ML18019B089

Petition Requesting Institution of Proceedings Per 10CFR2.206,requiring Util to Respond to Show Cause Order Due to Failure to Meet Required Stds in Areas of Emergency Planning,Plant Safety,Security & Personnel Stress

ML18019B089
Person / Time
Site:	Harris
Issue date:	07/02/1986
From:	Eddleman W, Julie Hughes, Katz S COALITION FOR ALTERNATIVES TO SHEARON HARRIS
To:	Office of Nuclear Reactor Regulation
Shared Package
ML18019B088	List: ML18019B089
References
2.206, NUDOCS 8607160412
Download: ML18019B089 (166)

Text

July 2, 1986 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE DIRECTOR OFFICE OF NUCLEAR REACTOR REGULATION In the Matter of Carolina Power&Light Co.and North Carolina Eastern Municipal Power Agency Docket No.50-400 (10 C.F.R.8 2.206)5i (Shearon Harris Nuclear Power Plant)REQUEST FOR INSTITUTION OF PROCEEDINGS PURSUANT TO 10 CFR 2.206pgo7 koMI~EDO-001906 OUTLINE CONTENTS I.II.III.IV.V.VI.VII.VIII.Introduction Statement of Authorization to Represent Persons, Organizations and Interests Standard Pursuant to 10 CFR 2.206/2.202.

Emergency Plannning/Preparedness Arguments Quality Assurance Program Arguments NEPA Psychological Stress Arguments Summary of Relief Sought Apppendix A.Organizational Document CASH B.Affadavit Ted Outwater C.Chatham County Commissioners Resolution 27 Nay 1986 D.Affadavits:

1.Dan Frazier 2.Barbara Keyworth/David Richardson 3.Ruth Thomas 4.Nitchell 6 Kay Riley 5.Clair a Edward Thomas 6.Anne Greenlaw 7.Rada Greenlaw E.Comment: Emergency Planning Zone: Kenneth G.Sexton, Ph.D.(June 30, 1986)F.Letter Patricia Niriello (January 1, 1986)

I.INTRODUCTION The petitioners request that Nr.Harold R.Denton;Director of Nuclear Reactor Regulations require CPaL to respond to a, show cause order pursuant to 10 CFR 2.202.In conforming with the requirements of 10 CFR 2.206, the petitioners will demonstrate that.CP&L, by acts or ommission, has failed to meet the applicable standards required by 10 CFR et.al..Petitioners will address the following issues: Emergency Planning, Plant Safety, Security, and Psychological Stress.Joseph Hughes and Steven Katz, are authorized by the Coalition for Alternatives to Shearon Harris, Calvin Regan, et.al., and Patricia hliriello, to assert the interest of the organizations'embership (1), (which includes CASH members residing in Chatham, Nake, Harnett, Lee, Durham, and Orange counties the principal population concentration of the organization lies within a 15 mile radius of Shearon Harris Nuclear Power Plant.See: Appendix A for organizational material.)

Calvin Regan, et.al., (2)(see petition for CASH's representation of residence of persons living within the five mile zone at Appendix B), Patricia Niriello (3), (see documentation of f1s.lliriello's request for CASH's representation in these proceedings), and the interests of Joseph Hughes and Steven P.Katz (4).(Joseph Hughes and Steven Katz are CASH members and are responsible for developing legal strategy, and reside in Durham and Orange Counties respectively.)

On June 9, 1986, CASH filed documents with the NRC: first, a petition for leave to intervene pursuant to 10 CFR 2.714 (a)and 2.715 (a).A document in the form of a motion to State the Immediate Effectiveness of the Final Licensing Board Decision was filed on June 9, 1986 and this motion was joined and signed by Wells Eddelman, pro se.The motion complied with the procedural requirements of 10 CFR 2.788 and 10 CFR 2.764.In light of these filings, CASH's viability as a multicounty organization, CASH's representation of its membership, Nr Regan et.al., and tls..Niriello, the petitioner clearly has the requisite interest to assert the following arguments.

III.Standard Under 10 CFR 2.206 to Initiate a Proceeding Section 2.206 provides a mechanism whereby members of the public may: 1.Request initiation of an enforcement action to modify, suspend or revoke a construction or operation licenses held by a utility;or;2.for other such action as may be proper.The Director of the appropriate NRC office is vested with the authority to institute action pursuant to 10 CFR 2.202 Show cause order.A show cause order, 10 CFR 2.202, should be issued by the Director where substantial health or safety issues have been raised.Consolidated Edison CL1758, 2NRC 173, 175 (1985).Additional health and safety requirements are set out in 10 CFR, and are relevent in determining whether adequate measures have been taken by the utility to protect public health and safety.IV.Emergency Preparedness and Planning A.Factual Background On May 27, 1986, the Chatham County Commissioners passed a resolution rescinding prior approval of the Emergency Management Plan.(See: Appendix).The operative language is as follows: Now, therefore, be it resolved that the Chatham County Commissioners hereby rescind all prior approvals of the Shearon Harris Emergency Response Plan pending further critical study of the unresolved issues.As a general proposition, local governmental entities are an integral part of emergency planning.See: 10 CFR 50.47.(b)(1);(primary responsibility for emergency response...by state and local organizations within the emergency planning zone (are)assigned, and specifically established and each organization has staff to respond and augment its initial response on a continuing basis).It is clear that Chatham County's emergency preparedness, as of this date, is fatally deficient.

The Commissioners have rescinded their agreement to participate in the plan.Supporting organizations will not be staffed.Without staff mere notice of a radiological emergency occurrence would result in chaos.In short, there is no means of assuring that the population of Chatham County would be protected by any organization in the event of a radiological emergency.

I 0 B.Adequacy of the EMP There can be no question that emergency preparedness, particularly in Chatham County is inadequate and fails to assure that any plan could be implemented.

10 CFR 50.47 (a)(2).Petitioner notes that FEMA found the E.M.P.adequate, as of May 1985.However, the FEMA finding has been mooted, by the Chatham County's rescision of May 27, 1986.C.Requirement of Reasonable Assurance It is clear that 10 CFR 50.47 (a)requires a finding made by the NRC that there be reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency.

FEMA did find that emergency planning was adequate in may 1985.Then presumption of adequacy and implementation is rebutted, due to the effect of the Chatham pullout.The EMP without Chatham County's participation cannot satisfy the requirements of 10 CFR 50.47 (b)(116)and 10 CFR part 50.(Supplemental documents will be forwarded to the Director analyzing the sixteen requirements for an EMP.)D.One Year Test Standard Emergency Preparedness New plants are required, to conduct a full scale exercise which tests the emergency plan.That plan is to be conducted within one year before issuance of the first, full power operating license (10 CFR Part 50, Appendix E Section Fl).It is understood that the emergency preparedness exercises are part of the operational inspection process and are not required for any initial licensing decision, 10 CFR 50.47(a)(2) however the language requiring a full scale exercise to be held within one year before full power operation.

Union of Concerned Scientists vs.NRC, 735 Fzd 1437 (D.C.Cir.1984), citing 10 CFR 50.47(a)(2).

Here, the FEMA approval of the EMP was made in May of 1985.Chatham County participated in that test.Any plan which does not include Chatham County is clearly not the plan which was tested in May 1985.The plain meaning of, 10 CFR Part 50, App.E.(F)(1), requires a test of the plan one year prior to granting of an operation license.The Director should withhold granting of any operation license until such matter is resolved.

l I The petitioners'inal argument runs to the implementation phase of the ENP.10 CFR Part 50 Appendix E.The applicable requirements of Plant Staffing assignments have not been clearly communicated to the operations staff.The requirements for"Activation of Emergency Organization" were tested during a given event occurring 28 June 1986.A preliminary analysis of the manner in which information is disseminated from the plant in the event of a'iren system transmission, clearly there is a lack of preparedness with respect to activation of the notification systemboth onsight and within the affected communities.

The events of 28 June 1986 are summarized as follows.See: affadavits at Appendix D.An alarm siren was activated on June.1986 at 1:55 a.m.Numerous persons were awakened during the siren transmission.

Persons living 2 miles north of the siren awoke and attempted to call various state and local authorities, and also called Shearon Harris Nuclear Power Plant.(See: affadavit of Barbara Keyworth and David Richardson).

The Chatham County Sheriff Department dispatcher had not been informed by CPGL of the siren's purpose.The dispatcher stated that she had received other pohone calls from concerned residents of Chatham County.Calls were made to Shearon Harris Nuclear Power Plant.The proffered explanation was, that a shift whistle sounding at 2:00 a.m.had roused persons eleven miles from the plant.(See affadavit of Keyworth and Richardson).

Confusion continued as calls were'made to the N.C.Highway Patrol which resulted in a particularly uninformed and condescinding response.Nr.Nac Harris, media manager CPSL, released a media piece which stated that vandals had tampered with the siren box setting the device off.This media release is contradicted by petitioners'ffidavits which tend to prove that the siren which was allegedly tampered with had no visible signs of forceful tampering with either the security locks or the siren itself.(See affidavits of Frazier, Keyworth, Richardson and Thomas).A continuing investigation of this matter continues.

However, a number of inferences are readily apparent.First, security, if one chooses to believe CP&L's version of the incident, at the siren locations is not adequately provided.If vandals were able to set sirens off at will, the underlying reliability and value of the emergency warning system would be rendered useless.Second, there is apparently no method to secure information upon the activation of an emergency siren.

I C 0 Clearly 10 CFR Part 50 App.E (c)requires the existence of message authentication scheme which includes notification of local emergency officials about unusual events, alerts, site area emergencies, and general emergencies.

Note that 10 CFR Part 50 App.E.(D)(3), states that";..where there is substantial time available for state and local officials to make a judgement whether or not to activate the public notification system.'here there is a decision to act'ivate.

..the state and local officials"will make the determination." This incident implicates the unrefined information gathering and dissemination process which is the central thrust of any emergency notification scheme.F.Conclusion and Requested Action: The ENP approved by FERA in Nay of 1985 is no longer viable.See Appendix E.It no longer provides for participation by Chatham County.The EHP has been flawed by an incident involving an emergency siren which sounded and residents of the EPZ were unable to ascertain definitive information with respect to the nature of the alarm or what action should be taken (evacuation, etc.;incidentally--no person from whom affidavits were taken turned to the emergency broadcast channel-petitioner will supplement this document as information becomes available).

Finally, 10 CFR 50.47(d)provides that a license authorizing fuel loading and/or low power operation may be issued after a finding that the state of emergency preparedness provides reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency.

This standard has not been met.Therefore, petitioner moves that: 1.The Director should issue a 10 CFR 2.202 show cause order upon CP&L to demonstrate why CP&L should not be required to proceed with a complete Preliminary Safety Analysis pursuant to 10 CFR Part 50 App.E.II (in light of the Chatham County pull-out).

2.The Director should issue a 10 CFR 2.202 upon CP&L to demonstrate why CP&L should not be required to comply with the requirements of 10 CFR Part 50 App.E.III{in light of Activation of Emergency Notification System).3.The Director should immediately revoke present or prospective authorization, for fuel loading and five percent testing of the Shearon Harris Nuclear Power Plant (lack of reasonable assurance that adequate measures can and will be take in the event of a radiological emergency due to Chatham County's pull-out).

4.That the Director proceed in a hearing upon the substantive issues raised by the petitioner in this and various pleadings filed with the NRC (pursuant to section 189 of the Atomic Energy Act).V.Former CP&L Employee Investigation/Document Falsification/CP&L Quality Assurance On January 1, 1986, Ms.Patty Miriello, a former CP&L worker at Shearon Harris and Brunswick nuclear reactors wrote to the presiding judge, James Kelley, Chairman of the ASLB panel in Docket 50400 OL alleging falsification of radiation exposure records and questionable practices relating to health physics and requested that her identity remain confidential.

See Appendix F.The Chairman, however, ruled, pursuant to 10 CFR 2.780(b)that the allegations were to be treated as ex parte communications and disclosed the information to all parties in the case, including the Applicants.

Although the NRC Office of Investigations (OI)has had documented evidence of Ms.Miriello's contentions since September 1985, the OI has yet to do a personal interview the the alleger.Moreover, the NRC OI has yet to issue a report of its investigation, which goes to the heart of the question of the Applicant's competence and integrity in operating the proposed Shearon Harris Plant.l.As a worker exposed to radiation of the Applicant's nuclear reactors, the facts which have been brought forward by Ms.Miriello create serious close questions which would implicate the effectiveness of the Applicant's proposed, radiation protection program for its employees.

Moreover, the assertions which Ms.Miriello make, if substantiated by the Office of Investigation report which has yet to be completed, would result in a finding by the Commission that the Applicant's request for an operating license"may be revoked suspended or modified, in whole or part, for any material false statement of fact requireed of the Applicant." (10 CFR 50.100)Miriello, a former employee of CP&L, alledged in September of 1985 that documents were falsified by the applicant.

The OI has yet to complete this investigation.

Among other allegations which have not been resolved, Miriello has been unable to obtain her complete record from the applicant and thus has been precluded from seeking positions within that field.Aside from the interest in freedom to pursue gainful employment, the applicant may be in violation of 10 CFR 50.100 (material false statements of fact), 10 CFR 0.735039(c)(disclosure of confidential information by the applicant), and a substantial possibility that the applicant may not have an adequate radiation protection program.All these issues may in combination or in part, amount to a substantial fundamental flaw in the final decision of the Licensing Board's decision.Moreover, petitioners allege that the Applicant may have a defective radiation protection program regarding the requirement for maintaining records of employee radiation exposure under 10 CFR 20.401.With regard to its former employee, Ms.Miriello, the Applicant may have violated 10 CFR 20.601 concerning falsification of employee monitoring records according to the attached affadavit.

Moreover, when Ms.Miriello left employment with the Applicant, she was not provided with accurate exposure data as required under 10 CFR 20.408.In each of these instances, improper recordkeeping in the Applicant's radiation protection program could constitute adequate grounds for withholding or revoking the Applicant's proposed operat'ing license.Beyond concerns about the Applicant's radiation protection program, Ms.Miriello also has provided the NRC with documentary evidence of improper inservice ultrasonic inspectiions of the large reactor coolant line welds as part of the Applicant's quality assurance program.Attachments fl and 02, which are both five pages long, show a discrepancy on the fourth page for the coolant line welds for numbers 09 and 412.As Ms.Miriello notes in her letter to Judge Kelley of January 1, 1986,"two level III Nuclear Energy Services NDE inspectors argued over these ultrasonic results.They had conflicting opinions." When the new page four was revised as shown in Attachment 42,"note that the mention of a weld or repair weld was eliminated from page 4 of the original Mel Perry (NES corporate inspector) turned in.""Also removed was the listings of indications in this weld, referring specifically to indications 09 and 412." According to Ms.Miriello's investigation, these design flaws in the Shearon Harris core coolant line are violations of the ASME Boiler and Pressure Vessel Code,Section XI, Articvle IWB3000,"Acceptance Standards for Flaw Indications", as quoted from the 1980 edition of the Shearon Harris PreService Inspection Manual.According to the information which Ms.Miriello has observed and obtained, approximately 10%of the welds in the inservice inspection program at the Shearon Harris plant are defective and improperly documented.

These inservice inspection records were altered and changed

without following the proper HRC procedure for record revisions on pipe welds.VTe feel that these violations of NRC regulations in the Carolina Power and Light Company's Quality Assurance Program are sufficient grounds for withholding an operating license until these critical plant safety violations are investigated.

