8% of which is in release category S7). A steam generator tube rupture (SGTR) initiating event can result in an early j bypass release if the secondary side is not isolated and the ability to cool the core is lost. The SGTR model has not changed since the SSPSA (SSPSA Sections 5.3.11 and 5 4.4). SGTR is an insignificant contributor to early release i f requency for two reasons. First, the frequency of core melt is low ( Aprox. 10-6/ year). This low frequency stems from a high degree of plant equipment availability for core cooling and a very long time available for the operator to terminate break flow and maintain basic safety functions. Second, the conditional frequency that the secondary side is open and the operators don't isolate paths to the environment is extremely low. Therefore, SGTR core melt sequences with bypass are insignificant contributors to early release frequency. External Events External initiating events have the potential of f ailing containment and starting an initiating event. Two such events (aircraf t crash and turbine missiles) are in the SSPSA model but neither provide a significant contribution to early release f requency or early health risk. It should be noted that other external events were considered. For example, the seismic capacity of the , Containment (SSPSA, Section 9.2) was found to be greater than 2g and excluded from the detailed event tree quantification. I i i .

    -- -.-_,,~,.m       . . - . _ . _m , _ _ . , . _ . _ . . . _ - - . _ - - - . _ _ . _ . . . .-___,,.._.__...m .. _,m..,...,   __._-___.r, _ . _ . _ _ - . _ _ - _ . _

[ Aircraf t crashes into the containment that can potentially penetrate the containment are assumed to cause a large LOCA initiating event (SSPSA, Section 9.3). The mean f requency of core melt and early release is 1.03 x 10-10/ year (SSPSA, Section 13 2). This analysis has not changed since the SSPSA. As shown in Table 1 air crashes ( APC) make an insignificant contribution to early release frequency. Turbine Missiles that penetrate containment are also assumed to cause a large LOCA (SSPSA, Section 9.9). Core melt frequency is assessed in the plant model (SSPSA, Section 5.3.7 and 5.4.3) with a mean core melt frequency equal to 5.23 x 10-10/ year (SSPSA, Section 13.2) In conclusion, external events impacting containment are insignificant contributors to early release f requency as shown in Table 1 and are minor contributors to early , health risk in RMEPS. To the extent that WASH-1400 (and NUREG-0396) risk curves do not explicitly contain risk from external events, the RMEPS results are conserv-ative in camparison. i Containment Isolation Failure Containment isolation system f ailure and f ailure to recover during a core melt - would result in a potential early release. The systems analysis without recovery is documented in SSPSA Section D.13. The systems analysis model was updated to include operator recovery that was included at the sequence level in the SSPSA. This update is provided in the Risk-Based Evaluation of Technical Specifications For Seabrook Station, PLG-0431 (Section 4.5). Presently, the update and the original SSPSA analysis are being integrated. The top contributor to early release frequency and risk is f ailure of containment isolation caused by seismic initiating events that fail support systems such as the SSPS. As shown in Table 1 containment isolation (CIS) f ailure dominates early release frequency (89%, all of which is S6). Again, RMEPS risk is more complete WASH-1400 and (NUREG-0396) since risk f rom earthquakes were not explicitly included in WASH-1400. Containment Structural Failure Containment structural f ailures due to core melt phenomena is assessed in SSPSA Section 11. This assessment includes an analysis of the ultimate strength of the containment, f ailure modes, and f ailure times as well as ef fects of hydrogen burns, and steam explosions. This analysis has not changed since the SSPSA. BNL reviewed . the SSPSA severe accident phenomena, containment response, and radiological source terms and generally concurred (NUREG/CR-4540) that early f ailure of Seabrook ! containment is very unlikely. As shown in Table 1, steam explosion makes an i insignificant contribution to early release frequency. I i

c. Other Sources As discussed in Section 5.2.2 of the SSPSA (PLG-0300), Initiating Event Identification, if a significant radiological release occurs at a nuclear power plant , "...it must originate either in a damaged core or in a non-core source of radioactivity such as the spent f uel storage pool or the gaseous, liquid, and soild waste f acilities. Past experience and analysis (Ref erence 5.2-1) have clearly shown that releases from the core are by f ar the only significant source of risk at a nuclear power plant

                                                                                              ...It is generally recognized that sources of radioactivity at the plant (other than the reactor core) and the possible mechanisms for their release are such as to provide a negligible risk of public health impact."     We also note that contributions f rom class 3-8 accidents, modes 2-6 events, external events             1 and many common. cause f ailures were not included in WASH-1400 or NUREG-0396.

Ref erence 5.2-1 is NUREG/CR-0603, " A Risk Assessment of a Pressurized Water Reactors for Class 3-8 Accidents," R.E. Hall, et al, October 1979. 5

Table 1 Distribution of Large Early Release Frequency Release Mean Fraction of Sequence Mean  % of Fraction of Category Frequency + Total Type

• Frequency + RC Total S1 5.8 x 10-9 .016 V (pipe break) 4.6 x 10-9 79 .013 Stm Exp1 5.9 x 10-10 10 .002 THLL 5.2 x 10-10 9 .001 APC 1.0 x 10-10 2 . 000 S6 3.2 x 10-7 .889 Earthquakes /CIS_ 3.2 x 10-7 100 .889

. S7 - 3.9 x 13-6 .108 V 3.9 x 10-8 100 ,gog (RHR pump-coal) 3.6 x 10-7

 + events per reactor year V- interf acing LOCA Stm. Expl- Reactor vessel steam explosion TMLL- Turbine missile resulting in'large LOCA APC- aircraft crash into primary containment (resulting in large LOCA)                                                 _

CIS- containment isolation system failure RC- release category a _ _, . . , . ._,...w .- .- ,..m_. -...---,-r-,. r- ..-w., ..--._.---e.e r_.._,--- - . - . -

l PRINCIPAL CONTRIBUTORS TO EARLY RELEASE FREQUENCY INITIATING EVENTS WITH CONTAINMENT . BYPASS

  - INTERFACING LOCAs                                                      *
  - STEAM GENERATOR TUBE RUPTURE 1

l l l EXTERNAL EVENTS WITH - POTENTIAL - 1 CONTAINMENT DAMAGE FOR

    - AIRCRAFT CRASH                                                      >    EARLY      -
   - TURBINE MISSLE                                                            RELEASE 1

i-i i LOSS OF CONTAINMENT ~

                                                                             ~

STRUCTURAL INTEGRITY l ALL OTHER INITIATING

EVENTS

                               '      CONTAINMENT ISOLATION            .

FAILURE

RAI 52

Page 3-7 contains a discussion of vault behavior in response to RHR system [ breaks. The emphasis is upon loss of equipment due to flooding. What consideration has been given to breaks which are small enough that the vault is not flooded, but there is a significant thermal energy release that may impact equipment operation? Picase include consideration that enough energy may be released to activate the fusible links in the ventilation system, thereby terminating ventilation and indirectly causing failure to pumps due to overheating of pump motors, and that this could occur at a time earlier than might occur due to flooding. RESPONSE 52 On subsequent pages in this chapter (e.g. pp 3-22 thru 3-28 and referenced tables) it described how environmental damage to pumps in the RHR vault due to causes other than submergence was assessed. Calculations are presented that show that the leak area of the RHR pump seals must be less than .09 in2 to enable the RHR vault sump pumps to keep up with the rate of floodig. In the event trees of Figure 3-4 and 3-5, leak areas of 0 to .9 in4 are covered by sequences in which the top event L1 is successful - (sequences No. 4-42 in the VI tree and 4-34 in the VS tree, respectively). In addition to submergence by RRR vault flooding, consideration was given to other f ailure modes to RHR vault pumps such as the thermal and moisture ef fects of the steam environment. The f ault tree in Figure 3-9 together with the quantifications in Tables 3-11, 3-12 and 3-13 of PLG-0432 were used to assess the probabilities of environmental failure of the RHR, CBS, and SI pump, respectively. The result of these assessment is summarized in Table 1. As seen in this table, there is a high probability assigned to environmental failure probability of each type of pump even for the range of 0 .09 in2 in which flooding of motors may not occur. While these assessments are highly subjective, they demonstrate adequate consid-erstion of these failure modes and are believed to be conservative. Note that for larger leak rates there is some chance of non-submergence in the injection path (VI) event tree because of the possibility of operator action to ISOLATE the leaking check valves. . In spite of the conservative assessment of environmental damage of. RHR vault pumps, it is clear that the conclusions of the sensitivity study are insensitive to these essessment. The total f requency of non-core melt sequences resulting from meess of event L1 is on the order of 2 to 3 x 10-8/per reactor-year. Even if no credit were taken for any unflooded pumps in this vault, the results would be unaf fected because the charging pumps located outside the vault would still be available for core cooling. Even if it is further assumed that the charging pumps would also f ail during these sequences, there would be no impact on the conclusions of the sensitivity study because of the low f requency currently assigned to the non-melt sequences in which the pumps in the RHR vault do not flood. { .. . .__