Petitioner moves the Director to proceed in a 10 CFR 2.202 sho cause proceeding, and 189 hearing, to consider questions of material fact raised by this argument.VI.Psychological Stress Argument A.It is national policy that each federal agency shall utilize a"systematic, interdisciplinary approach which will insure the integrated use of natural and social sciences and environmental design arts", in order to assure that governmental action which affects the health and safety of the persons within a particular zone will be adequately protected.

See: 42 U.S.C.4331(2)(a).

In order to implement this policy which the proposed action affects public health and safety, as a factor in determining whether the federal action significantl'y affects the human environment.

In Pe~o le Against Nuclear Energy vs.V.S.N.R.C., 678 F2d 222 (D.C.Cir 1982), the Circuit Court was called to consider a novel health and safety issue, in light of the National Environmental Policy Act,'2 USC S4321, et.seq., and the Atomic Energy Act, 42 USC s 2133.The issue ran to the possibility that renewed operation of the plant at TNI would cause severe psychological distress to persons living within the vicinity of the reactor.The operation of the reactor would harm the stability, cohesiveness and well-being of the communities within the vicinity of the reactor.678 F.2d at 226-227.The petitioners in PANE claimed that citizens had lost confidence that responsible institutions could function effectively during a crisis.That the area was becoming an undesirable location for residents and businesses; and, that the operation of the reactor was causing permanent damage to the economic and social health of the community were also alleged.Id.The Court in PANE held that the petitioners had alleged claimed within the meaning of the NEPA, and were allegations which rise to the level of environmental effects.B.The central question in evaluating issues of psychological stress are the potential that particular governmental action may effect health.Language in the case supports the notion that there are occasions for considering when psychological stress is to be considered as a factor in evaluating the propriety of governmental action by a government agency.First, it is clear that 10

congress intended to include psychological stress as an element of the calclus for determining what effect a governmental action has on'health'78 F2d at 230.It is equally clear that the severity of psychological harm, and the cognizability of that harm under the NEPA will not be satisfied by"mere dissatisfaction arising from social opinions, economic concerns, or political disagreements with agency policies".

Id.What does not seem clear is the extent to which psychological str'ess will preclude governmental'action in light of the recent disaster at Chernobyl, and the recent failures of CP&L to adequately inform the public of thhe nature of an early morning siren which left numerous residents of the Emergency Management zone wondering whether to evacuate, and subsequently wondering whether the plan as designed could adequately assure the health and safety of their person in the event of a radiological emergency.

In Netro~olitan Edison v Pep~le Against Nuclear Energy 460 US 766, 75 Led2d 534, 103 S.Ct.1556 (1983), the court held that the NEPA does not,require the NRC to consider whether the risk of an accident at a nuclear power plant may cause harm to the psychological health and community well being of residents of the surrounding area.The Supreme Court in so holding did not affirmatively prohibit the consideration of psychological stress by the NRC in their determination of whether to order.an Environmental Impact Statement or investigation.

C.Petitioners argument begins with the following premises: that the Commission must comply with the NEPA before it takes'major federal action'.That such"major federal action'reates a statutory responsibility with the NEPA.A'major federal action', includes, but is not limited to, new and continuing activities, including projects and programs, entirely or partially fininished, conducted, regulated or approved by federal agencies.40 CFR s15.08.18(a).

See also, 678 F.2d at note 14.(direct and immediate effect of psychological health or community well being).It is clear that responsibility to assure that nuclear power plants will operate without endangering the health and safety of the public lies with the Commission.

Where the Commission takes'major federal action'uch action is continually reviewable in accordance with the standards set out in the NEPA.The Commission is required to prepare a Supplemental Environmental Impact Statement upon the occurrence of either of the following conditions:

first, where the agency makes substantial changes in the proposed actions that are relevant to safety concerns;and, second, where there are significant new circumstances or information which are relevent to environmental concerns and bears to proposed action.40 CFR 1502.9(c)(1).

The petitioner argues that three significant new circumstances have developed within the time of the FENA approval of the ENP the and this date.(as will be argued later the Chernobyl accident and the false siren, 28 June 1986, in Chatham County, and the Chatham County pull-out are such significant new circumstances).

D.The factors employed in determining whether an event rises to the level of a significant new circumstance are;(a)the environmental significance of the new information;(b)the probable accuracy of the information;(c)the degree of care the agency used in considering the new information;(d)the degree to which the agency supported its decision with additional data.Harm S~rings Dam Task Force v.Gribble, 621 F2d 1017 (9th Cir.1980).These factors are relevent here to the degree that the Commission is required to take a'hard look't events which may rise to the level of significant new circumstances.

Furthermore, in reviewing environmental allegations the Commission should take a'hard look'here significant new circumstances are asserted.Alleged facts should be evaluated by the Commission, in a complete and comprehensive manner.See: 678 F2d 234, Note 20.1.The twin disasters of Three Mile Island and Chernobyl have raised compelling questions with respect to the dispersal of radiation.

The Director of Nuclear Reactor Regulation.

should take a'hard look't NUREG-CR-2239 and NUREG-CR-0956.

These documents concern data'with respect to severe accidents (of the Chernobyl and TMI type).The issue concerns the quantities of radioactive material which affect persons.The NRC's failure to consider as part of its environmental assesment NUREG-CR-2239 and 0956, which is current and accurate information.

In light of the particular argument NUREG 2239/0956, and the general argument that scientific understanding has been significantly advanced in light of TMI and Chernobyl (with respect to the dispersal of radiation), notions concerning the adequancy of a ten mile emergeny planning zone may be inadequate to protect the health and safety of those living around the Shearon Harris Nuclear Power Plant.Because the petitioner alleges a new, significant, environmental circumstance, supported by some particular data, it is moved, pursuant to 10 CFR 2.206, that the Director take action consistent with this new information and conduct an Environmental Impact Statement prior to any affirmative licensing action concerning Shearon Harris Nuclear Power Plant.The petitioner moves th'at the decision of the Licensing Board be stayed pending completion of the Environmental Impact.Statement.

Wherefore, the undersigned, individually, and in their representative capacity prays that you institute a proceeding pursuant to 10 CFR 2.202, based upon the moved issues raised herein.2 July 1986 Respectfully submitted, eph T.Hughes, r.04 W.Chapel Hil St.Durham<N.C.(919)98 3818 Steven P.Katz 604 W.Chapel Hill St.Durham, N.C.(919)682-3818 Wells Eddleman, pro se Durham, N.C.(919)688-0076 13

AAPPENOlX DRAFT MOTION CONSTITUTING THE COALITION FOR ALTERNATIVES TO SHEARON HARRIS (C.A.S.H.), CREATING AN INTERIM STEERING COMMITTEE, AND ESTABLISHING TNO THIRDS MAJORITY VOTE AS BASIS FOR DECISIONS.

P Whereas the impending loading and operation of the Shearon Harris Nuclear Power Plant is a threat to our health, safety, and economic well-being and necessitates quick, creative, and concerted collective action both within and across our communities this Emergency Regional Assembly hereby constitutes itself as the Coalition for Alternatives to Shearon Harris (C.A.S.H.), membership in which is open to all individuals and groups which endorse the Apex Declaration.

Further, until the convening of a second Regional Assembly, it creates an Interim Steering Committee to guide the Coalition's growth and activities to be comprised of representatives of those working groups which may be established to further the Coalition's aims and objectives, and representatives of those local organizations which may be created t'o implement them.Further, it establishes consensus as the ideal to be strived for in Coalition and Steering Committee'decisions and specifies that in the event consensus is unattainable decisions shall be based on two-thirds majority vote.

Affidavit My name is Ted Outwater.On Saturday, June 7, 1986, I contacted the following~~~~~s t-residents living within the Five Mile Zone around the Shearon Harris Nuclear Power Plant and obtained their signatures on the attached document.I am a member of the Coalition for Alternatives to Shearon Harris (C.A.S.H.), serve on the C.A.S.H.Steering Committee, and work out of our Durham Office at 604 W.Chapel Hill St.Durham N.C.27701.Ted Outwater State of North Carolina, Durham County I, Julia Borbely-Brown, a notary public, due hereby certify that Ted Outwater the affiant personally appeared before me this day and acknowledged the due execution of the foregoing affidavit.

Winess my hand and notarial seal, this the 8th.day of June, 1986 tary public State of North Carolina, Durham County Ny commission expires: m~i>/9fg AppfNDIX

We would like;the Coalition for Alternatives to Shearon Harris (C.A.S.H.)

to represent us and to intervene on our behalf before the Nuclear Regulatory Commission in the matter of licensing the Shearon Harris Nuclear Power Plant.We do not believe that the interests of the residents living within the Five Mile Zone around the Harris plant have ever been recognized or represented

.NAME ADDRESS DO YOU LIVE INSIDE THE FIVE MILE ZONE?-~cf.<~3 8J 7 7w./, Box 3s j Aj~ldi//hlC>><4~(si J+tv'P-gg.Lghr

)<deal-hurt4n'/s),rj.,-,~~,', M,ix~./n/2.'u'".,vJ,'(L

/J0 z,~('L APPEND I X I Coalition for Alternatives To Shearon Harris c/o Durham Research Office 919-682-3818 604 W.Chapel Hill St.Durham NC 27701 APPENDIX c C A Resolution Cc~oot'afng

-:ha shearon.Har ria,'Suc1ear F over Plant: MHERRAS, the'nunks,ear.

power ply tt accident on April P6~1986'in Chernobyl USSR has aroused widespread.concern with5.n the United States and throughout the.Morld about the safety~of'u clear power plants, and r t WHEREAS>Shel'I has bur/'aoed within, Chatham County I videspraad and int.ense opposition to,the nearly completed Shearon Harris Nuclear Power Plant:constructed by Caro1$M Power and Light Company, and MHEREAS, there are substantive~.

and unresolved issues about the Chatham County evacuation plan,.NON, THEREFORE, BE IT RESOLVED that the Chathem County.'oard of Commissioners hereby reaoinds a11 prior approvala of the Shelron Harris Emergency Response plan pending further cr it1cs 1.exam'.na tfon of the unr csol ved issues.This resolution shall be eff'eotive upon ac!optic'n.

This the 27th day of Hay, 3986, ar.omp on Chsirm~n axe~oone Clerk to the Board STATE OF NORTH CAROLINA COUNTY OF CHATHAhl AFFIDAVIT From: Dan Frazier Rt.9, Ol Jones Branch Rd.Chapel Hill, N.C.27514 962-2267, 967-9057 This affidavit is to indicate that at 3:00 pm on 28 June 1986 I heard the first reports that some of my neighbors living about three miles south of my home in Chatham County heard a Shearon Harris emergency siren at about 1:55 am on-28 June 1986.I was concerned that part of the evacuation system upon which I rely had malfunctioned.

I was also concerned that some of my neighbors were unable to find out what was happening for over 30 minutes.I was concerned enough to talk to some of the people who live near the siren and collect affadavits from them.I wanted to find out what happened and what effects the incident was having on those involved.At 10:33 am on 29 June 1986 I called Shearon Harris, 362-8793, to find out what had happened.The man who answered the phone said that he was in the guard shack;that he didn'know anything about any siren or alarm Saturday morning and that there wasn't anyone for me to talk to.He was basically uncooperative, uncommunicative and uninformed.

I had the distinct impression that he had been told that he didn't know anything.At 10:35 am I called the Chatham Sheriff dispatcher, 542-2811, and he said that an emergency siren on Pea Ridge Rd.had gone off Saturday morning.He didn't know which of the two alarms on Pea Ridge Rd.had gone off.At 12:10 pm I visited Barbara Keyworth and David Richardson on Hatley Rd.about 2 miles north of the siren which was reported to have gone off.At about.1:55 am Ms.Keyworth was awakened by a siren.She thought it was the Shearon Harris alarm because she had heard it before.She woke Nr.Richardson who also heard the siren.She estimates that the siren sounded about 3 to 5 minutes.They feared that they might be in danger since they knew there was nuclear fuel at the plant.They called the Chatham Sheriff, Shearon Harris, and Raleigh State Patrol.Only Shearon Harris had an explanation:

that they heard the shift change or break whistle.t4s.Keyworth did not accept this explanation since they live 11 miles from the plant.llore than 30 minutes after the siren sounded, they were finally told by the Chatham dispatcher that CPGL doesn't know why the alarm went off and that there was not an emergency.

Mr.Richardson then showed a cassette recording of the WRAL ll:00 news from 28 June 1986..In the newscast, Bill Lesley stated that CP&L officials had reported that vandals had broken into the Shearon Harris plant and set off an alarm.He also stated that CP&L planned to increase security at the plant.I was quite concerned since this was the third explanation that I had heard from CP&L.I was also a little amused.Amusement turned to shear entertainment when I read in the 29 June 1986 News and Observer a fourth and all new explanation.

Mac Harris, CP&L spokesman, was quoted: "We have clearly established that the siren was deliberately set off by some individual or individuals who vandalized the siren.Someone had to make a real effort to do it." I anxiously anticipate future explanations.

I am really intimidated to have my well-being in the hands of people who have given me every reason to mistrust them.At about 2:30 pm I visited with Mitchell Riley on Hatley Road about 2 miles from the siren.He and his wife Kay Riley were asleep at the time of the siren and were never awakened.They had their bedroom windows open and a quiet fan running.Mr.Riley stated that he had no faith in the evacuation plan and that they would probably move if the plant started up.At about 3:30 pm I visited Ruth Thomas on Pea Ridge Rd.Her house is located across the street (about 200 feet)from the siren which sounded Saturday morning.She was awake after 1 am Saturday morning and heard the siren go off for about 5 minutes.Within two minutes after the alarm started she went outside to her front porch to see if CP&L was testing the siren.Although the siren isn't quite visible from the front porch because of the trees, she was convinced that no one was at the siren.She heard no one and heard no vehicles.Also, her high-strung dog didn't start to bark until she was outside.She felt sure that the dog would have barked if someone had been at the siren.I was shocked that no one in her family was awakened by the siren.This includes her husband, Lieutenant Charles Thomas, of the Chatham Sheriff Department, and their son and daughter.The windows were closed and there were no fans or air conditioning running.It concerned me that one of the sirens, which we rely upon in case of a disaster, can't even wake people 200 feet away.Ms.Thomas knows Anne Wilke who was the dispatcher for the Ch Sheriff's Department at the time of the incident.Ms.Wilke told her that she was swamped with calls from people asking about the siren and had called in an extra dispatcher.

Ms.Wilke also told Ms.Thomas that she had called CP&L to find out what had happened and that they said that the siren had been turned on accidentally.

atham

Ms.Thomas and I then carefully examined the siren, the pole, the boxes on the pole and the area around the pole at 4:30 pm.I observed no breakage, no scratches or any physical damage at all.All of the locks were weathered.

There were no parts that looked new or replaced.Ms.Thomas said that everything looked the same as always to her.She had examined the siren closely.She concluded that she really doesn't believe that anyone vandalized the siren.I then drove to the south end of Pea Ridge Rd.to see the siren there about 1.5 miles from Ms.Thomas'ouse.