The conservative assessments of environmental damage in Table 1 accounts for the direct ef fects of water jets, steam, humidity and thermal damage. These ef fects were assessed to overshadow the rather indirect mechanism of fusible link actuation and overheating due to lack of ventilation. It is not clear whether the fusable links would actuate or not. Even if they did, the incremental thermal stresses acting on the pumps would be small in relation to the direct environmental stresses. Note that sensitivities were assessed on the impact of fusable links on source terms in Section 4 of RMEPS. e S

TABLE 1. Assessment of Environmental Damage to Pumps in RHR Pump Vault RHR Pump Seal Environmental Failure Proability of RHR Vault Pumps Leak Area RHR Pumps CBS Pumps SI Pumps (in 2) E . 0 .09 .55 .1 .1

  .09-1.05          .85                .44                  .33 1

1.05-2.6 1.0 .75 .64

     >2.6           1.0               1.0                   1.0 2

0 .09 .56 .11 .11

  .09-1.05          1.0               1.0                   1.0 1.05-2.6           1.0                1.0                  1.0
     >2.6           1.0               1.0                   1.0

RAI 54

The authors concluded on page 3-9 that presence of water in the reactor cavity will decrease (significantly?) the revaporization of fission products from RCS and perhaps RHR surf aces. We anticipate that a significant quantity of heat producing radioisotopes will remain in the wreckage of the reactor vessel, and this may be ef fective in heating what-ever gases or vapor are flowing toward the break. Has this been investi-gated? RESPONSE 54 Water present in the reactor cavity will essentially eliminate heating of the reactor vessel by the debris. Furthermore, volatile fission products trapped in the primary system and RHR piping will be cooled by the flow of steam from the cavity pool to the RHR pump vaults. Calculations indicate that this flow could cool somewhat more than half of the volatile inventory, were it trapped in the RHR line alone. Larger inventories can be cooled as the decay heat drops in long duration sequences or if credit was taken for direct heat losses from the primary system to the containment. Somewhat smaller inventories could be cooled if the fission products were concentrated at one location, although natural processes would tend to distribute the fission products. 4 r- -

RAI 56

There have been a number of indication (prior to and including page 3-11) that containment spray may be actuated due to RRR relief valve release into containment. What is the justification for this conclusion? Include the effect of containment heat sinks and containment cooler operation in the response. RESPONSE $6 In the early stage of the V-sequence analysis, it was recognized that there . could be some potential for containment spray (CBS) actuation since the RHR relief valves would be discharging to containment (through the pressurizer relief tank); in the interest of completeness, this possibility was included in the modeling. As discussed on page 3-11 (top), automatic CBS actuation occurs on a P-signal which is generated by a high containment pressure of 18 psig (33 psia). The V-sequence behavior was analyzed using the MAAP computer program as discussed at the bottom of page 3-11 and throughout Section 4. MAAP predicted containment pressure is plotted in Figure 4-13. MAAP does account for such af fects as heat sinks and containment air coolers , though for this particular case the coolers were assumed to be not operating. In actuality, 5 of the 6 coolers would typically be running and delay CBS actuation signal even further. It should be noted that CBS pump operation could have a negative af fect on the V-sequence since it directs Ref ueling Water Storage Tank (RWST) inventory to the containment. That is, away from the RCS where it could contribute to core cooling and away from the vault where it could contribute to fission product scrubbing. Thus neglecting the containment coolers is conservative. The discussion at the bottom of page 3-11 provides a somewhat simplified event timing discussion for a particular size of RHR leakage into the vault area. As discussed throughout section 3, the V-sequence analysis includes event trees modeling of the V-sequence over the complete range of RHR system overpressure failure modes and leakage rates. The MAAP model simply included both the possibility of CBS pump actuation and the possibility of CBS pump flood-out. In the MAAP model, if actuated, the CBS pumps operate as long as they are not submerged. (see Table 3-7). As discussed in response number 52, the event trees include pump failure probabilities as a function of RHR leakage for all pumps in the vault area. i l

RAI 58

The last paragraph on page 3-l'1 contains a number of timing of event state-ments. Please provide justification of each. Plots of plant behavior showing suitable parameters and indicating the event points are sufficient for most. Operator response information, in addition to RCS parameter information, is necessary to substantiate the statement that RCPs will be tripped within about 21 seconds of break initiation. RESPONSE 58 The last paragraph of page 3-11 discusses the timing of certain events as predicted by the NAAP program for one particular V-sequence event involving the maximum expected RHR pump seal leak area of 1.3 squared inches for each pump. This particular analysis and its results are discussed further throughout section 4 of PLG-0432. Table 4-7 provides further sequence event , timing information. Plots of several plant parameters versus time (as pre-dicted by MAAP) are provided in Section 4 including the following: Primary System Pressure (Figure 4-11) Core Water Temperature (Figure 9-12)

        - Vault Water Level (Figure 4-20)

Seal LOCA Flow Rate Into Vault (Figure 4-19)

        - Containment Pressure (Figure 4-13)

Enhanced plots of the information provided there are included here as Figures 1 and 2; they show RCS pressure vs. time with various set points superimposed and subcooling vs. time with Reactor Coolant Pump Trip criteria superimposed. CBS pump timing in Paragraph 3-11 is based on an approximate hand calculation for one particular break size. Further CBS pump operation is discussed in responses 52 and 56. Page 3-11 notes that MAAP predicts the RCS to be solid within 30 seconds; this particular event, which is of no significance to the V-sequence, is probably inaccurately predicted by MAAP because the RHR relief valve discharge to the Pressurizer Relief Tank was actually modeled via the PORV as discussed in Section 4.4.3 Operator Response Information: i The Response to Reactor Trip or Safety Injection (E-0) would be the first procedure utilized by the operators. The E-0 procedure specifies on the foldout page (see attachment) that the Reactor Coolant Pumps be tripped whenever the following criteria are met. TRIP ALL RCPs IF ANY CONDITIONS LISTED BELOW OCCUR:

• CCPs or SI pumps - at least one running

                      - and -

RCS subcooling - less than 30*F Phase B Containment Isolation (loss of PCCW)

These instructions are valid throughout E-0. For further information on this subject, see included excerpt from Westinghouse ERP Users Guide on RCP Trip Criteria, section 2.3.2, Evaluation of Alternate RCP Trip Criteria. The trip of the RCPS is a function of the present ERP sets, which is trained upon in both licensed operator training and requalification training. cm 9 [

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O 9 0 0 0 0 0 0 0 0 0 0 0 9 0 0 0 0 0 0 0 0 0- 0 0 4 2 0 8 6 4 2 0 6 6 4 2 4 2 2 2 1 1 1 1 1 RCS P ressur e ll?1l

Calculated Subcooling vs T ime for Worst Case RHR Interf ace LOC A 60 g S 55 ~ '~ ~~' ~ ~~~ ^ ~ ~ U 50 : . b A5 E -- -- - -- - - - -- - t  : g zo _ _- .. . ._. _ . . . _ . . .. . . .. . ._ 0 35 . i - I [ Legend 30 :- I

                                                                                                                                                                                                                                                    "" RCP Trip Criterea n  25 :         --                               -

tR g 20 _ i /\ /'s Subcoolina (calculated) o--o -- N - -- - -- -- -

                                                                                                                 /

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                                ~

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                                     .:.. e....i                        ... i....o . . . . ,                      ....e        ....          ,....e          ....e        ....           i....e ....i                      ....i                 calculated by Seabrook.

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                           -10     :-                                 -                                                                                  - - -       - -       - -                           --                              -

0 5 10 15 20 25 30 35 40 45 50 55 60 65 Time in I-liriotes t I

(E-0) OPERATOR ACTION

SUMMARY

FUK E-0 56 RIES PROCEDURES

1. RCP TRIP CRITERIA l

Trip all RCPs if ANY conditions listed below occur.

                                                                                         ~

e CCPs or SI pumps - AT LEAST ONE RUNNING l

                                            -AND-RCS Subcooling - LESS THAN 30*F e Phase B containment isolation (loss of PCCW)

2. ECCS ACTUATION CRITERIA .

Actuate SI and go to E-0, REACTOR TRIP OR SAFETY INJECTION , Step 1, if EITHER condition listed below occurs: e RCS subcooling - LESS THAN 30*F

                                                               -OR-                                                     .

e Pressuriser level - CANNOT BE MAINTAINED GREATER THAN ' ' 5% [(35)% FOR ADVERSE CONTAIhhENT]

3. EFW SUPPLY Commense CST askeup as soon as possible to avoid low inventory problems.

4. RED PATH

SUMMARY

- ATTACletENT F

5. KEY CAUTIONS e If offsite power.is lost after SI reset, manual action may be required to restart safeguard equipment.

e RCS pressure should be monitored. If RCS pressure drops below 200 PSIG, , RHR pumps must be manually restarted to supply water to RCS. l l

The appropriate instrument uncertainties should be added to the RCS pressure value established above. For normal containment conditions, the normal instrument uncertainties should be used, whereas with adverse containment conditiens, the instrument uncertainties associated with post-acticent containment conditions should be used. The instrument uncertainties should be determined for both the RCS pressure measurement and the steam generator pressure measurement, and the values should be combined in an appropriate i manner to obtain the total uncertainty. The resulting two pressures are the indicated RCS pressure setpoints at which the operator should trip the RCPs, depending upon the steam generator pressure and the containment conditions. To facilitate the use of this parameter, a curve or table can be used which . shows the RCS pressure setpoirt for RCP trip as a function of steam generator pressure for normal and for adverse containment conditions. The setpoint for this parameter could also be expressed as a RCS/ steam generator pressure differential. With this method, the RCS/ steam generator pressure differential setpoint for RCP trip would be equal to the pressure difference from the steam generator pressure measurement location to the RCS pressure measurement location established above plus the combined RCS and steam generator pressure measurement uncertainties. 2.3.2 Evaluation of Alternate RCP Trio Parameters Analyses have been performed to evaluate the effectiveness of the tnree alternate RCp trip parameters for small break LOCAs, SGTRs and non-LOCAs. For each of the accidents, a design basis accident was defined and analyses were performed for representative Westinghouse plants. The details of the accident analyses are presented in Reference 5. I l The objective of the small break LOCA analysis was to demonstrate that the alternate parameters would provide an indication of the need for RCP trip prior to the time when trip is,actually required. For acceptability per NRC letters 83-10c and 10d, the time available from reaching a setpoint that indicates the need for RCP trip to the time RCP trip is actually required i RCP TRIP HP/LP-Rev. 1 0069V:1 17 l

should be at least 2 minutes. To provide a conservative minimum time period to trip the RCPs for each of the alternate RCP trip parameters, tne small break LOCA analysis was based on Appendix K assumptions and the Westingneuse Small Break Evaluation Model WFLASH/LOCTA-IV Codes. The use of best estimate assumptions and mocels would result in longer time periods than tnose cotained with the Appendix K assumptions and models. The results of the small break LOCA analysis demonstrate that the tnree alternate RCP trip parameters (RCS pressure, RCS subcooling and RCS/ steam generator AP) are essentially equivalent in providing an indication to the operator to trip the RCPs during a small break LOCA transient. The results also show that each of the parameters will provide the indication for RCP trip

• sufficiently early such that more than 2 minutes are available for operator action between the time the RCP trip setpoint is reached and the time when trip is required. This was demonstrated for each of the RCP trip parameters without adding any instrument uncertiinty in determining the RCP trip set-points. Thus, each of the alternate RCP trip parameters will satisfactorily indicate the need for RCP trip for a small break LOCA with the instrument uncertainties based on either normal or adverse containment conditions.
• Because each of the alternate RCP trip parameters are adequate to quickly provide an indication of the need for trip during a small break LOCA, the choice of which one to implement at a given plant may therefore be based upon the discrimination capability for SGTRs and non-LOCAs and other plant specific instrumentation considerations.

For the SGTR and non-LOCA events, design basis accidents were defined and analyses were performed to determine the behavior of the alternate RCP trip parameters. The design basis SGTR was defined as a double ended rupture of one s' team generator tube on the outlet sice of the steam generator. The non-LOCA analyses were performed for credible steamline and feedline breaks since it was determined that these accidents result in the most limiting transients among the non-LOCAs considered. The design basis steamline break was defined as an unisolable break of approximately 4-1/2 inches in diameter in one steamline, which is equivalent to one steam generator PORV failing

                                                                                               )

RCP TRIP HP/LP-Rev. 1 0069V:1 18 l

RAI 65

Relative water levels in the RHR vaults and the RCS are mentioned on pages 3-35 and 3-36. What are the water volumes in these regions as a function of elevation? (Of particular interest is the level at the top of the core and at the elevation of the hot leg connections to the RHR.) RESPONSE 65 The volume of the RCS (Reactor Vessel and attached Cold Leg and Hot Leg piping) is 3,332 cubic feet. The center line of the reactor coolant piping lies at (-) 9 feet. The reactor vessel volume at the top of the core is 2,362 cubic feet. The top of the active fuel lies at -14' 2". The connec-tion of the RHR suction nozzle to the loop piping is shown in the attached figure. The equipment elevations, flood elevtions and flood volumes are

• shown in the attached figure for the containment building spray pumps, the CBS vault sump pump, the RHR pump and the safety injection pump.

I FIGURE 1 RHR CONNECTIONS TO REACTOR COOLANT LOOP PIPING

                                                 ??

I p t. . . . - ._ G.-.-. s 12' RESIDUAL HE AT - 12" RE5iDJAL HL AT REMOVAL , %MO/AL

                                                                , /                     .

LOOP 4 LOOP 1

                                                 -h.

1

I t.

                                             -                                                                                                   Safety insecten Pump j       '
                                                                            ,                                                                    Motor Center hne Elevation (-) 47 1S-
                                                . g   -
                                                                                                                                            / Flood Elevaten 47.6 FT
                                                                                                                                          /      Flood Volume 23634 FT3 LT A                     i l

I II Contamment SuMing Spray Pump Motor Center Lme Elevaten ti 58* 5' / Flood Elevaten 58 4 FT , i , / RHR Pump Flood Volume 4783 FT3 " p Disenarge Elevaten H 56' 6-

                                                                                      ."                                                  #          Seat Flood Elevaten 55 0 81
                                             \      ,           J,   I        g,    3            ?          ,

Seal Fiood volume 10729 FTl b (-) 61' 0" - # ' u! CBS Vault Sump Pump Flood Elevaten 59 5 FT Flood Volume 2898 FT3 I l - l Figure 2 RHR Equipment Vault Volumes and Elevations a I

l

RAI 66

4 What is the justification for the statement on page 3-36 that the water level in the vaults will be approximately the same as that in the RCS? (We do not agree because of the potential that pressure in the vaults and containment are not the same, and water temperature in the two locations ) l may differ). RESPONSE 66 Page 3-36 (and preceeding) discusses potential operator actions to recover the V-sequence and prevent core melt. It states that if the LOCA can not be isolated, the water level in the vault will rise unt'll the levels in the vault and in the RCS are approximately equal. This is a general statement made only in evaluating possible operator action; the exact levels are not important to operator action. As the question states, pressures and temperatures may differ in the vault and in the RCS/ containment; however, since these areas are communicating through the RHR leak, the dif ferences would not be significant. Also, the dif ferences might tend to force additional water into the vault area (since RCS/ containment pressure would be greater than vault pressure) providing additional fission product scrubbing. It should be noted that pressure / temperature dif ferences of interest here are between the vault and the RCS not between the vault and containment. Other than RHR relief valve discharges, there is no communication between the containment and the RCS or vault area (at least until vessel melt-through) unless a RRR relief valve should stick open. i e e a

  ?            y --y   e--,,-,. m- - , ,pm, _-g--pyv.w ,y w m,y,7,m.w. . , , , ----..a. -.-.g ,,--9,_w-. -,-.y,--,.w, mm .y-,,.       ,.,g- ---- ----   -w

_ - = .

RAI 74

A tacit assumption appears to be incorporated into References 1 and 2 that check valves are always closed. In reality, many check valves require a (substantial) reverse flow to force then to close, and they additionally of ten require a significant reverse pressure to keep them closed. Is this the case for any of the valves of interest here? If so, please discuss the implications. If not, what is the justification f or the conclusion? RESPONSE 74 It was not merely an assumption that resulted in the omission of sequences involving check valve failures to close but rather the result of a considered assessment. First of all, .our knowledge of the design characteristics and tech spec surveillance requirements that govern the RHR system during normal power operation give us high confidence that the check valves in the RHR and interf acing systems are initially closed. The check valves of interest here, those that separate the RHR system from the CVCS, SI system, CBS, RWST and containment sump are all designed to close without the need to apply reverse pressure. Hence it is necessary 1 to postulate that these valves f ailed to close the last time they were opened and that the failures remained undetected until the time of the postulated "V-sequence". Secondly, even if one assumes, conservatively that all these check valves are initially open, our knowledge of check valve failure rates for f ailures to close, which have taken into account the San Onofre Unit 1 experience, gives us high confidence that any potential for check valve failed open bypass pathways is greatly over shadowed by the probability already assumed for RHR piping f ailure given RHR pressurization. In the discussion attached we presented our assessment of check valve failure rates in light of San Onofre Unit 1 experience and our conservative bounding assessment of sequences involving assumed failed open check valves. In view of these assessments, the conclusions of the sensitivity study are generally insensitive to assumptions regarding check valve performance.

1 CHECK VALVE FAILURE RATES USED IN THE SEABROOK EPZ STUDY The following provides further documentation for the estimated failure rates of the check valves in the pressure boundary of the ECCS and RCS, as modeled in the Seabrook EPZ study (Reference 1). The failure modes of concern are disc rupture or gross leakage of a check valve that is initially seated and tested to verify its position and failure of check valves to reseat on demand.

                                                           ~

1. DISC RUPTURE / GROSS LEAKAGE 1.1 DATA COLLECTION The source of event descriptions for this analysis was Nuclear Power Experience (Reference 2). In this search, which was_done manually based on the NPE key word index, a total of 610 check valve failure events were identified ( Attachment A is an expanded list of these failure events, ,

based on the NPE-automated retrieval system--a total of 692 events). The initial list was then reviewed to identify leakage events (external and internal) in all systems of both PWRs and BWRs. A total of 163 events were identified. These events are marked "L" in Attachment B. This review also provided further evidence that a large number of check valve leakage events should not be considered for the failure mode of interest in the V-sequence analysis, either in terms of mode or cause of failure. ! It was then decided to limit the data base to those events involving check valve leakage in the ECCS and RCS/ECCS system boundary of PWRs. . These were judged to be the closest category to the initially seated and tested check valves modeled in the analysis. No disc rupture events were identified in these events, and the maximum observad leak rate was 200 gpm. The majority of events involved very small leaks. Those considered to be more significant are listed in Table 1. Even for the cases listed in Table 1, the exact leak rates were not always provided in the event reports. Consequently, leak rates were estimated, based on

 ,          other available evidence: the rate of boron concentration change, j            pressure reduction, and similarity to other occurrences for which the
!           leak rates were known.