I felt that since this siren was located further from houses than the Thomas siren, it would be a better choice for a vandal.I then drove to the siren on Big Woods Rd.about 3 miles from the Thomas-siren.This siren is isolated far from any houses and would have been the best choice of the three for a vandal.I can't help but conclude that many of the other 66 sirens are also isolated.Why would a vandal pick the one across the street from a Lieutenant in the Sheriff's Department'P At about 5:30 I talked with Claire and Edward Thomas who live on Hatley Rd.about 2 miles from the siren.The siren woke him up and she was already awake.They thought it was a wreck or something.

They never thought about Shearon Harris.The incident left them less secure about the evacuation plan.At about 7:00 I talked to Radd Greenlaw on Hatley Rd.about 2 miles from the siren who was asleep and never heard the siren.Hei husband Raymond Greenlaw woke up but didn't know why.Ms.Greenlaw very angry about the incident.She has never had any faith in the evacuation plan.At about 7:30 I talked to Robert Hatley on'wy.64 about 1/4 mile from the siren.He was awake, heard the siren, knew exactly what it was and called the Chatham Sheriff (911).Anne Wilke, the dispatcher, didn't know anything and put him on hold.Anne then came back on the line and said they were investigating.

Then the line was somehow cut off.Mr.Hatley got no explanation that night.On 30 June 1985 at about 8:45 am I called Mac Harris, CP&L'pokesman, 836-6189.I identified myself and said I lived near the siren and had collected affidavits from about twelve people and that I wanted to find out what happened from CPGL's viewpoint.

The following is not verbatim, but accurately represents the ideas that were exchanged.

Harris: What are you going to do?Frazier: I just want to find out what happened.Harris: If you'e getting signed affidavits you'e obviously taking action against CPaL.What orders are you going to bring against us?(very agitated)Frazier: No kind of action.I was thinking of handing the affidavits over to the media.Harris: Oh yes, oh well, okay, the press then.What is it you want to know?Frazier: There are four contradictory explanations about what happened: (1)Shift change horn, (2)Error at the plant, (3)Vandals at the plant, (4)Vandals at the siren.Which is correct and how do you explain the other versions?Harris: It is absolutely clear that someone forceably physically removed a lock (which was later replaced)on a control box at the siren and set off a 3 min.cycle at full volume.The 3 min.cycle cuts off automatically after 3 min.It was probably someone with a purpose and an agenda.We know this happened and I'm not interested in proving it.I then asked Mr.Harris to address each of the other explanations mentioned above.He answers: Explanation 1-It is reasonable that the people at the plant thought it was a change horn.People at the plant had no way of knowing the alarm went off (He was unaware of this explanation).

Explanation 2-He was also unaware of this version.He thought the Chatham sheriff had control of the switches.He doesn'know who the Sheriff's department talked to at Shearon Harris.Explanation 3-WRAL got it wrong.He personally related explanation 4 to WRAL.Xt must have been changed in translation.

Mr.Harris stated that the sirens are fired by radio signal but can be set off from the box at the siren.There is no feed back from the alarms.The only way to know if an alarm goes off is to hear it.Apparently, the next time one goes off like this the same thing will happen again.Frazier: The alarm didn't awaken 3 people right under it.Will it be effective in an emergency?

Harris: That's just incredible.

It's about 127 decibels.I don't know what those peoples'leep habits are.Harris later admitted that hot humid conditions like those of 28 June 1986 have great damping effect on sound and since the sirens weren't reliable under those conditions people within five miles were given special radios to warn them.Not all

of the people in the 5-10 mile zone are supposed to hear the sirens.He said that they aren't in as much danger anyway.I asked about people (these people were just outside the 10 mile zone)not knowing who to call and not getting good answers.He replied thag it is a real problem that people eleven miles from the plant don't know what to do.People in the 10-mile zone had been instructed to tune into the Emergency Broadcast System.Supposedly if they hear an alarm and turn on the radio and don't hear about an emergency then there isn't one.He said that the people who live eleven miles from the plant were a tough issue since they could hear the sirens but hadn't been informed of what to do.He said CP&L should do something about it.When Mr.Harris heard about the siren at about 2:30 am 6/20/86 he thought about calling the press but didn'know who to call at that hour and so called no one.I informed Mr.Harris of all the evidence (previously mentioned) that seemed inconsistent with the vandal at the siren hypothesis.

He was agitated and said I'd just have to accept his version as fact.Mr.Harris did not say how the siren was set off in the interest of not letting people know how to do it again.I asked if a system might be installed to notify some authorities immediately when an alarm goes off.He said he didn't know if such a system existed.Mr.Harris took my number and said he would contact me if he got any new information.

I thanked him and said goodbye.The information I learned from my neighbors leaves me very distressed.

The sirens will not reliably awaken us and many won't know what to do if we hear it.There may be more false alarms and the authorities will not have any immediate answers.If I hear a siren I'l evacuate first and ask questions from Virginia.Because of the four different explanations given for what caused the siren, I feel I cannot trust CP&L to let me know what is happening, even after the fuel is loaded.I have a strong fear that if there is a radiation leak from the plant that CPGL will take whatever action is in its interest and will not act primarily in the interest of threatened citizens.I have written the above statement and believe that it is a true and accurate statement of the events and occurences described therein.

STATE OF NORTH CAROZ INA COUNTY OF'CHATHAN Addendum to Affidavit of Dan Frazier 6/28/86 Contemporaneous to the printing of this affidavit I learned of new information which indicated that the siren in front of Ruth Thomas'ouse may not have been the one which sounded on 6/28/86.It was probably the siren on Hank's Chapel Rd.near Ns.Thomas'hich sounded.This information corroborates CP&L's explanation that a vandal set off the siren at the siren.I have written the above statement and believe that it is a true and accurate statement of the events and occurrences described therein.

AFFIDAVIT FROM: Barbara Keyworth David Richardson Route 4, Box 641 Pittsboro, NC 27312 This affidavit, taken by Dan Frazier at 12:30 p.m.on June 29, 1986, is to indicate that to the best of their recollection Barbara Keyworth and David Richardson heard a siren just before 2:00 a.m.on June 28, 1986.Their home is about two miles from the siren which was later reported to have sounded.Ms.Keyworth heard the alarm first and thought it was the emergency alarm for Shearon Harris because she had heard it once before She woke Mr.Richardson.

&>Ill ac%'cu8 o mc.ck'%l4iA9 l)(i'lo+M-/l&44>y wn/~ad.ig pQll/.At about 2:00 a.m.Mr.Richardson dialed the operator and asked for the police.He did not know which police he spoke with.The police expressed surprise and doubt that it was Shearon Harris.Mr.Richardson had a very eerie feeling and really felt that something was wrong.Ms.Keyworth though that if the alarm was going off then something must be wrong at the plant.Mr.Richardson.

called the operator and asked for a number for Shearon Harris.The number he tried was disconnected or no longer in service at that time.Ms.Keyworth dialed 911 and talked to Anne%like, dispatcher for the Chatham Sheriff Department.

She was surprised and did not know anything.Mr.Richardson called the operator and got two numbers for Shearon Harris, 362-2320 and 362-8891.He dialed 362-2320 and reached Murdoch Jones in security.Mr.Jones said there had not been an accident and that the horn was for the shift change or break.Shen asked for his supervisor, he ignored the request and restated that it was the break siren.Mr.Richardson called 362-8891 and reached David Dean of the payroll office, who said that the siren was for the shift change and that it went off at 2:00 and 4:00 every morning.Mr.Dean expressed irritation and was sure that Mr.I Richardson had heard the shif t change whistle.~'r)rg.zend><<<~L%P~~~-0+i<'~Wld Ck<V<'lO!~CLlddSgm Ct~1)lS.KaqiOPla i~i~Ji~p(gg During these phone calls, Mr.Richardson and Ms.Keyworth wondered whether they should go ahead and evacuate or stay and keep trying for an explanation

~They were aware that nuclear fuel was present at the plant.They really felt helpless.Since the Chatham commissioners had pulled out of the

evacuation plan, Mr.Richardson thought he should contact the people who would take over the evacuation plan, the State.;Patrol.Ms.Keyworth called the operator for the State Patrol number.The operator asked, for what city?Ms.Keyworth said, for Pittsboro.

The operator said there was not a patrol office there.Ms.Keyworth asked for Raleigh and got a number.She dialed that number and reached Trooper Mhitehouse, who laughed at her concerns and did not take her seriously.

He said that he lived six miles from the plant and that there was nothing down there.She replied that there was nuclear fuel there.He asked,"Mhere did you hear that?" in a tone which implied that she was misinformed.

He made no indication that he would do anything at all.Ms.Keyworth answered that it was public information and that there was the potential for a problem.She said that there had been problems at other new plants before fuel loading.q Ms.Keyworth asked Mr.Mhitehouse to call Shearon Harris and ask what happened.He agreed to.He called back quickly and said it was the break bell.Ms.Keyworth said that that was impossible, since she lived eleven miles from the plant.Mr.Mhitehouse replied that they were testing the, sirens all the time.She answered that she had never heard one at night.Mr.Mhitehouse suggested that maybe someone had pushed the wrong button, and then said that he was not going to argue at 2:00 a.m.V)g 5:i~>'.mt'ARS agAoiyc/<))<~I.pe pj~>'nQlu~id~n cx ch)cia~:o f)0'ilaw)4f'Alck~s~'I$(N>oc'8~~~/I Q)la<gJv ba J g<ggc<0 Ig&p.gY'p)anal P~Qc'<1'):<+<<~

Ms.Keyworth told Mr.Mhitehouse he had laughed at her.He replied,"No, I didn'." She asked his name.He replied,"Mhitehouse, and I'm the night supervisor." She hung up, angry.Mr.Richardson called the governor's hotline, (800)662-9952, and got no answer.Ms.Keyworth called the Governor at 733-5811 and got no answer.She called MRAL radio and got no answer.She called the 94Z radio station and got no answer.S>)v its iw:~I'kvSh:i9-i/