Also considered was an event that occurred in San Onofre Unit 1 in November 1985. The event involved failure of several main feedwater pump discharge check valves to reseat on demand resulting in overpressurization of the main feedwater system. A summary of the event, 4 as presented in Nuclear Power Experience (Reference 2), is provided as i j Attachment C. As can be seen from the event description, four of the five failed check valves failed to reseat when the main feedwater pumps tripped. These

failures obviously do not apply to the disc rupture / gross leakage mode of failure considered for the ECCS/RCS check valves. However, as it is described later, they are included in the estimated frequency of failure to reseat on demand. The fifth valve (feedwater regulating valve bypass line check valve) failed because of water hammer resulting from i

1 1423P091686

TABLE 1. CHECK VALVE LEAKAGE EVENT DATA BASE Sheet 1 of 2 NPE Plant * ** Reference (date) Event Description Range (gpm)- V11. A.126 Zion 2 A leak rate of ~0.25 gpa was detected from

                                                                                                                    ~

y --0.25 (October 1975) the "A" accumulator check valve - wrong size gasket installed. V11.A.32 Turkey Point 4 One of the three check valves in the high-head y ~~~0.33 (May 1973) safety injection lines to the RCS cold legs developed 1/3 gpm leakage with 180 psi of water pressure applied. Two other check valves showed only slight leakage - failure of sof t seats. V11.A.175 San Onofre 1 A tilting disc check valve located in the LPI y<5 (May 1978) system as the first valve inside containment, , failed to close with gravity - valve installed in a vertical rather than a horizontal pipeline. V11. A.114 Surry 1 Check Valves 1-51-128; 130 leaked causing boron y < 10 (July 1976) dilution in the "B" accumulator, y < 10 V11. A.182 Calvert C11ffs 2 The outlet check valves associated with the y < 10 (September 1978) safety injection tanks 218 and 228 leaked y < 10 reducing the boron concentration from 1,724 and 1,731 ppe to 1.652 and 1,594 ppe in 1-month period, respectively. V11.A.306 McGuire 1 Discharge check valves associated with the cold y < 10 ( April 1981) leg injection accumulator A leaked - cause y < 10 unspecified. V11.A.343 Point Beach Check valve 1-853C, serving as the first-off y < 10 (October 1981) check valve from the RCS for the low head safety injection. Y11.A.291 Surry 2 Check valve associated with the safety y < 20 (January 1981) injection accumulator "C" leaked, resulting in accumulator boron dilution - cause unknown. V11.A.63 Ginna Accumulator "A" check valve leaked leading to y < 20 (September 1974) boron dilution (from about 2,550 down to 1,617 ppm) - cause unknown. V11.A.85 Surry 1 Check valve associated with the IC accumulator y < 20 (August 1975) failed to seat, resulting in increase in accumulator level - cause unspecified. developed 6 gpu leakage. V11. A.105 Robinson 2 "B" safety injection accumulator check y < 20 January 1976 valve developed leakage - cause unspecified. V. A.122 Zion 1 Discharge check valve on the accumulator 1D y < 20 (June 1976) developed back leakage - cause unspecified. V.A.407 McGuire 1 Cold leg injection accumulator check valve 20 < y < 50 (May 1983) leaked, resulting in low accumulator boron concentration - cause unspecified. V.A 452 St. Lucie 2 The SIT outlet check valve developed excessive 20 < y < 50 (December 1984) leakage - foreign material caused ball galling leading to joint binding. 1422P091686 2

i TABLE 1 (continued) Sheet 2 of 2 NPE Plant **

                                                      "     **"'             "                     *"I' Reference             (date)

(gpm)- V.A.456 Calvert C1tffs 2 SIT check valve developed excessive leakage - 20 < y < 50 (January 1985) ethylene propylene 0-ring material degradation. V.A.437 Farley 2 Loop 3 cold leg safety injection check valve 50 < y < 100 (September 1983) developed excessive leakage - incomplete contact between disc and seat. V.A.273 Davis Besse 1 Gross back leakage through core flood check 20 < y < 50 (October 1980) valve - cause unspecified. V11. A.384 Calvert Cliffs 1 JIT outlet check valve leaked at the rate of y -~200 (July 1982) 200 gpa ring deteriorated. l f I, 4 l

1422P091686 3 l

the failure of other valves. This failure also does not apply to the failure model of interest here. In the process of reviewing the available data, a recent review of eight BWR events (Reference 3) was also con.sidered. These events, li.sted in Table 2, involved testable isolation check valves in the pressure boundary and could be considered as precursors to an interfacing LOCA.- These events were judged to be inapplicable for this study because the valves involved are different from those considered here both in terms of desiga and operation. The reasons for inapplicability of each of the events are listed in Table 2. In summary, the BWR check valves have air operators, whereas the PWR ECCS/RCS check valves are enclosed and cannot be operated from outside. The latter group is verified seated, either continuously (for the upstream valve) or during startup (for the downstream valve). Thus, the same mechanisms that cause the eight BWR check valves to be open and undetected do not apply to the PWR ECCS/RCS check valves. 1.2 SUCCESS (EXPOSURE) DATA To estimate the total check valve hours, the information provided in NUREG/CR-1363 on the number of valves in the ECCS and RCS in various PWRs was used. The details are prgvided in Table 3. The total number of check valve hours is 1.0 x 100 1.3 FAILURE RATE ESTIMATE 1 The various leakage events were grouped into five leak ranges, as shown in Table 4. For each group, a frequency per hour was estimated using the exposure time discussed above. Table 4 also provides the corresponding cumulative frequency points that are also shown in Figure 1. The curve fit on a log-log scale was done using an IMSL code, which uses the least square method. The parameters of the line obtained from this method correspond to the Bayesian most probable values based on a uniform prior distribution. The equation of the line is

                    -y = ax +b where x is the logarithm of the leak rate (gpm) and y is the logarithm of the frequency of exceedance per hour.

Using the data of Table 4 the following values for a and b were derived:

e Parameter a Mean = 0.0976 95th Percentile = 1.0127 Sth Percentile = 0.6915 f

Jf 1423P091686 4 7 --F --

TABLE 2. SUPNARY OF OPERATING EVENTS Event Percent System l Date Power Involved Status Cause Reason for Inappifcabf11ty Vermont Yankee 12/12/75 99 LPCI/RUR Open Unknown PWR ECCS/RCS check valves are tested i LER 75-24 and verffled seated initially. They l I can not be left open undetected. j Cooper 01/21/77 97 HPCI Open Loose Part LER 77-04 PWR ECCS/RCS check valves are tested 1 Obstruction and vertfled seated. Any initfal I leakage or failure to be in the seated posf t1on will be discovered before the } plant goes to power. l LaSa11e-1 10/05/82 20 HPCS Open Orfed Lubricant and LER 82-115 PWR ECCS/RCS check valves do not have Insufficient Preload air operators. They can not, therefore. in Air Operator; be opened externally. Opened Bypass Line LaSa11e-1 06/17/83 48 HPCS Open Thermal Binding; j LER 83-066/03L Check valve failed to close due to disk En Opened Bypass thermal binding. The PWR ECCS/RCS check Line valves are required to hold against RCS j pressure af ter being veriffed seated initially. These valves are closed and stay closed. They are not cycled; i therefore, the failure modes are different. I LaSa11e-1 ' 09/14/83 0 LPCI Open Maintenance Errors PWR ECCS/RCS check valves are tested LER 83-105/01T and vertfled seated before the plant ' goes to operation. Pfigria 09/29/83 98 HPCI Open Rusted Lfnkage on LER 83-48 Air Operator .l Hatch-2 10/28/83 90 LPCI Open Maintenance Errors LER 83-112/03L PWR ECCS/RCS check valves do not have on Air Operator air operators and will not open due to a sinflar maintenance error. j Browns Ferry-1 08/14/84 100 LPCS Open Maintenance Errors PWR ECCS/RCS check valves do not have 1 LER 83-032 on Air Operator air operators and will not open due i to a sinflar maintenance error. [ i l 1422P091686 j ,

TABLE 3. CHECK VALVE EXPOSURE DATA Nisaber of Start of -Number of Check Valves Total Number of Plant Name Code Comunercial Operation Years in ECCS Check Valve Hours Arkansas Nuclear One 1 AR1 December 1974 10 20 1.75+6 Crystal River 3 CR3 March 1977 7.75' 23 1.56+6 Davis-Besse 1 081 November 1977 7.08 s 26 1.61+6 Oconee 1 OE1 July 1973 11.42 20 2.00+6 Oconee 2 CE2 March 1974 10.25 21 1.89+ 6 Oconee 3 DE3 December 1974 10 21 1.84+6 Rancho Seco RSI April 1975 9.67 30 2.54+6 Three Mile Island 1 TII September 1974 10.25 19 1.71+6 Three Mile Island 2 TI2 December 1978 6 19 9. 99+ 5 Arkansas Nuclear One 2 AR2 March 1980 4.75 30 1.25+6 Calvert Cliffs 1 CC1 May 1975 9.58 45 3.78+6 Calvert Cliffs 2 CC2 April 1977 7.67 45 3.02+6 Fort Calhoun FC1 September 1973 10.25 45 4.04+6 Millstone 2 MI2 DecenL2r 1975 9 47 3.71+6 Maine Yankee MY1 December 1972 12 49 5.15+6 Palisades PA1 December 1971 13 21 2.39+6 SL1 December 1976 8 30 2.10+ 6 it.Lucie1 eaver Valley 1 BV1 April 1977 7.67 36 2.42+6 D. C. Cook 1 DC1 August 1975 9.33 34 2.78+6 D. C. Cook 2 DC2 July 1978 6.42 34 1.91+6 ! Haddam Neck HN1 14 27 3.31+6 Indian Point 2 IP2 Januar5741968 July 1 10.42 36 3.29+6 Indian Point 3 IP3 August 1976 8.33 45 3.28+6 Joseph M. Farley 1 JF1 December 1977 7 33 2.02+6 Kewaunee KE1 June 1974 10.5 19 1.75+6 , North Anna 1 NA1 June 1978 6.5 3 2.05+6 Prairie Island 1 PRI December 1973 11 23 2.22+6 Prairie Island 2 PR2 December 1974 10 23 2.01+6 Point Beach 1 PT1 Deces6er 1970 14 21 2.58+6 Point Beach 2 PT2 October 1972 12.17 21 2.24+6 R. E. Ginna 1 RG1 March 1970 14 21 2.58+ 6 1 H. 8. Robinson 2 R02 March 1971 13.75 25 3.01+6 Salem 1 SA1 June 1977 7.5 32 2.10+6 San Onofre 1 501 January 1968 14 18 2.21+6 Surry 1 SUI December 1972 12 25 2.63+6 Surry 2 SU2 May 1973 11.58 25 2.54+6 Trojan TRI May 1976 8.58 22 1.65+ 6 Turkey Point 3 TU3 December 1972 12 34- 3.57+6 Turkey Point 4 TU4 September 1973 11.25 34 3.35+6 Yankee Rowe YR1 June 1961 14 17 2.08+6 Zion 1 211 Decenter 1973 11 50 4.82+6 Zion 2 ZI2 September 1974 10.25 50 4.49+6 Total 1.08+8 NOTE: Exponential notation is indicated in abbreviated forin; i.e., 1.75+6 = 1.75 x 106 , 1422P091686 < l

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

l l l TABLE 4. STATISTICAL DATA ON CHECK VALVE LEAKAGE EVENTS IN PWR, ECCS, AND RCS SYSTEMS F Leak Rate Number of gf 0c u re ce Frequency of (gpm) Events Exceedance (per hour) 5 3 2.94-8 2.06-7 10 7 6.86-8 1.77-7 20 5 4.90-8 1.08-7 50 4 3.92-8 5.90-8 100 1 9.80-9 1.96-8 200 1 9.80-9 9.80-9 NOTE: Exponential notation is indicated in abbreviated form; i.e., 2.94-8 = 2.94 x 10-8 7 1422P091686

l l l l 9 - i 4 4 i: e iilij\i a ei6iilj l

                                                                                              -   6 I i i iil l            4             4 4 44 4iL 6      -             s\ N               \                                                                                                  :      o I

5 - g 4 - N , 3 -

                          \                       \g                     N                                                                               -

2 - \

• N. s -

N e N - N N N *

                                                                       \
                                                                          \

1 a 10'9 g 7

                                         \                                   N g                                                                        -

6 - g \ o N g s 5 -

                                                \                                           N                                                           -

4 -

                                                  \                                             \                                                      -

3 .

                                                    \g                                                   N N
                                                                                                                         \                             -
                                                                                                                                                        =

2 -

                                                                                                               \

N . s -

                                                  \           \                                                      \

e 1 a 10 -- s g e N N

                                                                                                                                               \  --

c  ? : E 6 - N s  :

  <           5     -
                                                                \            \                                                               \

0 4 -

                                                                                  \                                                            \ -    -

3 - y 2 - N Ng - E

                                                                                                  \
  $ i ioj
                                                                                                          \                                      --

8 - 7 6 LEGEND: \ \ - 5 -

                          - --- - - STATISTICAL 80UNOS                                                           N                                   -

4 . AT 90% CONFIDENCE \ - 3 - 8EST FIT g \ - 2 -

                                  - - ASSUMED 95TH AND STH PERCENTILES                                                       \                      ~

i io-', -- N \ -- 8 - 7 - N '

                                                                                                                                                    ~

6 - ~ 5 - - 4 - , 3 - , 2 - 1 10' " ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' i io too ' ' ' ' ,000 1 10.000 CHECK VALVE LEAK RATE (GPMI FIGURE 1. FREQUENCY OF CHECK VALVE LEAKAGE EVENTS 8

e Parameter b Mean = 13.6943 95th Percentile = 14.2862 . 5th Percentile = 13.1024 Based on the above values, the "best-fit" line is -

           -y = 13.6943 + 0.0976 x with the following bounds:

95th Percentile = -y = 14.2862 + 1.0127 x 5th Percentile = -y = 13.1024 + 0.6915 x These lines are shown in Figure 1 as the "best-fit line" and " statistical bounds at 90% confidence." To account for uncertainty in the assessment , of the leak rates, classification of data, estimation of exposure data, and the applicability of the data to the check valve and ' failure mode of concern in this analysis, the statistical bounds were further stretched by increasing the range factor of the frequency at 150 gpm from 3.7 to 10 and increasing the range factor of the frequency at other points proportionally (to RF = 14 at 1,800 gpm). The resulting new bounds are also shown in Figure 1.

2. FAILURE TO RESEAT ON DEMAND To estimate the frequency of check valve failure to reseat on demand, two types of data were used: (1) estimates from several generic sources of failure data, and (2) experiential data from eight U.S. nuclear power plants based on plant-specific PRAs performed by PLG.

Since the majority of data sources provided information on check valve failure on demand without specifying failure to open and failure to close modes separately, the distribution developed here is based on failure on demand data. Review of check valve failure events from several plants indicate that the distribution is a good (and perhaps even conservative) estimate of the failure to reseat frequency. An additional piece of information provided by the San Onofre event of November 1985 (Attachment C) was also incorporated into the estimate of - check valve failure on demand frequency. Four of the five check valve failures (failures involving pump discharge check valves) apply to this mode of failure. NPE was reviewed for the period January 1,1971, through June 30, 1986, to see if there have been other check valve failures in the San Onofre main feedwater system. None were found. l The corresponding success data (number of demands) were developed by - assuming an average of 10 system-wide demands per year, a population of eight check valves, and 15.5 years of operation from January 1,1971, 4 through June 30, 1986. This resulted in an estimated 1,240 check valve 1423P091686 9

demands. The corresponding failure frequency estimate (counting four of the five check valve failures) is A so

              =
                           = 3.2 x 10'3 per demand.             '

This value was used together with the generic estimates as well as plant-specific data from other plants in a Bayesian updating process - described in Reference 4 to develop the failure on demand frequency distribution. '- , The following summarizes the data used. e Generic Estimates Source Estimate Assigned Range Factor

• WASH-1400 1.00 x 10-4 5 NUREG-1363 1.10 x 10-4 3 EPRI-81 7.00 x 10-5 10 e Data from Nuclear Power Plants Number of Plant Events Number of Demands Oconee 3 6,855 7 Zion . 0 6,970 Indian Point 2 0 1,440 Indian Point 3 0 1,550 Beznau (2 Units) 7 28,978 Pilgrim 0 2,394 TMI-1 12 8,716 e San Onofre Unit 1 Main Feedwater System Check Valves Estimate = 3.23 x 10-5 Assigned Range Factor = 5 l

(A moderate range factor is used to represent higher degrees of uncertainty than indicated by the estimated four events in 1,240 demands.) 1

  *The assigned range factor (ratio of the 95th to the 50th percentile of lognormal) represents our uncertainty of the accuracy of the estimate.

See Reference 4 for the details of the methodology. 10 1423P091686

The resulting distribution is shown'in Figure 2. Some key characteristics are: Mean Median Per e tile Percen ile _ 5.46-4 1.18-5 1.58-4 1.63-3 NOTE: Exponential notation is indicated in abbreviated form; i.e., 5.46-4 = 5.46 x 10-4

3. REFERENCES

1. Fleming, K. N., A. Torri, K. Woodard, and R. K. Deremer, "Seabrook '

Station Risk Management and Emergency Planning Study," prepared for Public Service Company of New Hampshire and Yankee Atomic Electric Company, confidential, PLG-0432, December 1985,

2. Nuclear Power Experience, S. M. Stoller Corporation, updated monthly.

3. U.S. Nuclear Regulatory Commission, "Prelimin3ry Case Study Report, "

Overpressurization of Emergency Cocling System in Boiling Water Reactors," February 1985.