Il~d>Uuu~ij i<E~ii Dl.0 n~'>coo'g~nH wo c k.ace~pn.c o~-I i<

~i At 2:35 a.m.Ms.Keyworth called Anne Milke, the Chatham Sheriff's dispatcher. Ms.Milke said she had someone from CPSL on the line who wanted to know what the siren sounded like and how loud it was.Ms.Keyworth imitated the slowly oscillating, wailing sound.Ms.Keyworth then asked if there had been other calls.Ms.@like said"several," and then said it was not an emergency and CPSL did not know why the alarm had gone off. Ms.Keyworth and Mr.Richardson got back to sleep after 3:30 a.m.At 9:00 a.m.on June 28, 1986, Ms.Keyworth called 911, the Chatham Sheriff, about the alarm again.She was told that"someone down at, the plant set it off accidentally." Qf 4)pl<.Eo<I<i)~+~@ncaa Ill k~~f<l4&us~w~'c v.<&j chc~wcI.orle Ai il I((cq(.4>~8~l'~cPsr-eh@n+~~an)Rfawd>m'Lj mucker'o ganu4fa"1<(a. ef!OJl~g~g:g Q~$g;J pQ~Spa/lg d~<Q.,/J~r<~Q~~piq PIBglgaLADn Pdzl L'i g'kill]Mf)4(~d/2e ant Hundt Lf~1 a~/i+<~il!, C.844 gi~l~Mc~~d~(Neo~rupia'yean% >~+4I~egg.qgqg, cJ To the best of my knowledge this statement accurately reflects the substance of my conversation with@gp P-8 Q.X.E i~~a7//~L fC Qlr i)i Jg8/Q~I have read the above statement and believe it is a true and accurate statement of the events and occurrences described therein. AFFIDAVIT From.Ruth Thomas Route 4, Box 835 Pittsboro, NC 27312 542-4030 This affidavit, taken by Dan Frazier at about 4:00 p.m.on June 29, 1986, is to indicate that on June 28, 1986 after 1:00 a.m., Ms.Thomas was awake and heard a siren go off.The siren is across the street from her house on Pea Ridge Road and is about 200 feet from her house.The siren sounded for about five minutes.She knew immediately that it was the Shearon Harris emergency siren and went outside to see if CP8L was testing the alarm.From her porch she saw and heard no one and no automobiles. There are trees that block the view of the siren from the front porch but she believed no one was there.Her excitable dog was sleeping outside in front of the house and did not start, barking until she went outside.She felt sure that if someone had been present the dog would have barked.She noted that the alarm did not seem as loud as it had when she had heard it previously. She did not feel that it was loud enough to awaken people.In fact, her husband Charles Thomas and their two children never awakened during the incident.Their windows were down and no fans or air conditioning were on.She was not concerned that there was an emergency because she was monitoring a police scanner and she believed that the Sheriff's Department would have to have been called before the alarm could have been sounded.Since there was no news she assumed the siren to be a test or an error.She is worried that the sirens will not wake people up in an emergency. Ms.Thomas does not believe that anyone vandalized the siren.She examined the siren, the pole, and the boxes on the pole I\ carefully at about 5:00 p.m.on June 29, 1986.She stated that everything looked normal to her and she saw no evidence of tampering. She had examined the siren previous to the incident.At some time long af ter the siren sounded Ms.Thomas called Chatham Sheriff dispatcher Anne Milke.Ms.@like told her that she had been swamped with calls from people concerned about the siren and had had to call in an extra dispatcher. She also stated that she had called CPLL and that they had said the siren was accidentally turned on.To the best of my knowledge this statement accurately reflects the substance of my conversation with I 1 I I r l I have read the above statement and believe it is a true and accurate statement of the events and occurrences described therein. AFFIDAVIT From.Anne Greenlaw Route 4, Lot 2, Jordan Moods Hatley Road Pittsboro, NC 27312 542-3465'I~This affidavit, taken by Dan Frazier at 11:00 a.m.on June 30 p 1 986 p is to indi ca'te that Anne Green 1 aw was awake at 1 55 on June 28, 1986, and heard a siren which was very faint.Her home is about two miles from the siren.Her windows were closed and an air conditioner and fan were on.She never considered that the siren might be from Shearon Harris.Mhen she learned that it was an emergency siren for Shearon Harris she felt much less safe because the alarm malfunctioned and because it was too faint to elicit an evacuation response.To the best of my knowledge this statement accurately reflects the substance of my conversation with I have read the above statement and believe it is a true and accurate statement of the events and occurrences described therein. st I q i~/3 I AFFIDAVIT From.'laire and Edward Thomas Route 4, Box 638 Hatley Road Pittsboro, NC 27312 542-3637 This affidavit, taken by Dan Frazier at about 500 p.m.on June 29, 1986, is to indicate that Claire Thomas was awake at about 2:00 p.m.on June 28, 1986, and'hea'r'd a siren.Thomas ,Edwards was awakened by the siren.Their windows were open and no fans were running.Their home is about two miles from:.~the siren4 They thought the siren was from a~reck or~J something and never thought of Shearon Harris.'hen they learned that the alarm was from Shearon Harris, they felt less secure about the evacuation plan.To the best of my knowledge this statement accurately >': reflects the substance of my conversation with"a~r'I I have read the above statement and believe it is a true and accurate statement of the events and occurrences described therein. 'Comment on Outdated Federal Guidance for Size of the Emergency Planning Zone Kenneth G.Sexton, Ph.DE Research Associate Dept.Environmental Sciences and Engineering School of Public Health University of North Carolina June 30, 1986 Q."IS A 10-MILE EVACUATION A1KA ADEQUATE?" A.NO ONE REALLY KNOWS.Why not?There are many uncertainties in predictions of nuclear-power-plant-accident consequences. These result from uncertainties in the prediction techniques and in input data.The NRC is currently attempting to resolve major uncertainties for risk assessment. Generic rather than site-specific calculations were performed (using some outdated techniques and over-simplifying assumptions) to help determine the distance.The 10-mile evacuation plan is supposedly adequate to use as a base for evacuating additional areas outside the 10 miles as needed on a"ad hoc" basis when an accident does occur.No one knows if it will work until-an accident happens because there are no required formal, predetermined, evacuation plans in place outside the 10-mile area to evaluate.No one claims that deaths'nd injuries will not occur outside the 10-mile EPZ in the case of a more severe accident.There are several important points that should be made very clear to all officials concerned about protecting the safety and health of the people in the countie" surrounding a~n nuclear power plant.These facts come from reports and regulations from the Nuclear Regulatory Commission and the North Carolina Emergency Response.Flan (NCERP).The immediate concern is with the Shearon Harris Nuclear Power Plant (SHNPP).However, the following discussion app'lies to any nuclear power plant of comparable size because the 10-mile EPZ is a generic distance which applies to all U.S.nuclear plants of comparable size. The 10-mile emergency planning zone (or EPZ)is based on findings of a joint NRC-Environmental Protection Agency (EPA)Task Force which were published in 1978 (NUREG-0396). They concluded that the 10-mile EPZ was more than adequate to protect the public.However, it is also made clear that: Although most early fatalities and injuries will occur inside the 10-mile EPZ, the NRC (NUREG-0396, pg 17;NUREG/CR-2239, pp 1-3 to 1-6)and the NC Emergency Response Plan (NCERP, Part 1, pg 1)acknowledge that some of the early severe health effects (injuries or deaths)which would result from the more severe accidents will occur beyond the 10-mile EPZ."In addition, the EPZ is provide for substant health effects (injuries of the more severe Class (NUREG-0396, p 17), of sufficient size to educt'on in early severe or deaths)in the event 9 accidents." 2)The size of the EPZ and the emergency plan are not restricted to, nor designed specifically for protecting only the people in, the 10-mile EPZ.They are designed for the protection of all areas and all people that could be affected by an accident.The NRC assumes that any emergency plan deemed adequate for a 10-mile radius is sufficiently detailed to be adequate to cover emergency needs in areas beyond the 10-mile EPZ (NUREG-0396, pp 15-16).The NRC, CP&f, and NCERP acknowledge that emergency response outside the 10-mile EPZ may be needed."The size of the EPZ represents a judgment on the extent of detailed planning needed to assure an adequate response base" (NCERP, Part 1, pg 1).The concept in the NCERP and NRC guidance is to use the EPZ planning as a"base for expansion of response efforts if necessary" (NCERP, Part 1, pg 1)and to respond on an"ad hoc" basis (NRC, NUREG-0396, pg 16).3)The size of the 10-mile EPZ is"tempered" by probability (NUREG-0396, pg 15).Some amount of risk was determined by the NRC to be acceptable. Their decision was affected by low-probability estimates of the occurence and nature of severe accidents (NUREG-75/014). More recent NRC reports indicate that many of these earlier accident estimates may be too low (NUREG/CR-0400 cited in NUREG/CR-4199, pp 1;and NUREG/CR-4199, pp 6-9).There is much uncertainty in risk and probability estimates, as well as disagreement among experts on this matter (as indicated in different NRC reports).The inclusion of a greater accident probability could result in the establishment of a larger EPZ upon reevaluation. Also, it should not be implied that the term"low-probability accident" indicates that a long time will pass before such an event occurs.It is therefore reasonable to expect that consideration of emergency plans be"tempered" by these uncertainties. Local officials should plan accordingly, especially when highly-populated 'areas are very near but beyond the presently-accepted 10-mile EPZ.4)The latest NRC regulations published January 1, 1986 cite~onl this 1978 Task Force report as a basis for determining the EPZ (10 CFR 50.47 and its Appendix E).No report is cited which discusses oz suggest a smaller EFZ for nuclear plants the size of the SHNPP.Simple techniques and information now known to be inappropria e, or a least not the best, were used for generic calculations used in determining the 10-mile EFZ.Furthermore, seemingly inconsistent NRC regulations do require"state-of-the-art" (current). computations be performed after an accident using site-s ecific information (eg.information specific to SHNPP)(NUREG-0654, Appendix 2, pp 2-2 and 2-3)."State-of-the-art" models (NRC-sponsored) have been used in recent years to estimate radiation doses to the public under a variety of accident and normal operation conditions, but evidently have not been used for reevaluation of the EPZ (NUREG/CR-2239, NUREG/CR-4199, NUREG/CR-3344, NUREG/CR-4000). Uncertainty is a major problem in accident predictions (NUREG/CR-2239, pp 2-7 to 2-10).There is, in fact, an on-going program for reevaluation of nuclear accident risk at the NRC, but work to date has been"greeted with skepticism... There is a disagreement over the credibility of some computer modeling codes that are the basis for all the predictions that will come out of NUREG-0956" (~Bci nce, April 1986, pp 153-154, attached).Therefore, there is justification in requesting the NRC to review and update the 1978 Task Force Report, and consequently the justification for the size of the EPZ.Current thinking would suggest that the NRC should require the SHNPP and all other plants to reevaluate the 10-mile EPZ using on-site and national weather service weather data specific to the area. Local officals are responsible for deciding if this type and size of emergency planning is acceptable and adequate.There should be demonstrable. assurance of ad hoc capability being adequate.For example and specifically related to the SHNPP, consideration should be given to the effect on local emergency response efforts if it were determined that Raleigh (and the state government) needed to be evacuated. Local officials must, decide if they accept the very low NRC accident-risk and probability estimates which were determined before the Three Mile Island accident--a serious accident which occurred despite its"low probability" of occurence. Those responsible for assuring the health and safety of the public should be aware that current techniques have not been used in establishing the EPZ and that there are serious questions in regard to some of the assumptions under which it was established. The obvious implication is that these calculations and the resulting 10-mile recommendation are therefore suspect and uncertain for purposes of protecting public health.ADDITIONAL DISCUSSION The 10-mile Emergency Planning Zone (EPZ)is recommended by the Nuclear Regulatory Commission (NRC)as follows: "Generally, the plume exposure pathway EPZ for nuclear power plants shall consist of an area about 10 miles (16 km)in radius, and the ingestion pathway EPZ shall consist of an area about 50 miles (80hm)in radius.The.exact, size and configura ion of the EPZs surrounding a particular nuclear power reactor shall be determined in relation to local emergency response needs and capabilities as they are affected by such conditions as demography, topography, land characteristics, access routes, and jurisdictional boundaries." (10 CFR Part, 50.47"Emergency Plans")This regulation recognizes that, approximately a 10-mile radius is appropriate, but, also implies that alternate sizes and configurations may be very significantly more appropriate. Although the regula ion requires consideration be given to several area-specific fac ors, no mention is made of local meteorology. This is in contradiction to regulations for siting and post-accident calculations (10 CFR 100.10 and 10 CFR 50.47, respectively), and the findings of more recent accident-consequence estimates (NUREG/CR-2239, p 1-3), all of which consider local meteorology. Local o ficials must carefully determine local emergency response needs and the adequacy of.emergency capabilities in approving a plan specific to a given nuclear power plant. The 10-mile EPZ is based on the report of a joint NRC-Environmental Protection Agency (EPA)Task Force which was published in 1978.The report's principal meteorological references are dated 1968 and 1970 (USAEC, 1968;Turner, 1970).The report concluded that the 10-mile EPZ was more than adequate to protect the public.However, they used 1)meteorological techniques that are now outdated, and 2)nuclear-reactor-accident estimates developed before the Three Mile Island accident experience and before subsequent a'dditional experiences with nuclear reactor problems.These early calculations and EPZ estimates depend on the estimates of the amount of radioactivity that would be released during accidents and the probabilities of different types of accidents occurring. Assumptions were made which now may be incorrect or inappropriate. Very simple assumptions were made concerning the behavior of the radiation plume that might be released in an accident.The atmosphere and its weather systems are very complex, and a wide range of plume behavior is possible."The weather conditions at the time of a large release will have a substantial impact on the health effects caused by that release" (NUREG/CR-2239,. pg 1-3).Given a plume released during an accident that would result in injury within the 10-mile EPZ, there are meteorological conditions which could result in significant exposure at distances beyond the 10-mile EPZ and even hundreds of miles"downwind". The plume can meander rather than travel in a straight line, making predictions of exposure difficult and allowing for multiple exposures to the population. Also, important considerations such as the effect of rain were mentioned but not included in calculations used in the final distance determination in the 1978 report (NUREG-0396, pp I-25 and I-26).The importance of the effects of rain on downwind radiation doses to the public are now documented by the NRC (NUREG/CR-2239; NUBEG/CR-1244). Significantly-larger doses to the public can occur further downwind if the radiation release is"washed-out" of the air by rain (rain can clean the air of radioactive particulate as it falls, creating"hot spots" on the ground).On the official average, North Carolina receives rain on one of every three days.As another example, it was assumed in the report that the major dose exposure would occur within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the accident.This assumption is debatable and has several implications. The evacuation time estimate for the NC Emergency Management Plan for the SHNPP is almost 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.Sheltering in place until the released radiation pa ses may be the best strategy under some adverse conditions, but some meteorological conditions could result in long and uncertain sheltering times (waiting)while some lower-level exposure continues. Therefore, careful dose estimates and monitoring, accurate evacuation-time estimates, and good management by emergency personel are needed to minimize personal injury not only within the 10-mile EPZ but also at distances beyond the 10-mile EPZ.Unfortunately, beyond 10 miles these types of decisions and management will be performed ad hoc after an accident occurs.With a mean wind speed of approximately 7.5 mph in this area, there will not be much time (1-2 hours)before there could be a problem beyond 10 miles.It is prudent t;o be able to respond to problems beyond this distance for this reason, if for no other.All nuclear units operating in this country are subject to the same type of plan.The calculations used for determining the 10-mile EPZ were performed for hypothetical accidents and meteorological systems.The generic 10-mile-distance calculations obviously do not use meteorological parameters or other factors specific for the Shearon Harris site and power plant.There are now better methodsfor modeling a specific site which result in more appropriate calculations. The NRC now uses more up-to-date (more correct)techniques and computer models to estimate site-specific radiation releases and doses to the public.Several of these models were developed by the NRC itself but evidently have not been used for reevaluation of the 10-mile EPZ.Even with these improved techniques, it is recognized and'ocumented by the NRC that the reliability of the rish and dose estimates is still limited by the uncertainty of the amounts of radiation that will be released during accident-(NUREG/CR-4199, p 8).These uncertainties are further increased by the uncertainties of the meteorological estimates (NUREG/CR-4199, p 9;NUREG/CR-2239, p 1-3).The obvious implication is that these calculations and the resulting 10-mile recommendation are therefore suspect and uncertain for purposes of protecting public health.Reevaluation with more current methodologies and recent experience could result in a larger EPZ distance which would require modification of the emergency plan and required participation out ide a 10-mile radius before licensing of a plant.Part of demonstrating that an emergency plan is adequate is to show that the size of the area affected by the plan is appropriate. The problems and limitations of the older methodologies are now well documented. Xh~os-es ons'b e o ass t he~+and safet of the ublic should be aware that current techni ues have not been used in establishin the EPZ and that there are serious u.tions n re ard to some of the assumption under which it was estab ished.Conse uentlg 1 serious in the case of the SHNPP because heavily-populated areas including the state government systems exist so close to the presently-accepted 10-mile EPZ. An appendix is being prepared which further documents these statements, includes additional findings and comments, an)contains references which document the widely accepted criticisms of the older and simpler assumptions, dispersion parameters, and methodologies. These criticisms are found in 1)reports from the NRC, EPA, AMS (American Meteorology Society), a joint AMS-EFA workshop, and a Department of Energy (DOE)-sponsored DOE-AMS workshop;and 2)a statement from Herschel Slater, formerly of the Monitoring and Data Analysis Division, Office.of Air Quality Planning and Standards, EPA, a meteorologist who co-authored the guidance document for EPA Air Quality Models in 1978 (This"tatement is attached). Statement by the author: I am a research associate in the Department of Environmental Sciences and Engineering at the School of Public Health, University of North Carol-'na, Chapel Hill, where I received my Ph.D.My research field is atmospheric chemistry and computer modeling of pho ochemical smog.This report represents an independent study not done in connection with my work at UNC.My personal interest in the emergency plan for the Shearon Harris Nuclear Power Plant (SHNPP)is in regard to the techniques used to establish the size of the emergency planning zone.My reason for preparing this report is a sincere concern that the present plan and zone may be less than adequate to protect the general public in the event of an accident at the SHNPP.I am neither an anti-nuclear activist nor a member of the Coalition for Alternatives to Shearon Harris Steering Committee. Kenneth G.Sezton, Ph.D l References Cited In This Summary NUREG-0396; EPA 520/1-78-016,"Planning Basis for.the Development of State and local Government Radiological Emergency Response Plans in Support of Light, Water Nuclear Power Plants," December 1978.NUREG/CR-2239,"Technical Guidance for Siting Cri eria Developmen ", SAND81-1549, December 1982.NUREG-75/014,"Reactor Safety Study: An Assessment ot Acciden Risks in U.S.Commercial Nuclear Power Plant,s, WASH-1400, U.S.Nuclear Regulatory Commission, 1975.NUREG/CR-0400,"Risk Assessment Review Group Report to the U.S.Nuclear Regulatory Commission," NRC, 1978.ilUREG/CR-4199,"A Demonstration Uncertainty/ Sen-itivi"y Analysis Using the Health and Economic Conseauence Model CRAC2," Hay 1985.T'tie 10 CFR, Chap e , Nuclear Regulatory Commision, Part 50.47,"Emergency Plans", 1-1-86.Title 10 CFR, Chapter 1, Nuc ear Regulatory Commision, Part 50, Appendi..E, Emergency Planning and Preparedness for Produc ion and Utilisation Facilities", 1-1-86.NUREG-0654/REV-1, Appendix 2, including ANNEX I,"Criteria for Preparation and Evaluation oi Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," November 19SO.NUREG/CR-3344 HUREG/CR-4000 Science, April 1986, Vol.232, pp 153-154,"Nuclear Meltdown: A Calculated (and Recalculated) Risk".(HCERP)Horth Carolina Emergency Response Plan, In support of the Shearon Harris Nuclear Power Plant Feb.1984, Rev.l Sept,.1984.USAEC.Heteorology and Atomic Energy-1968.D.Slade, ed.TID-24190. National Technical Information Service, Springfield, Va.22151 Turner, D.Bruce, Workbook of Atmospheric Di"persian Estimates. Ap-26.USEPA Office of Air.Programs, Research Triangle Park, HC 27711.1970 Revision. NUREG/CR-1244,"Impact of Rainstorm and Runoff Modeling on Predicted Consequences of Atmospheric Releases From Nuclear Reactor Accidents, U.S.Nuclear Regulatory Commission, February 1980."Guideline on Air Quality Models", J.Tikvart and H.Slater, EPA-450/2-78-027, OAQPS No.1.2-080, Research Triangle Park, NC, April 1978.Some Additional References Referred to In Last Paragraph of Summary Which Will Be Cited in the Appendix EPA/600/S3-85/072,"Research on Diffusion in Atmospheric Boundary Layers: A Position Paper on Status and Needs," Project Summary, G.A.Briggs and F.S.Binkowski, December 1985.EPA/600/S3-85/056,"Atmospheric Diffusion Modeling Based on Boundary Layer Parameterization," Project Summary, J.S.Irwin, S.E.Gryning,.A.A.M.Holtstag, and B.Sivertsen, December 1985.Hanna, S.R., G.A.Briggs, J.