4. Mosleh, A., "PRA Data Base," Pickard, Lowe and Garrick, Inc.,

PLG-0500, 1986. t a 1423P091686 11

m. m. - -

WASH 1400 NUREG-1363 EPRI-81 SAN ONOFRE 1 i;; LUMPED DATA FROM 8 PLANTS (22 FAILURES IN 56.903 DEMANDS) C 2

                 "  E                                                                                                      .

8 i E t l l l 10-6 10-5 10'4 10-3 10-2 i FAILURES PER DEMAND FIGURE 2. CHECK VALVE FAILURE ON DEMAND FREQUENCY DISTRIBUTION \ _ _ _ _ . _ _ _ __ _____1_

4 Q 0 ATTACHMENT A

           , e-- , -,- .--      - , - -   - . - - .    - -   n.

PhEE 1 NPE C4ECK VALVE LISTI16 - 493 ARTICLES TOTAL THERE ARE 73 ARTICLE ($3 TO GE DISPLAYED: BWR 3.04.s.0010 : VARIOUS CONTROL SLADE, DRIVE PR03LEMS

                                                                                            - CNECE VALVE FAILURES - 19 b-19F5
                         --- 3RESDEN I - 1960-72                                                                                               4 8.04.e.0036 :                                                                                               .

DISK AND SEAT SURF 4CES DIRTY - CRB CNECE WALVE LE4fAGE *

                         --- DRESDEN 2 - 14T F5 - REFWELING SNUTD0d4 9.05.E.0002 : CHECE---

VALVE LEAE4GE DRESDEs 2 - SPRING 72 i

3.05.E.0003

FW CNECK VALWE LE4K4GE

                        --- DRESDEN 2 - SPRING 71                          -

9.05.E.0014 : FW CNECC V4LWE LE4E ASE - CHA1GEDT3 SILIC3M SEAL RING i --- DRESDEN 3 - SPRING 1974 3 j 9.06.E.0018 : 3-R!gGS ADDED TO FW CHECC VALVES ! --- DRESDEN 3 - 157 1/2 0F 1773 '# 5.06.E.0022 : CMECC VALVE 3 LUG LEAC

                       --- OYSTER CREEK - JULY F3                                                                                   ~

J 5.06.E.0023 : SILICONE

                       --- DRESDEN   0-RINGS UN4CCEPT48LE - VALVES LE4KED                                                                    J,

. 2 - N4R S N4Y 74 t.0$.E.0024 : FW CHECK DISC DID NOT MATE WITN SEAT ()

                      --- MONTICELLO - SPRING F4                                                                                                   <
;   5.05.E.0029 : LEAKT GASKET IN FW CHECE W4LVE                                                                                            C)     ,
                      --- 4RNOLD - JULY 74 (POWER ESCAL4T!3N TESTING)                                                      #

{ 8.06.E.0034 : i LEAK 4GE BETWEEN CNECE VALVE SEAT RINS AND V4LVE S0DY p 4 i

                     --- 20AD-CITIES 1 - APR 74 EREFUELING SNUTD3WN) 3.05.E.0040 : FL4W IN FW CdECK VALWE                                                                                                    J
                     --- GUAD-CITIES 2 - FEs 75 - REFUELING $NUTDOWN j s.06.E.0042 : CNECE VALVE SEAL RING LEAEAGE                                                                                              .)
                     --- 3UAD-CITIES 2 - DEC F6 & 75 - REFUELI4E SNUTDOWN 3

S.06.E.0043 : l SEAT LEAKAGE AND CRACEED SELLOWS IN FW CNECES >

                     --- PEACM SOFTOM 2 - MAY i                                                   & JUNE 73 - SMUTDOWN AND 45E P3WER 8.06.E.0048 : ERODED CONTR3L VALVES TARAPUR 1 8 2 - 1970 J     l 5.05.E.0058 : ---   FW C4ECK VALWE 0-RINGS C04*RESSED - EXCESSIVE Lent 4GE DRESDEN 3 - APR 8 NAT F5 - REFUELING SHUTOOW9 g
' 3.05.E.0059 : FW---CHECK V4LfE 0-RINGS DISASSOCIATED DRESDEN 2 - N3V 74 - REFJELING SMUTD0d4 II e.05.E.0075 : PIN MOLE LEA (5 IN FW CHECC VALVE SEAL PLATE, CRACKED C01V3 LUTE
                    --- PEACH SOFTON 3 - AUG 75 - STARTING UP

_v, _

                                                                                      .                                               (> '

o PAGE 2 D.06.E.0001 : . CRACCED FW VALVE SEAL PLATE BELLOWS

                          --- PEACM DOFTON 2 - JUNE F5 - F40 MWE 5.0$.G.0002 :                                                                                                                3
                          .9ISC. TURDINE CVCLE STSTE15 PRODLEis
                          --- NUNLEDER3 - 1972-F4 i

' D.07.4.0015 : s) CHECC VALVE DISC-NINGE JOINT FRACTURED

                          --- OUAD-CITIES 1 - APR F4 (SMUTD0dm FOR REFUELING) 3.0F.4.0020 :                                                                                                              g)

RCIC TURDINE ENNAUST CHECC ASSEN9 LED INC0tRECTLY NONTICELL3 - JAN F5 - SMUTDOWN D.07.4.0031 : O CHECC VALVE SLOWI15 STEAM, R3D CONTR3LS N3T CNECEED j

                         ---   FITIPAftICE -

SMUTDOWN JULY 75 - (POWER ESCALATION TESTINS) - j I D.07.A.0032 : EXNA'JST LINE CNECC VALVE CLAPPER DISCONNECTED

                         --- 3RUNSw!CC 2 - APRIL F5 - (POST CRITICAL TESTING)DE P3WER                                            #

S.07.A.0034 : EINAUST CNECC VALWE FLAPPER WAS J41NED

                        **- DRUNSWICC 2 - OCT F5 - 63t POWER                                                                     #

D.07.A.0040 : RCIC TURDINE DISCNAREE CHECK VALVE FLAPPER DISC STUD DR3CEN

                        -** 3RUNSd!CC 2 - FED F5 - M3r STAND 37                                                                  >

p.0F.4.0043 : TORUS

                       ---    COOPERDISCNARGE        VALVES
                                       - OCT F5 - SMUTD0dN    NAD DIRTY SEATS AND MISSING FASTENERS J

D.0F.C.0014 : SCALE ON VALWE SEATS

                       --- 10NTICELLO - SPRING F4                                                                              I t

! D.07.C.0030 : SCALE ON CHECK VALVE SEATS

                      --- 10NTICELLO - JAN F5 - SMUTDOWN D.07.C.0042 : IMPROPERLY 43 JUSTED CNECE VALVES
                      --- 3RUNSWICC 2 - NARCH F5 (POST CRITICAL TESTING)J 58 P0 DER s.0F.C.0043 : CHECC VALVE SEAL LEACED
                     --- TARAPUR 1 8 2 - 1769 OR F0                                                                          ~

S.07.C.0044 : STUCC (DECC dALVES T.AAPUR 1 & 2 - PRIOR 70 1973 D.07.C.004F : SCALE ON SEATING

                     --- 9ONTICELLO        - SEPTSURFACES F5         OF CORE SPRAY CNECE VALVES

D.0F.C.0061 : CORE SPRAY DEPRESSURIIE3 C1ECE VALVES SEATED INPROPERLY - DISCNARGE MEADER FG DRESDEN 2 - JAN F5 - REFUELING SNUTD0d4 --- DRESDEN 3 - AUG 1

J \ D.07.D.0016 : WIRE AND G1110ER ARDOR F3UND IN RNR SYSTEM

                    --- JUAD-CITIES 2 - JAN F3 SP 9.07.0.0035 : SCALE ON VALdE SEAT                                                     ,

90NTICELLO - SPRING F4 4, 0 9.07.0.0061 : SCALE ON CNECC VALVE SEAT MONTitiLLO - JAN FS - SMUTDOWN C) I

5

PAGE 3 i

5.0F.D.3076 : RNR AND NPCI VALVE PACKING LEAES

                                                      --- 3RUNSWICC 2 - SEPT F5 - 521 P0 DER

• 3.07.D.0082 : VALVE LEAKAGE, LPCI LOOP 3VERPRES$URIZED
• NEAT EXCNANGER e

GASKET FAILED

                                                      --- VERNONT TANEEE

• DEC F5 - 993 PodER gj 3.0F.D.3104 i

TORUS SPRAT VALVE OPERATOR NOUNTINS PLATE WELD FAILED, CRACKED T3EE ARM - 1150FFICIENT dELD PENETRATION, UNDERSIZED MOUNTING SOLTS

                                                     --- 3ROWNS FERRY t - MAT $ SEPT 76 - COLD SNUTDodu

$ 9.07.E.000T : C) l STEAM SJITCM VALVE DAMAGE DISC PIN FAILURE - RUPTURE DISCS RUPTURE - TEMP ! 10NTICELLO - JULT F2

• 3 3.07.E.0003 : STEAM VALVE LEAKAGE J

                                                     --- 10NTICELLO - JULT F2

' 3.0F.E.0011 : LOOSE fuRBINE RUST - CNEtt VALVE LEACED TORUS bATER BACE TOWARDS 1PCI

                                                     --- PILGRIM - JULY F2 (P0 DER ESCALATION TESTING) i 8.0F.E.0018 : VACUJM IN TUR8INE EXNAUST - d4TER WAMMER                                                                                   .*
                                                     ==- VERMONT TANKEE - 1971 (PREOPERATIONAL TESTING) 4.0F.E.0025 :                                                                                                                             J 4

WATER

                                                    --- 8ROWNSHAMMER FERRT OF                 TUROINE EENAUST - INSTALLED      CONDENSING SPARGER 1 - OCT F2 (PREDP TESTING) 9.0F.E.0048 :                                                                                                                           $3 4
                 ,                                  ELAND CONDENSER GASKET FAILED - P3SSIBLE INJECTISN VALVE LEAEAGE
                                                    --- 3ROWNS FERET 2 - NOV F4 - 422 POWER s.0F.E.0053 : SURFACE FLAWS IN VALVE DISC AND SEAT                                                                                      O
                                                    --- PEACM BOTTOM 2 - FEB F5 - SNUTD0J4                                                           '
,                        0.07.E.0054 :                                                                                                                            a SENT MINGE PIN AND SAD DISC ON TURSIME EMMAUST CNECK VALVE 10NTICELLO - JAN SNJTDOWN 9.07.E.0058 : MISSING     CNECC, WELD IN SWING CNECK, INSUFFICIENT TACK WELDS IN STOP VALVE DIS $4SSEM8 LED WITH C30LANT, GREATER TNA1 212                                                      I 3                                                   DEGREES F j                                                   --- ARN0tD - APR 75

• 9.07.E.0061 : STE49 VALVE LEAKED

                                                   --- SUAD-CITIES 2 - DEC F4 - REFUELI1G SHUTDOWN                                                               .)

8.07.E.00FF : SCRATCHED --- MONTICELLO SEATING $URFACES - TURDINE EXNAUST VALVE LEAKAGE

                                                                                   - SEPT F5                                                                     J 5.07.E.0130 :

MPCI TURSINE EXNAUST CHECK VALVES LEAKED

• WORN VALVE SEATS I

                                                   ---                                                                                                           J l ,                                                 1 - SUAD-CITIES JAN F6 - REFUELING         2 - DEC F4 & SEP FS - REFUELING --- JUAB CITIES 4

S.07.F.3007 : VARI 3US PROBLEMS WITH VALVE LIMITOROUE OPERATORS b

                                                  --- DRESDEN 2 & 3 - DEC 70 - JAN F1                                                         .

s.0s.t.0055 : DIRT IN SERVICE WATER CNECE VALVE c) j ,

                                                  --- STSTER CREEx                      -

MAT r5 - 390 mWE i -

I PAGE 4 3.08.C.3064 : CROSS s THREADES PIPE S LOOSE WALVE 90LTS - LEAKS

                              --- BRUNSWICC 2 - JUL 75 - Ft POWER S.07.F.0006 : CNEtt VALVE NAD MISSING DISK AND P3P*ET                                                                                    I
                              --- 3UAD-CITIES 1 - JUL F4 - SNUTD3WN B.09.F.0007 :

INSTRUMENT BAD SEAT CNECK WALVES NAD dEAK SPRINGS, PITTED POPPETS AND $3

                             --- 3UAD-CITIES 2 - APR 75 - COLD SN3fDOWN

9.09.G.3147

DAM 43ED SEAT ON INSTRUMENT N2 VALVE

                             --- PEACM 30TTOM 2 - MAT F5 - 100E P3WER

() ' t a.11.4.0057 : STUCC DIESEL FUEL SALL CNECK VALVE

                             --- WERMONT TANKEE - JUL F4
                                                                                                                                                )
)

9.11.A.0103 : STUCK DIESEL AIR CNECK VALVE

                            --- 3RUNSWICK 2 - DEC 75 - 340 MWE S

B.14.8.001F : SECONDARY C01TAIN1ENT VIOLATIONS

                            --- DRESDEN 2 8 3 - OCT FI - MAR F2 1

5.14.8.0037 : CNECK VALVES LEAKED

                            --- DRESDEN 3 - J AN F3 8.14.8.0056 :       C3NT AINMENT CNECK VALVE MISALIGN4E4T MONTICELLO - NOV F3 8.14.8.0059 : LEAKING VACUJM SREAKER PENETRATIONS - INADVERTENT RELEASE
                           --- PEACM 90TTOM 2 - DEC F3 (POWER ESCALATION TESTING)                                                            gj 8.14.8.3072 : VACU3M BREAKER LEAKAGE
                           --- 3YSTER CAEEE 1 - APR F4 - 442 9WE

! () 3 3.14.8.3081 : DEFORMATION

                          ---              3F CHECK VALVE RUSSER SEAL

• PEACM BOTTOM 3 - MAR F5 - 1005 P3WER gg 8.14.0.0088 : REVERSE FLOW CHECCS REMOVED FROM SGTS - M3 TOR OVERLOAD

                          --- COOPER - MAY 75 - COLD $NUTD0d4 J

a.14.s.3115 : CNECK VALVE PISTON FOUND NUNG UP ) --- PEACH 80TT3M 3 - OCT F5 - SSI PodER s.15.A.0007 : TNIN WALLED WALVES - BWR$ IN GENERAL - 19F0 ((P.15.33) 9.15.A.0037 : NITROGEN ACCJ4ULATOR WALVE LEAKED s a

                          --- SRUNSWICC 2 - DEC 75 - SSE POWER

! 3.16.C.0001 : J QUESTION 04 SUITABILITY OF CERTAIN PIPING AND VALVES i --- OfSTER CREEK 1 - JUL 59 j S.16.C.0036 : ' PROCEDURAL PROBLE1 - LPCI VALVE MOTOR SURNED OUT DURI4G T!ST. J

                         --- VERMONT FANKEE - MAY F3 s.1$.C.0099 :                                                                                                                   e PROCEDURAL LEFT    OPEN    PROBLEM - EXCESS FLOW CNECK VALVE'S SYPASS VALWES
                         --- PEACM 10TTOM 2 & 3 - NOV F4 - 1035 8 $0E POWER fl
     ............. ALL DONE. PRESS CRETURN> (EY TO G3 T3 MENU ,,,,,,,,,,,,,,,,,7 G) l a                             v.                                          ,         - - - - - - _ - - _           _ _ - .

e

                                                                                       . s SAGE 5 i.

THERE 4RE 123 ARTICLE (S) TO SE DISPLATES: 19-4Ug-1984 13: 37: 3 7 l BWR CNECK V4LVES, 19F6 - 1980 3.04.3.3047 : CRD i RETURN LINE ISOL4T!0N VALVE M43 WORM SEAT, RJST ON PISTON

                                        --- 3RESDEg 3 - OCT T6 - REFdELING SMUTDOWN 3.04.9.0114 :

4 CONTROL R00 3ID NOT L4TCM - DIRECTIONAL V4LVE FAILED

                                        --- BROWNS FERRY 2 - JUN B0 - 401 POWER                                                                                                                              '

i 5.04.5.0117 : WEST SDV NEADERS DID NOT DRAIN AFTER M4NU4L SCRAN

                                        --- 3RESDEN 5 - JUL 80 - SHUTDOWN 4.04.s.3145 :              CONTROL ROD 3 RIVE SYSTEM 94LFUNCTIONS 3
                                        --- 3RUNSWICK

{ 1 - AUG 80 - SJ0 CRITIC 4L (STARTUP) --* 3RWNSWICK 2 bb

                                       - FES 81 - 1.53 P3WER --- OTSTER CREEK -N3V 80 - 63X P0 DER s.05.C.0257 :             THRE4DED LOC (ING DEVICES 3M VALVES, PUMPS, W4LVE OPER4T3R$

4 FAILED

                                       -** SWR'S IN GENERAL - M4R 83 ((P.089.2933) 5.05.C.3261 :                                                                                                                                                                                   .

l ADS 41R SUPPLT ACCUMULAT3R CNECK W4LWES LEAKED

                                       --- NATCH 2 4PR 93 - COLD SNUTD0dN 1'

3.35.C.3266 : ADS 4 AIR J j SUPPLT ACCU 1ULAT3R CNECK WALVES LEAKED ' --- COOPER - APR 40 - REFUELING j > a.06.E.0054 : RUST ON FW C4ECK WALVE SE4TS ]' --- PEACM BOTTOM 2 - MAR T4 - SMUTD0dN ji 3.06.E.0035 : FWDRESDEN

                                      ---    CHECK V4LWE 2 - MAR LE4KAGE           - EXCESSIVE DISC-73-SEAT CLEARENCC 76 - REFUELING SMUTDodN g, ;

s.06.E.0060 : '

,                                     FW CNECK VALWES LEAKED - WITON 0-RINES DETERIOR ATED I
                                      --- 3U40-CITIES 1 - JAN TS - REFUELING SNUTDOWN                                                                                                                     g) ;

i 3.06.E.0064 : DIRT IN FW CNECK VALVES, IMPROPER DISC-TO-SEAT CLEAR 4MCE, ' { EECESSIVE LE4KAGE - TESTING METHODS dERE INADEGU4TE