Deardorff,

B.A.Egan, F.A.Gifford, and F.Pasquill,"AMS Workshop on Stability Classification Schemes and Sigma Curves--Summary of Recommendations," Bulletin American Meteorological Society, Vol.58, No.12, pp 1305--1309, December 1977.Weil, J.C.,"Updating Applied Diffusion Models*", J.of Climate and Applied Meteorology, Vol.24, No.11, pp 1111-1130, November 1985.~June 1985--This paper is an overview of the review and recommendations arising from the AMS/EPA Workshop on Updating Applied Diffusion Models held in Clearwater, Florida, January 24-27, 1984."Proceedings of the DOE/AMS Air Pollution Model Evaluation Workshop", Kiawah, South Carolina October 23-26, 1984, Volume 3, Summary, Conclusions, and Recommendations, DP-1701-3, Robert J.Kurzeja, and Allen H.Weber, Approved by A.L.Boni, Research Manager, Environmental Technology Division, Sponsored by the Office of Health and Environmental Research, U.S.Department of Energy, Publ'ication Date: December 1985, E.I.du Pond de Nemours 5, Co., Savannah River Laboratory, Aiken, SC, 29808, Prepared for the U.S.Dept.of Energy under contract DE-AC09-76SR00001.

Statement Concerning the Procedures for Selecting the Size and Configuration of an Emergency Planning Zone (EPZ)Herschel H.Slater, Consultant Air Pollution and Heteorology Chapel Hill,NC 27514 June 28, 1986 (X am a meteorologist, specializing in air pollution matters with experience and training that spans four decades.Hy experience includes service with the US Weather Bureau;US Air Force, as a career officer;Environmental Protection Agency;Adjunct Associate Professor, School of Public Health, UNC-CH;and Logistics Hanager for Project GALE for NCSU and the Natonal Center for Atmospheric Sciences.)

I am concerned about the size and configuration of the emergency planning zone (EPZ)as it applies to the Shearon Harris Nuclear Power Plant.CPL and the State of North Carolina apparently have accepted the Nuclear Regulatory Commission's suggested plume exposure pathway EPZ, NRC suggests an essentially circular area having a radius of'bout 10 miles.Fortunately, meteorological data and analytical techniques have been developed over the past decade that enable more definitive configurations of EPZ's.CPL has the data and the competence to apply more sophisticated methodologies to this problem than the generic approaches suggested in NRC-promulgated regulations.

CPL should be required to re-evaluate the proposed boundaries of the EPZ.I expect the result would be a more realistic and effective emergency response plan.Since the NRC regulations that pertain to the size of an EPZ were issued, most nuclear power facilities collect meteorologica data on site~Not only are the date site-specific, but they are designed to be applied directly to the problem of estimating the transport and dispersion of a.cloud or plume of radioactive material..Until such weather data began to be collected by commercial nuclear facilities, the weather data used to assist in choosing the boundaries of an EPZ usually came from the nearest official National Weather Service station.Xn the case of SHNPP, this is the station at the Raleigh-Durham Airport.Data collected at RDU is of highest quality.The equipment is well-designed, excellently maintained and the observers are well-trained and dedicated civil servants'.

The problem ,is two-fold: 1)The data are not observed where, in the event l

of an accident, the radioactive plume will generate and 2)The equipment is not designed to sense the.meteorological phenomena that determine the rate that a plume of nucle'ar material<<ill disperse, The equipment and observation procedures used at RDU are designed to meet the needs of aircraft operations and safety and to meet the needs of forcasters in preparing forecasts for the general public.The scales (or size)of atmospheric motion sensed for these purposes are much larger than those which control the dispersion of a plume.The wind equipment at the airp'ort is designed to be inscnsitivc to the small gusts that are significant in determining the dispersion process.Observations are generally made at hourly intervals.

This is much less frequent than needed to characterize the power of the atmosphere to disperse pollutants and to sense the rapid changes of gustiness during periods of the day when this phenomena changes rapidly.Also, the wind observations are made at 10 meters, about 32 feet, above the ground, far below the height that a plume likely may travel, CPL has a body of meteorological data gathered by sensing equipment specifically designed to study and estimate the dispersion and transport of clouds or plumes of pollutants.

Unlike the equipment at RDU it is sensitive to the important small-scale motions of the atmosphere.

Also, some data are sensed at heights where a plume is most likely to occur.The rate a cloud disperses is often determined by the character of the surrouding topography.

The character of the gustiness is influenced markedly by the roughness and the thermal response of" the surrounding surface.Is it farmed or forested?Plowed or covered with vegetation?

Is a body of water nearby?The nearby SHNPP lake must have a significant affect on the way the atmosphere would disperse pollutants in the event of an accident.The lake's effect varies with season, time of day and cloud cover.With these considerations, good judgment dictates the use of available on-site data rather than data from a distant point when developing the optimum EPZ.NRC documents stress the importance of crainfall on peale concentrations.

A shower may immediately create a surface"hot spot".If a plume is emitted into a rain situation, little of the radioactive material may leave the site itself.Mith rain occurring on the average of about one day in three in central North Carolina (e-cept in 1986!), careful analysis of rainfall statistics may dictate EPZ boundaries different than a circle.Notwithstanding current NRC regulations, CPL and thc State can take the initiative to fine tune the configuration of the SHNPP EPZ.CPL has the data and the professional competency to do so.In light of the concerns of so many, it is prudent for CPL so to do.

In addition, ncw hunch criteria will be cstablishcd at thc outset, Trulv said."When it's time for the first flight, we are going to do it as safely as possible.Wc are going to launch in thc daytime from Kenned>[Space Center in Florida], ivc'rc going to have a conservative flight design,[and]ive'rc going to have a repeat payload, one that wc have cxpcricnce with." No civilians will fly during thc first year, and all flights will occur in warm weather, he indicated.

Truly cxplaincd that thc rules arc neces-sary to restore the agency's credibility in the wake of thc Challenger disaster (Scig;>su, 28 March, p.1495).Thc agency's present plan is to conduct roughly nine flights a year, beginning a year from now.First priority will bc given to launching military satellites, as well as a tracking and communicauons satellite dcstroycd by thc accident.'Wc can-not print enough mona~to make thc flights risk-free, Truly added."But wc certainly arc.n".going to correct any mistakes that wc may have made in the past, and wc arc going to gct going again just as soon as wc can." r':.R.JEFFREY'SMITH Panel Sees Decline in Undergraduate'Education A National Science Board committcc rc-port says that the nation's undcrgraduatc programs in science, mathematics, and engi-neering"havydecline'd in quality and scope to such anI>extent that they are no longer meeting~IIational needs." This poses a"gravepkong-term threat to thc nation's sci-entific and tcchnical capacity, its industrial and'cono'IIIic competitiveness, and the length of its national defense," thc panel/watns.On thc basisepf evidence gathered in its inquiry, the cominittce pinpointed three ar-eas that require highest prioritv attention.

r Laboratory Instruction was described as"often uninspired, tedious, and dull." In-strumentation and facilitics werc found to be obsolete and inadequate

-thc need for ncw Instruments was put at$2 billion to$4 billion.r Faculty members in too many nses werc seen as unablc to maintain their teach-ing skills, currency in their disciplines, and command of ncw technology.

Serious short-ages of qualificd faculty werc noted in some disciplines.

r Courses and curricula werc dcscribcd as"frequently out-of.date in content, uniinagi-nativc, poorly organized for students with'iflcrcnt interests, and (they)fail to rcflcct rcccnt advances in thc understanding of tnching and lcarning.-

Briejg: New Shuttle Director Promises Emphasis on Safety A ncw emphasis on safety will be'thc hallmark of thc space shuttle's operations when flights resume, according to Rear Ad-miral Richard Truly, thc new associate ad-ministrator for space flight at thc National Aeronautics and Space Administration'NASA).

Speaking on 25 March beforcyan enthusiastic crowd at thc Johnson/pace Center in Houston, Texas, Truly outlined a scrics of activities that hc said are~required to establish a rcalisuc and achiyrabie hunch rate that will bc safely sustairNfble." r Specifically, the entire budget and pro-gram management"philosophy, structure, reporting channels"'rfd decision-making process will be tho ughly rcvicwcd," hc said.All shuttfe co poncnts considered vital to thc safety of c orbiter and thc crew will bc rcasscssed, will all waivcrs of engineer-ing redund cy.Inspection and test rcquirc-ments w'c reviewed, and the booster joints, Idely recognized to have been the nus f thc shuttle accident in January, will bc cdcsigncd under thc direction of thc arshall Space Flight Ccntcr in Huntsville, Alabama..Roughly$60 million of the neev funth s ught for this year are to bc transfcrrcd fr the Pentagon to DOE,,presumably for onc r morc underground tests in Nevada, yo thc nvo to four tests alrcad>schcd-ulcd fo this fiscal year at a cost of$157.8 million.fiscal year 1987, thc under-ground t ting account will jump to$226 miiiiony or nough for three to fiv explo-sions.(The dgct for underground testing of the weapon has cxcceded that for labora-tory research fo scvcral years.)In addition to the x-ray laser, variety of nuclear-driven weapons such as article beams, micro-waves, hypcrvclocit)

Ilets, and optical la-sers arc also under i vcstigation and may~eventually bc tested.'These nuclear power urces, if you want to consider them that wa (they arc cxplo-sions but they act as powe sources)," may ulumatcly bc unnecessary for a ballistic mis-sile defense, Wagner testifie.ut"thc first stages of the SDI program, wh h...may last decades, I bclicvc and thc D artment believes will have this nuclear component, this new kind of nuclear-driven tlirccted energy weapon as onc of its very im rtant options." r R.JEEPREv SMITH ccording to the rcport, institutions of all in all regions of thc countIy arc affect-'d.

e problems of enginccring disciplines were'd to be most serious.Thc ommittee was formed last May to assess th state of undergraduate education in science, mathematics, and cnginecring and make r ommendations nn thc role thc National Scic cc Foundation should take in improring it.chairgnan was Homer A Ncal, provost the-State University of Ncw York at Sto y, Brook.Thc committee reported to the,.'ational Science Board, which is thc jIdlic)making body for thc foundation.>'n its rem~mmendat ns, thc committee said that N SF lacks the resources to solve the problems itself, but should take a Icadcrship role in stimulating the state's and thc private sector to increase their investment in under-graduate science, cnginecrin and math'education.

Thc panel docs recommend that NSF expenditures in thc field%increased by$100 million a year in"Icvcraged" pro-gram support.Some$5.5 million for college instrumentation is thc only program in un-dergraduate education in the NSF budget this year.NSF director Erich Bloch'.is charged irith converting the committcc rec-ommendanons into proposals to be incor-'<porated in net year's NSF budget.~7oHN w~H t y Nuclear Meltdown: A Calculated (and.ruj Recalculated)

Risk For yearsy the nuclear indusuy has been trying to persuade the govcmment to sec a silver lining in thc cloud that gathered over Three Mile Island.Broadly, thc argument is that the 1979 nuclear accident was much less dangerous than oflicial risk cstimatcs would have Ied pcoplc to expect.Therefore, thc risk studies should bc rewritten.

Eventu-ally, if analysis confirms what thc accident at Three Mile Island suggested, safety regula-tions may be adjusted to reflect a calmer view of what would happen in a meltdown.An exercise of this kind has begun at the Nuclear Regulatory Commission (NRC), called the"source terms" review (Scicyyu, 5 April 1985, p.31).The phrase refers to mathematical terms used to calculate leakage from radioactive sources.This project was inspired by the fact that radiation escaping from Three Mile Ishnd was only a fraction of what might have been expected.Also, radioactive iodine was less volatile during thc accident than many had predicted.

Rath-er than venting to thc atmosphere in a pure II hPRII l986 Sc.i ee~t~e'o'e y3ugNEws ic'.cohthIENT I33 J.

~~form, virtually aH of it combined with other chemicals and stayed in thc plant.On 26 March, NRC heard a stafF rcport on the work done so far in the source term renew.Thc NRC staffcrs said they definite-ly could scc a glimmer in thc darkness, but they could not bc sure whether it was thc glint of a silver lining or just another light-ning bolt.Dcspitc their uncertainty, they promised to have some ncw risk cstimatcs ready for publication this fall.Last year, thc NRC released the first draft of a source term document that is meant to serve as the new scienufic basis for work in the area.The report, called NUREG.0956, docs not deal at aH with risks.(These wiH bc calculated in a separate.document duc in October, designated NUREG-1150.)

In-stead, thc scientific document provides de-tailed forecasts of hoiv radioactive chemicals might behave in 16 types of accidents and in six t)~of reactors.When it is complctc in July, it will scrvc as the starting point for thc risk analysis.While the future version of this NUREG report may bc sound, the present edition has been grcctcd with skepticism.

Thc nuclear industry, which has sponsored its own re-search, calls it outdated and alarmist.Thc antinueiear groups sec it as underplaying hazards.And a number of scientists describe it as simply unripe.In this regard, the file of public comments reveals an inhcrcnt prob-lem that may keep the project unripe for a long ume.This is a disagreement over thc credibility of some computer modeling codes that are the basis for aH thc predicuons that will come out of NUREG-0956.

There arc two levels of disagreemcnt.

First, some researchers chaHcngc the codes on a mechanical basis.Thc codes arc so complex, tedious to rcvicw, and obscure, eriucs say, that they have been reviewed by almost no one except those paid to do so, that is, by N RC contractors.

There may bc a hidden bug in thcsc models that no onc has detected.Furthermore, it is impossible to"validate" thc codes fully, for no one is going to stage nuclear accidents to scc how well the numbers represent reality.For this reason, it is important that they be thor-oughly vctted by independent scientists.

Several commissioners stressed this point during thc briefing.Last year, a committee of the thc Ameri-can Physical Society (APS)reviewed some of this work, issued a rcport, and then disbanded-long before the game was over, it turns out.These APS members werc consulted, according to the NRC stafF, bout the final version of NUREG-0956.

But some of the APS group felt the consul-tation was perfunctory and fell far short of Full pccr review.For example, onc member of the APS committee, Fred Finlayson of thc Acrospacc Corporation, wrote to thc NRC in January to explain why he considered thc task unfin-ished.The codes have not been thoroughly peer-reviewed, Finlayson wrote, and their technical assumptions have not bccn adc-quatcly disclosed.

Hc concluded that there werc"too many uncertainucs to provide a rcasonablc basis for revised risk analysis at this time." Nothing has changed his opinion since January.-Another, broader problem with the codes is that they distort natural phenomena by simplifying them.(The codes must bc sim-plified to suit thc computer.)

Thus, knotty problems arc sometimes omitted.Howcvcr, these knotty ones could be important in an accident.For example, onc such hard-to-model event is the scenario in which a molten core interacts with a limestone con-crete floor to produce volumes of gas, heat, and a radioactive aerosoL In thc right cir-cumstances, these fumes could burst through thc containment and pose a serious threat to public health.Indeed, thc codes are inadequate to cope with fuel-concrete interactions, onc NRC official says, because the tcchnical issues arc unresolved.