                                     --- DRESDEN 3 - SEPT & OCT 76 - REFUELING SNUTDOWN 9.06.E.0065 :            FW $4MPLE PR30ES LODGED IN NPCI E FW CNECK VALVES, FL4NSE I'                                    GASKET BLOWN                                                                                                                                                          #

76 -COOPER SHUTDOW1 - J4N FF - APPROMIRATELY 762 POWERJ SRUNSWICK 2 - FEB

       ~   3.05.E.3076 :            'dORN FW VALVE SEAT / DISC ASSE19LT PINS, DEFORMED SEAT RIMGS

{ i --- DRESDEN 2 - OCT ?? - REFUELING SNUTD0dN

s e.06.E.0086

FW CNECK VALWES SEATS MACNINED, SE4T SEAL 0-RINGS CH44GED TO K4LREI

. --- 3UAD-CITIES 2 - SEPT F6 - REFUELING J 9.06.E.0093 : W3RN SEAT / DISC ASSEM0LY PINS IN FW CNECKS I

                                    --- DRESDEN 3 - MAR 78 - REFUELING                                                                                                                                   .)

j 9.06.E.0097

• WDRN SEATS 410 RINGS IN Fd CNECK WALVES ,

       ~
                                    --- MILLSTONE 1 - MAR 78 - REFUELING                                                                                                                                 OP i           e.06.E.0107 : DIRT             04 FW C4ECK VALVE SEnf ou40-CITIES f - JAN FP - REFUELIVG                                                                                                                          f)

. 8.05.E.0109 : WORN SUSHINGS - FW CNECK WALWES LE4KED 4 w -,. - -_u-- - - - - - - - - - - - - - - -

I I dVSI 9 EunNSA333 8 - Vdu le - 53JA31IhS 6*C9*3*O!!O

• ADUN ---

JM 3M333 AvlA3 6nSN!hSS , tunNSMI3) 2 = WVA 24 - m3An31!hS L P 8*09*3*CL12 3 AM --- 3N333 AV173$ 13v336 - SEVAS u3dVI53e* 53V1 u3d1V334 N111510N3 t - roN3 26 - assn 31 INS e*09*3*0119 t JM 3N333 AV1A3 MIMS DIN 3CABW 13V3 = NISM 3ONAVIMh3N A 13Nd g

                       --- EunNSM133 2 - vnS la - 6tS 40M3N E*0S*3*OLZJ 8 JA 3N332 AV1A3 13VE - A0uh 53V1/ TIS 3 VSS3ht11 DINS                                                            )
                       --- 853563N S - J35 80 - 53dn31 INS 8*05*3*01SL 8 JA   ---13V330 !klO 10u0$ nNG3J33134 - 3N333 AV1A3 e0hN31 33V1 AVI134 g

A1uW0NA AVN333 - fnN] fO - CSI dCM35  ! i E*0$*3*CLCS 3 AA 3M333 AV1A3 53V1/0153 VSSIM811 d!hS ROWN g

                       --- ON3$03N C - kVW 80 - 53Jn31 INS t*09*3*01VL : MCuN 53V1* eluA IN AP 3N333 AV1A3$
                       --- :nV0-2III35 1 - 53dA 80 - 83An31 INS SNR100PN E*0J"V*0058 8 e3NA 0!53 RVSN3W* INduod3N SEVAINS 1h 53I3 3RNVASA AV1A3
                       --- k0N1133110 - C31 JJ - u3Jn311NE SHn10CMN 8*OJ'V*0054 8 WORDN 5391IN$ $nuJV33$ lk 53I3 813VN AV1A33                                                                     .g
                      --- 3OOd3m - 031 42 - v34c31 INS SNc100nN s'04'V*C092       u2I3 inuEINE 511vh 3XMVASA 3N333 AV1A3 13V334 - A1Vdd3u tb033                                               gq tnV0-3III3$ 2 - 53d1 J9 - u3JA31IMS E*OJ'v*00JL 8 ucnSN VNO 3uV333$ IXNVn$A AV1A3 S3VAS                                                              ,
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                     --- 300435 - tdu 48 - 53Jn31 INS f*OJ'V*004$ t tluA ON inugIN3 3NHV031 AV1A3 SEVAS                                                                             gg
                     --- WONAI33110 - C31 JS - 53JA31!hS 5*OJ V*C029 8 u3IJ AV143 DIN SNIW436                                                                                            f
                     --- tsoMNS dluuA L - 033 AS - s3A.r311NS 8 04 V*0090 8 O!ulA        AV1A3 IN13uNV1$

tnva-31113$ L - fVN 4( - 53Jn31IND 8*OJ*V*C005 8 10053 AV1A3 01$3 0103330 53I3 !nuGIN3 3XWVnlA lIN3 a10Mb. unalps1 GIS3

                     --- MV13H 2 - rnhE 46 - CI dCM3W e*OJ'V*009$

• 03J331IA3 10ugnE SAI13N* s005M SIVAS - 313VN tnedlA VNG 33M1051 AV1A3$ 13V330

                     --- 3OOd3u - Vdu 26 - u3Jn31 INS                                                                  i         e E*OJ*V*006v : 5313 anusIN3 sVI1]o 10 savul - s011 10oS3e IN 513VN 33NVnsI 3N333 AV1A3                                                                                                  9
                    --- GunhSM]3) ! - NOA 26 - SN0100Ph E*OJ*V*COCS 8---3333 luId - 3M333       AT1A3 1C33 95033ks AIGuV130 310$30                                                    )(

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                                --- NATCM 2 - MAY 80 - 99E P3WER 2

I 9.07.D.0236 : RNR --- CMECK VALVE DID NOT SEAT WERMONT TANKEE - SEPT 80 - 898 P3WER s.07.D.0238 : INAD!eUATE

                                ---             SGPPORT - LPCI DRAIN LINE WELD LEAE                                  '#

DRESDEN 2 - MAT 03 - SMUfDsJM 3.07.D.3259 : CNECC VALVE STUCE - PLANT MODIFICATI3N INSTALLED

                               --- BRUNSWICC 2 - NOW 80 - 6tt P0 DER 9.07.E.0076 : HPCI  ---

STEAM C1ECK VALVE GASKET FAILED 3UAD-CITIES 1 - JAN 76 - REFUELI16 $NJTDOWN 5.0F.E.0081 : 4 IMPR3PER MOU1 TING - CORROSION AND PITTING 04 MPCI VALVE SEAT

                               --- DRESDEN 2 - MAR 76 - SMUTDOWN                                                       #

i l e.07.E.3090 : BINDING IN M*CI TURBINE EENAUST WALVES - RUPTURE DISC SL0d4 # t --- 3R0WNS FERRY 3 - AUG F6 - 3I P3WER j t.0F.E.0095 : LOOSE RETAINING MUTS, SROCEN RETAINI16 STUSS DISLODGED DISC IN I MPCI TUR3INE EXMAUST CMECC VALVES .! BRUNSWICC 1 8 2 - OCT F4 - COLD SMUTD3WN 1 e a.07.E.0113 : MPCI TUReINE ENNAUST WALVE DISC STUD FAILED, MISSING PARTS

                              --- COOPER - APR FF - COLD $40TD0d1 i

I 5.07.E.3114 : NPCI FRACTURED FLOW OSCILLATIONS AT LOJ SPEED - CNECK WALVE DISC NI1GE

                             --- 3RuMSWICc 1 - MAY 77 - C3LD $NuTDOWN                                              O 4

8.07.E.3121 : DIRTY EENAUST VALVE SEATS I

                             --- MONTICELLO - SEP FF - REFUELING SMUTDOWN j           5.07.E.3122 : R3UGN AND SC4ATCHED WALVE SEATS                                           .

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                            --- stuMSWICc 2 - MAR Ts - $NUTD341 8.07.E.0139 : MECMANICAL 11TERFERENCE IN CNECK VALVE
                            --- 3ROWNS FERRY 2 - MAR FS

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3.0F.E.0152 : GLAND SEAL CNECK WALVE LEAKED l

                            --- PEACM SOTTOM 2 - OCT F8 - $1X PodER i
   ~      8.07.E.3155 : DIRT ON TUROINE EINAUST VALVE SEATS i

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                            -*- 10NTICEL*.0 - 3CT F4 - REFUELING                                                    #

i s.07.E.0141 : CNECC VALVE 34!KETED SEAT FAILED

                           --- OUAD-CITIES 1 - JAN FF - REFUELI1G                                    '              8 5.07.E.0209 : TORUS SUCTIO1 VALVES FAILED DUE TO STEAM LEAK FROM A CNECC VALVE
                           --- NATCM 2 - MAY 00 - 99I POWER N

s.07.E.3213 : FOREIGN MATEttal SETdEEN CHECK VALVE DODY & SEAT . ] sROWNS FERRY 2 - SEP 90 - REFUELING SNUTDOWN f) 1 i 3.07.F.0037 : UNIO 4 BETWEE4 IS3LATION FLOW CMECC VALVE AND STEAM LI4E F.0W SENS3R WAS CR355-TMREADED

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