Research on this topic is now in progress in West Germany and at thc Sandia National Laboratory in Ncw Mexico.Simi-lar unccrtaintics plague thc issues of contain-ment building integrity, high-pressure ejec-tion of fuel from the reactor vessel, hydro-gen production, iodine and lanthanum chemisuy, and rcvaporization of deposited fission products.AH arc being researched.

Ciung the code's deficiencie in dealing with chcmisuy, R.Potter, a Briush official at the Atomic Energy Establishment at Winfrith, wrote of the trcaunent of iodine chemisuy: "At best this is an oversimplification, and at worst, wrong." Unless this and other aspects were improved, hc concluded that it would bc difFIcult to have the necessary confidence in thc results." Thc NRC stafF, induding the acting exec-uuvc director Victor SteHo, assured thc commission that corrccuons and cmenda-tions of document NUREG.0956'ill bc finished by July.Unresolved technical is-sues, such as thc interactions of thc fuel with concrete, will bc handled by setung wide uncertainty margins around relevant terms in the analysis.Work on the risk estimates themselves has already begun and will bc completed within 6 months.Finally, in the bureaucratic tradition, a policy paper issued by StcHo also promised that thc stafF would begin to propose regulatory changes right away, or, in any case,"as soon as the avail-ab!c information warrants such changes." a ELIOT MARSHALL Insurance Drought-Fosters Self-Help Plari for Biotechnology Firms'he insurance crisis that is cu ndy af-fecting a host of industries has not f>assed up biotechnology.

Faced with exorbitant pre-miums and in many instances thc mability to obtain insurance, small biotechnology firms arc turning to insuring themselves.

The As-soeiatihn of Bioteehnoloiiy

/Companies (ABC)plans to sct up an ofFshorc insurance venture'to provide liability coverage to 20 member'zompanics.

/Warren, Hyer, managing director of ABC, says that this plan hopefully:

wiH solve the V member companies'mmediate insurance crisis.Furtflermorc, it also Iiiay pave thc way for thc insurance indusuy tn provide at least limited supplemental underwriting to com-panies for upgrading general liability cover-age, protecting corporate,cxecutivcs and di-rectors as individuals, bringing ncw prod-ucts to market,'or sealinp'up experiments for field and clinical trials.,'nsurance is hard to"gct, says Hycr, be-cause the insuraiicc indusuy"does not know much about bioteehnblogy.

Thc risk right now cannot be identified." But insurers may bc more wiHingaio jake on biotechnology concerns, hc says, after thc association's new at I insurance operation starts functioning.

Dis-cussions with two Ncw York-based interna-uonal brokers-Maarsh 8c McLennan, Inc.and Johnson Bc Higgins-indicate that cov-erage on potentiaI liability claims exceeding Sl million might bc availablc from private insurance companies in the future, says Hycr./'ABC's tentative plan calls for each mem-ber companyl'to bc in-..urcd for liability claims up to/Sl miHior:.Each company would pay an annual prem um of$100,000.Thc companies wiH rcvicii'ach other's re-search portfolios and will esebHsh"a strong risk-prevenu'on program" that sets out gen-eral guideli.ies for thc conduct of rcscarch.Thc affiHati'.

of thc trade assoc iation is likely to be loca".ed in the Bahamas or Bermuda, Hyer indi ated, to avoid U.S.'-tax laws that would tre'at a surplus in thc in':urancc enti-ty's trust'funds as a taxable profit.Thc ir,'surancc crisis cxtcnds to biotcchno-log)~s larger players, including pharmaceuti-cal and;chemical giants."Everybody is hav-ing insurance problems," says Sus'an Racca, an analyst at the Industrial Biota;hnofogy Assodation.

Member companies of rhc IBA arc scheduled to meet next week to discuss a.p s self-Insurance plan.Thc associauon:helvcd thc jdea several months ago but is ta.dng it up again, says Raeca,"because things have go cn so bad." r MAMt CRAvmoari SCIENCEs VOL.232

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Page 1 of 2 AMENDMENT NO.37 3USTIFICATION The Cooling Water Canals and Reservoirs Section is revised to reflect the current design.Table 3.2.1-1, Classification of Structures, Systems, and Components, is revised to delete most references to Note 0 in response to an NRC question and also to reflect the appropriate QA requirements for items classified as seismic.Other minor corrections update information to reflect current design.This section is revised to add an appropriate reference.

Table 3.9.3-10, Non-NSSS Supplied Class I, 2, and 3 Active Valves, is revised to designate a containment isolation valve in the safety injection system as active based on changing its normal position to open and to change information on valves for the RCPB Leak Detection Radiation Monitor as a result of design changes to increase flow and meet particulate sampling requirements.

This section is revised as a result of design changes for the RCPB Leak Detection Radiation Monitor.Table 5.0.13-1, Pressurizer Valves Design Parameters, is revised to provide consistency between the FSAR and Technical Specifications concerning the Pressurizer PORV throat area.This section is revised to reflect design changes to the Containment Heat Removal System.Table 6.2.0-1, Containment Isolation System Data, is revised to show a'afety injection valve as normally open based on results of startup testing and as a result of design changes for the RCPB Leak Detection Radiation Monitor.This section is revised as a result of design changes for the RCPB Leak Detection Radiation Monitor.This section is revised to delete references to a reduced pressure ILRT because this was not used for preoperational ILRT nor will it be used in the future.LLRT changes are made per IE Information Notice 85-71 to ensure determination of"As-Found" Type A Leakage Rate.Also, changes are made to clarify packing leakage and globe valve testing requirements to provide consistency with the preoperational and surveillance test programs.This section is revised to reflect the as-built design of NaOH isolation valve logic.Table 7.3.1-5, ESF Actuation Systems-Safety Injection Signals, and Table 7.3.1-7, ESF Actuation Systems-Containment Isolation Phase A, are revised as a result of design changes for the RCPB Leak Detection" Radiation Monitor.(1092NEL/Aif)

Page 2 of 29.1.3 9.1.0 Table 9.1.3-2, Fuel Pool Cooling and Cleanup System Parameters, is revised to reflect final system parameters for Fuel Pool Cooling Pump flow rate and Total Developed Head (TDH).These values are consistent with those used in final system analyses.Rewording in this section is provided to clarify intent, provide~consistency with plant nomenclature and technical manuals, and correct typographical errors.9.5.1 Commitment to pr ovide on-site air for self-contained breathing equipment is revised to comply with NUREG 0800, 10CFR50, Appendix R and to reflect actual conditions.

9.5.5 12.3.0 13.1.1, 13.1.2 R 13.1.3 This section is revised to provide additional details of.'.he as-built design of the diesel generator cooling water system and suppcrt preoperational testing.Editorial Change These sections are revised to reflect recent management organization changes and provide consistency with Technical Specifications.

13.2.2 10.2.12~~15.6.5 15.7.0 TPe description of the Licensed Operator Requalif ication Training is revised to reflect 10CFR55 requirements.

This section is revised to provide compliance with IE Bulletin 80-06 and to provide consistency between design and testing requirements.

Typographical Error This section is revised to incorporate changes as a result of NRC Technical Specification review-related concer'ns regarding containment ventilation isolation for a fuel handling accident.(1092NEL/Qf

)

0 SKIP FSAR 2.4.8 COOLING WATER CANALS AND RESERVOIRS The safety related cooling water channels (canals), reservoirs, and water controL structures within the reservoir system of the Shearon Harris Nuclear Power Pl.ant consist of the Main Reservoir, the Auxiliary Reservoir>

the Auxiliary Reservoir Separating Dike, the Auxiliary Reservoir Channelxtl,the Emergency Service Water Intake and Discharge Channels'he design bases and operating modes of the reservoir system are described in relation to the safety-reLated Emergency Service Water System, Ultimate Heat Sink, and the Cooling Tower Makeup Water System', these discussions appear in Sections 2.4.11, 9.2.1, 9.2.5, and L0.4.5..Shearon Harris Nucl.ear Power Plant complies with NRC Regulatory Guide).127 (refer to Section 1.8)and Ebasco Specification CAR-SH-CH-24,"Reservoir, Dams and Dike Instrumentation Program (Non-Nuclear Safety)." In addition, the North Carolina Utilities Commission requires a dam inspection program invol.ving private consultants.

As a minimum, the inspection program will include the water-control structures discussed in Section C.2 of Regulatory Guide 1.127.Periodic monitoring of embankment instrumentation will be performed.

The Emergency Service Water Channels and Auxiliary Reservoir are monitored for sediment buildup.The Shearon Harris Nuclear Power Plant reservoir system constitutes the only-water bodies that are of concern regarding protection of plant facilities from fLood and wave runup, discussion of the protection of channels and reservoirs is contained in Sections 2.4.2, 2.4.3, 2.4.4, and 2.4.5.The only locations where potential bLockage is of concern to safe plant operation are the Emergency Service Water Intake and Discharge Channels,~andqkuxiliary Reservoir Channels These channel.s are Category I structures and are designed to remain stable when subjected to the Safe Shutdown Earthquake or the most severe cases of other postuLated natural.phenomena (see Section 2.5.6).In the unlikely event of a sl.ide of the earth slopes, the size of the channels is sufficient to pass the minimum required service water flow at a maximum velocity of 2 ft.per second under the conditions of maximum drawdown of the Main Reservoir and the Auxiliary Reservoir, as indicated in Section 2.4.11.Channel pLans and sections are shown on Figures 2.5.6-6, 2.5.6-7, and 2'.6-8~37 The use of screens for the Emergency Service Water Screening Structure and the Emergency Service Water and Cooling Tower Makeup Intake Structure, the location of the intake structures, and the maximum veLocity of 2 ft.per second in the channels provide assurance that no blockage of the intake structures, damage to the intake structures or damage to the emergency service water pumps can occur.The effects of failure of the AuxiLiary Separating Dike are discussed in Section 2.4.4.II The design bases for reservoir operation during periods of low water level are discussed in Section 2.4.11.2.4.8-1 37 Amendment No.P6 TABLE 3.2.1-1 (Continued)

CLASSIFICATION OF STRUCTURES STSTEMS AND COMPONENTS r Structures Safety Class (1)Code Desi n and Construction and 0 eratlons Quality Quality Code Seismic Quality Class Assurance Class C~ate or (2>Assurance (3>(23>(24>Remarks Diesel Fuel Oil'Storage Tanks and Tank Building NA See>lots (3C>Containment Air Locks, Equipment 2 ASME I I I MC Hatch and Valve Chamber 3 3.33 S Note (29)Containment Internal Structures Containment Crane Supports Cooling Tover NNS E Electr Ical Manholes tor Emergency Pover and Control Cables See Note (30)S stems and Components Reactor Coolant S stem Reactor Vessel" I ASME ill Steam Generator (tube side)(shel I side)I ASME'I I I 2 ASME III Q Q S<<<o<<(~>I~Pressurizer I ASME III

TABLE 3.2.1-1 (Continued)

CLASS IF ICATION OF STRUCTURES SYSTEHS AND COHPONENTS S stems and Com onents Safety Class (I)Code Desi n and Construction and 0 eratlons Code Se I sml c quality Close~Cote or (2)Sssureuue (3)Rsssrks ()ue((ty ()ue((ty Class Assurance (23)(24)Reactor Coolant Hot and Cold Leg Piping, Flttlngs and Fabrication ASHE ill I I Surge Pipe, Spray Pipe, Fittings, and Fabrication ASHE ill I P, See Note (5)4J I Crossover Leg Piping, Fittings I and Fabrication RTD Bypass Hanlfold Pressurizer Safety Valves Pressurizer Power Operated Relief Valves and Block Valves ASHE I I I I ASHE I I I 1 ASHE III 1 ASHE III 1:A Q h P Valves of Safety Class I to Safety Class 2 Interface ASHE III.1 Pressurizer Relief Tank NNS , ASHE VIII Reactor Coolant Thermowell I 5 g'ux 1 I I ary Reactor Coolant 2 Piping (Drains, etc,)Pressurizer Rel lef Valve Discharge 1 Lines (between Pressurizer Nozzle and Relief Valve Only)4v ASHE III 1 ASHE III 2 ASHE III I TABLE 3.2~I-I (Continued)

CLASSIFICATION OF STRUCTURES SVSTEMS AND COHPONENTS Desi n and Construction and 0 eratlons I Remarks S stems and Com onents Safety-Code Class (I)Code Class Seismic~Cate or t2)Pual(ty Quality puallty Class Assurance Assurance (3)(23)(24)Steam Generator Forging Type A Chemical K Volume Control S stem I ASME I I I B A 0 See Note (9)Regenerative HX Letdoxn HX (tube side)(shell side)2 ASHE I I I 2'ASME I I I 3 ASHE III Hlxed Bed Demlneral I zer Cation Bed Demlneral Izer 3 ASHE ill 3 See Note (7)3.'SHE III-3 See Note (7)e Reactor Coolant Filter 2 ASHE I I I 2 Volume Control Tank Charging (High Head Safety Infection)

Pumps 2 ASME I I I 2 ASHE III Charging Pump Hotors Seal Mater InJectlon Filter Seal Mater Return Filter Boric Acid Blender Letdoxn Orlf lees IE 2 ASHE III 2 ASHE III 3 ASHE III 2 ASPIC III B TABLE 3,2'-I (Continued)

CLASS IF ICATION OF STRUCTURES SYSTEHS AND COMPONENTS Desi n and Construction and 0 eratlons I Remarks S stems and Com onents Safety Class (I)Code Code Cl ass Sel sml c~Cate or (2)Quality ()uallty ()uallty Class Assurance Assurance (31 (23)(24)Excess Letdown HX (tube side)2 ASHE III (shell side)2 ASHE III Seal Mater HX (tube side)(shell side)Chemical Hlxlng" Tank 2 ASHE III 3 ASHE III NNS ASME VI II 2 3 Chemical Hlxlng Tank Orlflce Boron Heter NNS NNS ANSI B3I~I Boric Acid Tanks 3 ASME III Boric Acid F I l ter 3 ASHE II I Boric Acid Transfer Pump Boric Acid Transfer Pump Hotors 3 ASME III IE B A Boric Acid Batchlng Tank NNS ASME Vl I I Reactor Coolant Pump (RCP)Standpipe NNS ASME Vl I I RCP Standpipe Orlf Ice RCP Seal Bypass Orlflce I ASHE I I I 8 37

S stems and Com onents TABLE 3.2~l-l (Continued)

CLASSIFICATION OF STRUCTURES SYSTEMS AND COMPONENTS Desi n and Construction and 0 eratlons Puality Quality Safety Code Seismic Quality Class Assurance Class (l)Code Glass~Cate or (2)Assaraaaa (3)(23)(24)I Remarks lA I System Piping and Valves a)Part of RCPB b)Required for reactor coolant letdown and makeup c)Required lor providing boric acid for the letdown and makeup loop d)Normally or automatically Isolated from parts of system covered by a, b or c Instrumentatlon Operators for Safety-Related Active Valves I ASME III 2 ASME III 3 ASME III NNS ANSI B31~I IE IE I 2.B 8 Q Q s~~~A A Q See Note (15)A Q See Note (31)31 37 Boron Thermal Re eneratlon Subs stem Moderating HX (tube side)(shell side)3 ASME III 3'ASME III 3 See Note (7)See Note (7)P Sa~~(2P)Q Sem8&~9 Letdown Chiller HX (tube side)(shell side)3 ASME III NNS ASME V I I I See Note (7)A'~4ccto=~E II" Letdown Reheat HX (tube side)(shell side)Thermal Regeneration Demlnerallzer 2 ASME I I I 3 ASME III 3 ASME III I See Note (7)See Note (7)Q~eke=~Chiller Pump

TABLE 3.2'-l (Continued)

CLASS IF ICATION OF STRUCTURES STSTEMS ANO COHPONENTS Desi n and Construction and 0 eratlons Rssarks S stems and Com onents Safety Class (l)Code'e I sml c Coda Class"~Cats or (2)(Puallty'uality Quality Class Assurance Assurance (3)(23)(24)Chiller Surge Tank NNS ASME Vl I I Chiller Unit a)Evaporator b)Condenser c)Compressor NNS NNS NNS NNS ASHE'Vl I I ASHE Vill E E E E Ll I c System Piping and Valves a)Not normally or automatically Isolated from safety class components b)Other 3 ASHE III NNS ANSI B3l~I 37 Boron Rec cle S stem Recycle Hold Up Tank Recycle Honltor Tank 3 ASHE III NNS AWA D-l00 31 Recycle Honltor Tank Pump Casing NNS ASHE Vill'b Recycle Evap, Feed Pump 8 Recycle Evap Feed Demlnerallzer ft 0 3 ASHE III 3 ASME I I I 3 See Note (7)3 See Note (7)

5 TABLE 3.2'-1 (Continued)

CLASS IF ICATION DF STRUCTURES SYSTEMS AND COHPONENTS S stems and Com onents Safety Class (I)Code Code Class Seismic Rate(ear (2)Quality Quality Quality Class Assurance Assurance (3)(23)(24)Desi n and Construction and Operations Reearka I 24 lm Recycle Evap.Feed Filter Recycle Evap.Condensate Deminerallzer 3 ASME III NNS ASHE VIII See Note (7)B A Q Recycle Evap Reagent Tank NNS.ASME Vill Recycle Holdup Tank Vent EJector 3.ASME III 3 See Note (7)Recycle Evap.Condensate Filter NNS ASHE Vl I I Recycle Evap Concentrate'F I lter NNS ASHE Recycle Evaporator Package VNI E a)Feed Preheater I)Feed Side 2)Steam Side b)Gas Stripper c)Submerged Tube Evap, I)Feed Side 2)Steam Side d)Evaporator Condenser I)Distillate Side 2)Cooling Water Side 3 ASME III NNS ASME VIII ASHE III 3 ASHE III NNS ASHE VIII 3 ASHE III 3 ASME ill 3.See Note (7)3 See Note (7)3 See Note (7)3 See'Note (7)3 I Q 4ee=Hc4e~+

TABLE 3,2,1-1 (Continued)

CLASSIFICATION OF STRUCTURES SYSTEMS AND COMPONENTS S stems and Components Safety Class (I)Code Desi n and Construction and 0 eratlons pballty Quality Code Seismic Quality Class Assurance Class~Cata or>2>Assurance (3>>2>>tra>Raaarks e)Distillate Cooler I)Distillate Water Side 2)Cooling Water Side f)Absorption Tor)er g)Vent Condenser I)Gas Side 2)Cooling'Water Side h)Distillate Pump I)Concentrate Pump J)Piping I)Feed 2)Distillate 3)Concentrate 4)Cool lng 5)Steam k)Valves I)Feed 2)Distillate 3)Concentrate 4)Cooling 5)Steam 3 ASHE III 3 ASHE III 3 ASHE III 3 3 3 3 NNS ASHE III ASHE III ASHE III ASHE III ANSI 831'3 3 3 NNS ASHE III ASHE III ASHE III ASHE Ill ANSI 831'3 ASHE III 3 ASHE III 3 ASHE III 3 ASME III 3 See Note (7)3 I 3 See Hote (7)See Note (7)I See NCTe (7)See Note (7)3 See Note (7)3 See Note (7)3 See Note (7)3 I 3 See Note (7)See Note (7)3 See Note (7)3 I 8 8 8 8 B.8 8 8 e Note (Note (3'A 4/I 4'l ee Note (4 ee Note (4 ee ot (4 ee T (4~A~p A echo e (4 eeh a I-'a T ee cT (4 PA QQ e Nota 4 P'A~()I)e Note (P'A.P'e Nota (

TABLE 3,2.l-i (Continued)

CLASS IF ICATION OF STRUCTURES SYSTEHS PAD COMPONENTS S stems and Components Safety Class (I)Code Desi n and Construction and Operations Quality Quality Code Seismic Quality Class Assurance C(ess~Cate or (2)Assurance (3)(23)(24)Reearks System Piping and Valves a)Not normally or=automatically Isolated from safety class components b)Other 3 ASHE III NNS'.ANSI B31~I 3 See Note (7)Q Safet In ection S stem Accumulators 2 ASHE III Boron Infection Tank (BIT)2 ASHE III B Hydro Test Pump System Piping and Valves a)Part of RCPB I ASME III Q Q

TABLE 3.2.1-1 (Continued

)CLASS IF ICATION OF STRUCTURES SYSTEHS AND COHPONENTS S stems and Com onents Safety Class (I)Code Desi n and Construction and Operations Quality Quality Code Seismic.Quality Class Assurance Class~Cate or (2>Assurance (3>(23>(24>Remarks 43 I t4 c)Piping and valves required for 2 performance of satety tunctions of SC2 components and which are not In service during any normal mode of plant operation and are not testable d)Operators for Safety-Related IE Active Valves ASHE III Reactor Coolant Drain Tunk Ht~Exchanger (shell side)2 ASHE III 2 I Q See Note (31)Instrumentatlon IE A.Q See Note ()5)Containment Penetration Pressurlznt(on S stem System Piping and Valves Connected to Penetrations 2 ASHE III 37 Instrumentat ion NNS Waste Process ln Bul l din (WPB)Cool in S stem WPB Cooling Pumps NNS Heat Exchnnger (tube d shell side)IINS ASHE VIII n>Piping and Valves NNS ANSI B31,1 0

TABLE 3.2'-1 (ntlnued)-CLASSIFICATION OF STRUCTURES SYSTEMS AND COMPONENTS Desi n and Construction and 0 eratlons Remarks S stems and Com onents Safety Class (I)Code Code Se I sml c Class~Cere or (2>-Quality,'uality Quality-'lass.Assurance'-Assurance (3).(23)(24)Fuel Pool Coolln and Cleanu S stem Fuel Pool Heat Exchanger (tube side)(shell side)3 ASME Ill 3 3'ASHE I II 3 8 8 A~A~Q Fuel Pool Cooling Pumps.3.ASHE.I I I', 3 Fuel Pool Cool lng Pump Hotors'.IE Fuel Pool Demlneral lzer Filter.-NNS.ASME VI I I E Fuel Pool Demlnerallzer Fuel Pool Refueling Water, Purl f ication F I i ter NNS ASME Vl I I'NNS ASME Vl I I.," E~Fuel Pool Stralners 3 ASHE I I I.8 Fuel Pool Sklmmer Filters L Fuel Pool Sklmmer Pumps'NS:-ASME Vll I~'NS r Fuel Pool and Refuel lng Water.NNS'urification Pump E E P-E Fuel Pool Skimmers.-'NNS Fuel Pool Liner NNS 8-Q See Note (21)Fuel Pool Nozzles 8 Q See Note (21)and (21A)System Piping and Valves a)Required for cooling and makeup to the fuel pools b)Hakeup from RWST.3 ASHE III 3 ASHE III 8 8

'TABLE 3 2,1-1 (Continued).CLASS IF ICATION OF STRUCTURES SYSTEHS AND COHPONENTS Desi n and Construction and 0 eratlons I Remarks S stems and Com onents Safety Class (I)Code Se I sml c Class~Cate or (2)Quality Quality Quality Class Assurance Assurance (3)(23)(24)c)Required for fuel pool cleanup NNS and normally Isolated from a)ANSI 831'Instrumentatlon IE Q See Note (15)Fuel Handiin S stem Hanipulator Crane E Reactor Vessel Internals Lifting Device Rod Cluster Control Changing Fixture Reactor Vessel Stud Tensloner NNS E Spent Fuel Handling Tool Q.See Note (10)Fuel Transfer System a)Fuel Transfer Tube and Flange 2 b)Portions of Conveyor System and 3 Controls ln Fuel Handling Building c)Remainder of System NN ASHE I I I B.B h A Q.See Note (ll)Q See Note (12)New Fuel Elevator New Fuel Racks Portable Underwater Lights TABLE 3,2.ntlnued)CLASSIFICATION OF STRUCTURES SYSTEMS AND COMPONENTS Desi n and Construction and 0 eratlons Remarks S stems and Com onents Safety Class (I)Code New Fuel Assembly Handling Fixture NNS Code Class Sel sml c~Cote or ttt I Quality Quality Quality Class Assurance Assurance (3)(23)(24)New Rod Cluster Control Handling NNS Fixture Lower Internals Storage Stand NNS Upper Internals Storage Stand Load Cel I Linkage Spent Fuel Storage Racks Refueling Cavity Seal Ring Instrumentatlon IE A Q See Note (15)LI uld Waste Processln S stem NNS See Note (25)See Note (25)-Reactor Coolant Drain Tank Pump NNS ASME III Reactor Coolant Drain Tank Heat Exchanger (shell side)2 ASME I I I (tube side)NNS ASME V I I'I B.p'-sot 6'oto CeQ (37 System Piping 8, Valves a)Not normally or automatically 3 ASME III Isolated from SC-3 components b)Other NNS 831,1)37 Gaseous Waste Processln S stem Gas Compressor Gas Decay Tank 3 ASME III 0 TABLE 3,2,1-1 (Continued)

CLASSIFICATION OF STRUCTURES SYSTEHS AND COHPONENTS S stems and Com onents Safety Class (1)Code Desi n and Construction and 0 erations Quality Quality Code Seismic Quality Class Assurance Glass~Cate or (2)Assaraaoa (3)(23)(24)Rasarks Hydrogen Recomblner (Catalytic)

NNS ASHE I I I System Piping and Valves a)Not normally or automatically 3 isolated from SC-3 component b)Other NNS'31~1 Solid Waste Processln S stem Containment Cool in S stem NNS See Note (26)See Note (26)a e.Note (27)3 Containment Fan Coolers a)Fans and Casings b)Supply Fan Hotor c)Cooling Coils d)Ductwork and dampers up to concrete alrshafts e)Ductwork and dampers downstream of concrete alrshafts 2 IE 2 2 NNS ASHE I I I B 4 8 B B A A A A Q Q Q Q 422 a<nz((8$j 82 Containment Fan Col I Units Instrumentlon IE B ,~<<m<<sj (ar Q See Note (15)Containment Ventilation S stem Airborne Rad I oact I v I ty Removal NNS System E P-See Note (IS)

TABLE 3,2 l-l (Continued)

CLASS IF ICATION OF STRUCTURES SVSTEMS AND COMPONENTS 4 Desi n and Construction an3 0 eratlons Remarks S stems and Com onents Safety Class (I)Code Quality Qual Ity Code Selsmlc Quality Class Assurance C(ass~Cate or (2)Assurance (3)(23)(24)~es+CROM Cool lng Systems Containment Combustible Gas Control S stem/'B Electric Hydrogen Recomblner Instrumantatlon (In part)'2 IE B B A A.Q Q See Note (l5)Hydrogen Monitoring System (0-IOS a)Piping and Valves b)Hydrogen Analyzer Cabinet c)Remote Control Panel d)Remote Sample Dilution Panel range capability) 2 ASME I I I IE IE NNS A A'A E Q Q See Note (l5)See Note (I5)Containment Vacuum Relief (except blind flanges and valves for leak testing)ZivCET 8eDWe27iI Instrumentation 2 ASME III 3 jl~Ill IE Primer Shield Cool ln S stem Instrumentatlon Reactor Su orts Coolln S stem 3 Instrumentatlon s TABLE 3.2,1-1.(Cont inued)CLASS IF ICATION OF STRUCTURES SYSTEMS ANO COMPONENTS S stems and Com onents Safety Class (1)Code Code.Se I smic Class'~ete or (2)Quality Assurance (3)Desi n and Construction and Operations Raaarks Quality Quality Class Assurance (23)(24)Reactor Auxlliar Bulldin (RAB)..Ventilation S ste neo foR~g AH hJ I lus2 RAB Normal Ventilation System a)Isolation dampers b)All other components RAB Steam Tunnel Ventilation

'NS A E RAB Emergency Exhaust System-3 A RAB ESF Equipment Cooling Syste s 3 ESF RAB Batter R xhaust Fbns 3 y RAB Computer d Communications Room HYAC>r<WA ToRAHDO PR~TEC7XOP/

DA~PE'R>RAB Sultchgear Room Ventilation System Ir'vcrvdm a)Smoke purge solatlon valves Kcup b)Smoke purge i o at dampers~A s 0 RAB Electric Equipment Protection 3 Rooms Ventilation System>litic(Mig a)HV-equipment room ex a s b)SmOke purge iSOlatiOn valveS h~d HhmPE<>Instrumenta 2on IE P See Note (15)

TABLE 3.2.1-1 (Continued)

CLASS IF ICATION OF STRUCTURES SVSTEHS AND COHPONENTS S stems and Components Maste Processin Buildin Desi n and Construction and Operations Puallty Puality Safety Code Seismic puallty Class Assurance Claaa (1>COde Cleea~Cate Or (2>eseaeraeee (3)(23>(24)NNS Reearas Sff ivy%C+)Ventilation S stems MGC~ct T~~R~~u lfACk 4>ieFW l~l Coo~a Control porn HVAC S stems pd(d e'X A(AS j Normal Supply Subsystem a)Supply Fans d Casings b)Cooling Coils c)Electric Heating Coils d)Ducts and Dempers e)Valves for Outside Air Intakes f)Chlorine d Radiation Detectors g)Smoke Detectors 3 3 IE 3 3 IE NNS ASHE III ASHE ill 3 Control Room Smoke Pur e and Exhaust~up a)Boundary Isolation Valves b)Other 3 ASHE ill NNS p~~i37 OFF Ikon (/Sg I Control Room Emer enc Filtration S stem Instrumentation IE p See Note (I5)Fuel Handlin Buildin HVAC S stems Air Conditioning System for the Operating Floor a)Air Handling Unit NNS f2)OFF f fKTHgNAIIFA f&r<<AFAT&e QoICs i)&An W~JAILS,OuCT>

~d'Op~pF+S iVNS SSy uorS(S8$sar A61i(J8$A TABLE 3.2.l-l (Continued)

CLASS IF ICATION OF STRUCTURES SYSTEMS AND COHPONENTS Desi n and Construction and Operations Remarks S stems and Com onents Safety Class (I)Code Code Cl ass Seismic Sateraor (2)Puality Puality Quality Class Assurance Assurance (3).(23)(24)b)Exhaust Fans c)Ductwork and Dampers I)Isolation,Dampers 3)Other Ouch Unl gen y x aust System for the Operating Floor.3 NNS E A E 0 smzN Ti(IBJ Normal Ventilation System for Areas Below Operating Floor a)Air Handling Unit b)Exhaust Fans c)Ductwork and Dampers I)Isolation Dampers 2)Other Spent Fuel Pump'oom Ventl I ation System NNS NNS 3NNS 8 E E A..P 2~.s(5 R(Ii(IBJ A P Instrumentatlon Fuel Ol I Transfer Pum House Ventilation S stem Diesel Generator Bulldin Ventilation S stem a)DGB-Electric Room Ventilation b)DGB-F.O.Day Tank and Silencer Room Ventilation

)o(B-0 LG~&7o R Vzw7i'67ioi sy~7im.-IE-.3 I a S.()See Rote ((5)I~A P A~P

TABLE 3,2'-I (Continued

)CLASSIFICATION OF STRUCTURES SYSTEMS AND COMPONENTS Desi n and Construction and Operations Remarks S stems and Components Safety Class (I)Code Code Se I sml c Class~Cate or (2)Quality Quality Quality Class Assurance Assurance (3)(23)(24)IO I 4J Ln Chilled Mater Piping and Valves a)Required to provide chilled uater to safety related air handling units b)Required only for RAB NNS Ventilation Systems and automatlcaliy isolated from a)c)Operators for Safety-Related Active Valves 3 ASME III IE Instrumentation IE Non-Essential Services Chilled NNS~aster S stem A stXPo7E (I8$)~3/g K See Note (l5)Vl spy/Vole'(18/

Containment Atmos here Pur e and M~akeop S stem Ductwork Inside Containment

+~F to the isolation valves Containment isolation valves and piping P 3.2 ASME III jB pA/e A')M 37 0 From I sol at Ion va I ves outs I de'Containment to floor pene-tration at RAB Elevation 286 ft (puagEN4KEuP)

~i d RAB H PQ a;Cr~r S A~Kg s umen ion (iso a n E valves only)Other~NNS A 3i Q See Note (l5)sa P,f~ggs TABLE 3,2.1-1 (Continued)

CLASSIF ICATION OF STRUCTURES SYSTEHS AND COHPONENTS S stems and Components Safety Class (I)Code Code Se I sml c Cleat~Cate ar (2)Qudllty Puallty Ouallty Class Assurance Assurance (3)(23)(24)Desi n and Construction and Operations fiemarks lm lm Operators for Safety-Related Active Valves apl A AT Containment N dro en Pur e and-IE Q See Note (31)Containment Isolation valves and piping F rom I so I at i on va I ve outs I de Containment to floor pene-tration at RAB Elevation ft.Instrumentatlon (isolation valves only)2 ASHE III IE A.Q q See Note (15)37 Other E.-Szp Amok (I8$R.37 6 0 5)8~A R1~1~KXS~R uTAIurnF~7 Y~Cuu~2)F ANO PffR6<S)'Slt g s 0 C ToR n/E 80Jldi~ate SySggRIS NdS

TABLE 3.2.1-1 (Continued)

CLASS IF ICATION OF STRUCTURES SVSTEMS AND COMPONENTS Desi n and Construction and 0 eratlons Remarks S stems and Com onents Safety Class (I)Code Code Seismic Class~Cate or (2>Quality Quality Quality Class Assurance Assurance (3)(23)(24)b)From the MSIV up to and including the last seismic restraint In the Turbine Building c)Downstream of last seismic restraint in Turbine Building d)Operators for Safety-Related Active Valves e)Turbine Gland Sealing System See Note (16)NNS.ANSI 831~I IE B31~I See Note (16)A Q See Note (31)C Instrumentatlon Steam Generator Slowdown S stem IE A Q See Note (15)System Piping and Valves a)From steam generator to.and including containment isolation valves b)From containment isolation valves to RAB wall 2 ASME II I 3 ASME III Condensate and Feedwater S stem h4 0 Condensate and Feedwater Pumps E I ectromagnet I c F I I ter Condenser Evacuation System NNS NNS ASME Vl I I NNS 831~I (27)C R See Note (27)

+hf IntFIVcjffoM

'I/AJVE'AGg iI 4l47 gfigiekiC gE5~AIRT)iW fu~62~F ger)LCh~g d)up7f,rW7;I4ZS.7 i'uabuK 8w~cfieg~/V5 Sff ATE SFFIVO1F (gg (it J AWE AMBI 631 I TABL 3.2.I-I (Cont lnu CLASSIFICATION OF STRUCTURES.SYSTEMS AND COHPONENTS S stems and Components Safety Class (I)Code Desi n and Construction and 0 eratlons Code Class Se I sml c C~ate or<2)Reearka Quality Quality Quality Class Assurance Assurance (3)(23)(24)System Piping and Valves')Feedwater piping from the steam generator back to and Including the HFIV check valve;all branch connections from this section up to and Including the first normally closed shutoff valve b)HFW control valves and bypass control valves;-44ew-2'SHE-III 3~S.ASHE I I I ,See Note (4)c)t)~Operators for Safety-Related IE Active Valves, Q See Note (3l)Instrumentation IE Q See Note (I5)Auxlllar Feed22ater S stem AFW Pumps (Hotor d Turbine Driven)3 ASME III B A;Q AFW Pump Motors Condensate Storage Tank AFW Pump Turbine Driver IE 3 ASHE III 3 ASHE III Q See Note (2S)System Piping and Valves a)From steam generator up to 2 and Including the containment isolation valves ASME I I I

TABLE 3.2~I-l (Cont.lnued)

CLASSIFICATION OF STRUCTURES SYSTEHS AND COMPONENTS S stems and Com onents Safety Class (I)Code'ode Se I sml c Cl less~Ceto or (2)Puality Assurance (3)Desi n and Construction and Operations Remarks Pual Ity Puality Class Assurance (23)(24)b)Other c)Operators for Safety-Related Active Valves 3 ASHE III IE A A P See Note (31)Instrumentation s IE P See Note (15)Condenser C 1rcu I at In Water~Sstem Demineralized Water Stora e~et tees Demlnera1 Ized Water Storage Tank NNS E'eactor Hake-up Water Storage Tank 3 ASHE III 37 Instrumentatlon (in part)IE See Note (15)Reactor Hake-up Water Pump, Pipes/3 ASHE III Valves Reactor Hake-up Water Pump Hotors NNS Chlorine Leak Detection (In part)IE A P Radiation Honitorin S stem Safety Area Monitors IE See Note (15)

TABLE 3.2.l-l (Continued)

CLASSIF ICATION OF STRUCTURES SYSTEHS AND COHPONENTS S stems and Com onents Desi'n and Construction Safety Code Seismic Class (ll Cods Class C~sts or (2>and Operations Rssarks Quality Quality Quality Class Assurance Assurance (3)(23)(24)Piping and valves up to and including second isolation.

valve All other piping ASHE III 2$31.'1 NNS X Instrumentatlon IE Inadequate Core Cooling System.IE (in part)Q See Note (15)Q See Note (l5)Associated piping and valves ASHE III 2 A SHNPP FSAR Notes to Table 3.2'-l (Continued)

(18)Those portions of this system whose failure may have an adverse effect on a nearby safety related component are seismically supporte Avo sos~icdDy a'f'AS~/tv g~s Jg~70 f/ir Ro gi<5 gg Rq)Mpj (19)The reinforced concrete mat and walls of the Unit 1 Turbine Building between column line 42 (approx.)and 43 (approx.)are designed and constructed to Seismic Category I requirements due to the presence of the diesel'generator service water pipe tunnel and Class 1 electrical cable area above the pipe tunnel (see Figure 1.2.2-60).

This area is designed and constructed to withstand the coLlapse of the Turbine*Building concurrent with a SSE.(20)Provides mechanical support Eor Safety Class 1 component.

(21)Mill be designed and fabricated to the applicable portions of ASME III, although it is not classiEied as ANS Safety Class 1, 2, or 3.(21A)Fuel Pool Nozzles wiL1 be considered from the Fuel Pool Liner to the first shop girth weld.(22)Provides support to the Safety Class 1 pressure boundary conduit.(23)Quality CLassification (Operations Phase)e A" Safety'related.<<~P'~~'~B-Non-safety seismic or falls under Regulatory Guide 1.97.C-Radwaste.D-Fire protection.

asap'~+<~E-Non safety, non-seismic.

(24)Quality Assurance Requirements (Operations Phase)Q-QA requirements will meet 10CFR50 Appendix B criteria.R-QA requirements will meet ETSB ll-l QA requirements as a minimum.Optionally"Q" requirements may be imposed.F-QA requirements will meet Fire Protection QA requirements as a minimum.Optionally"Q" requirements may be imposed'A requirements of 10CFR50 Appendix B are not mandatory.

(25)The cod'e and code class Eor individual components in the Liquid Waste Processing System can be found on Table 11.2.1-7.(26)The code and code class for individual components in the Solid Waste'rocessing System can be Eound on Table 11.4.2-4.(27)The ETSB 11-1 QA applies to components listed in Table 11.4.2-4 except those listed as manufacturer's standard.(28)ASME III Code applies to oil cooler and trip/throttle valve only.(29)Not Stamped.3.2.1-47 Amendment Nn.~J N SHNPP FSAR direction.

Two additional sets of statistically independent accelerograms, developed for the east-west and vertical directions, are presented on Figures 3.7.1-25 through 3.7.1-28..A comparison of the spectral values of the SSE statistically independent horizontal east-west and vertical time histories, and the corresponding design, response spectra, is presented on Figures 3.7.1-29 through 3.7.1-34, for two, four, and seven percent damping, using the frequency intervals discussed above.The comparisons discusse4 above show that none of the points fall below ten percent of the design response spectrum, and no more than five.points fall below the design response spectrum.A The earthquake accelerograms used in the analysis of the Seismic Category I dams and dikes envelop the horizontal and vertical design response spectra presented on Figures 3.7.1-5 through 3.7.1-8.Figures 3.7.1-35 through 3o7o1-37 show the SSE horizontal accelerograms for one, two, and five percent damping, To demonstrate that these time histories envelop the design response spectra, a high resolution response spectra analysis was performed.

Each time history was analyzed at 247 discrete period points between the period range of 0.014 to 3.000 sec These period points were spaced at 0.0005 sec.intervals at the short pe'riod end and at O.l sec.intervals at the long period end.These period intervals were established by performing response analysis at both half resolution (124 period points)and full resolution (247 period points).It was found that there was essentially no change in the general shape of the response spectra.Therefore, these 247 closely spaced period points are considered to be sufficient to detect all the peaks and valleys of the response spectra.Comparison of these time histories with the horizontal design response spectra for the SSE are indicated on Figures 3.7.1-38, 3.7.1-39 and 3.7.1-40, for one, two, and five percent damping, respectively.

3.7.1.3 Critical Dam ing Values The damping ratios, which are expressed as percentages of critical damping and used in the dynamic analysis of Seismic Category I structures, are consistent with those of Regulatory Guide 1.61, and are shown in Table 3.7.1-1.For the Seismic Category I Main Dam, Auxiliary Dam and Auxiliary Separating Dike, the seismic analysis is presented in Section 2.5.6.For the Seismic Category I reactor coolant.loop system, Seismic Category I piping systems, and Seismic Category I equipment not purchased as of March 1, 1977, the SHNPP complies with the damping values of Regulatory Guide 1.61.In accordance with the provision of Regulatory Position C2, documented test data have been provided to and approved by the NRC which justifies the use of a damping value higher than three percent critical for large piping systems under the faulted condition.

A conservative value of four percent critical has been justified by testing.for the Westinghouse reac o nt loo as resented in WCAP-7921-AR"Damping Values of F<r Secs~<'c C~teq~g<c~Q e+r~a~))y<~r, da~piwg v a4ao z p+" ec4+cl Vomer Corqaro4oa Cc ale>rc g ow)Couku'i4~ewkVrogcaM (Report+1+$3-2(.(-~)'~+o 4e.'~1~>)(p~g A~e~k~e~t Qo.E1 TABLE 3.9,3'-l4 (continued)

NON-HSSS SUPPLIED CLASS I 2 AND 3 ACTIVE VALVES~ta llueber~tutee.Env, Looatloa gual.~T Oe Operator uaaufaaturer Safety Class Valve Design Rating (ANSI S)System Design Size Conditions (Inches-ID)

Function tA I Vl'tet 0 I CS-V711SN CS ICS-V70SN 2CS-V129SH CS RCB (4)Check RCB (4)Check RAB (3)Check hP hp Rockwell Rockwell Rockwell 152 I I 1521 2 1500 2485 pslg 8 650 F 2485 pslg 8 650 F 220 pslg 8 200 F RCPB Boundary RCPB Boundary Safe Shutdown 3CS-V222SN'S 3CS-V223SN CS RAB (3)Check RAB (3)Check hp Rockwell Rockwell 3 1500 3 1500 150 pslg 8 250 F 150 pslg 8 250 F Safe Shutdown Safe Shutdown ISI-V39SA V45SB V51SA SI RCB (4)Check Rockwell I 1521 2485 pslg (I 650'F RCPB Boundary Q I S I-V63SA V69SB SI V75SA 0 g~-u~2, SZ Mg RCB (4)Check RA&-Cl~L<Rockwell I 152 I 2485 ps lg ii 650 F Copes-Jwlcc a Q (5'oo Zoo ps ig p 2ooF RCPB Boundary CO~4Cal&MCMW WS OL O$I~

TABLE 3.9.3-14 (continued)

NON-NSSS SUPPLIED CLASS 1 2 AND 3 ACTIVE VALVES~ta N eaer~eatee 3SM-V870SA-I SM 3SM-V871SB-I SM 2CS-V136SN CS Valve Design Rating (ANSI I)3 600 RAB (3)Check hp Rockwell 3 600 RAS (3)Check Rockwell 2 1500 Env, Safety Loaatlou Qual.~T e 0~orator Maoufaoturar Class RAS (3)Check hP Rockwell System Design Conditions 150 pslg 8 140F 150 ps ig&140F 2735 pslg 8 200 F Size (Inches-IO)

Function I ESF Operation ESF Operation ESF Operation lA I Ln pc'CS-V137SN CS 2CS-V138SN CS 8 lgtSA 25P-IC308SS-I SP v NSOSh 2SP-~OSS-I VMSISB 2SP-~NB I SP SP V'W'I SQ 2SP-~&I SP RAB (3).Check RAB (3)Check RCB (5)G I obe RAB (3)Globe RCB (5)G lobe RAS (3)G lobe Rockwell hp Rockwell Solenoid Target-Rock Solenoid Target-Rock Solenoid Target-Rock Solenoid Target-Rock 2 1500 2 1500 2 600 2 600 2 600 2 600 2735 pslg 8 200 F ,2735 pslg 8 200 F 90 pslg 8 400 F 90 pslg 8 400 F 90 pslg 8 400'F 90 ps lg 8 400 F ESF Operation RcqsLe 4'b t.RM.Atty%ttr RCPT.Lc 4b gc.h.Ho~itpr%31 RC'PB Lac.k'b cf gcuh Ho l4r gceB L V.be QQ.Ho~t4r A~cchptcs K ESF Operation o~3CH B2SA I ESCMS Supply RAB (3)Sutter f I y Ol aphragm ITT/Hamme I Dahl 3 150 150 ps I g 8 125 F Isolation 3CH-B4SS-I ESCMS RAB'3)Butterfly Diaphragm ITT/Hammel Oahl 3 150 Supply 150 ps I g 8 125 F I so I at Ion SHNPP FSAR After collection in the containment sump, the'collected leakage is pumped to the floor drain collection tank.The combined sump pump discharge flow is recorded in the Control.Room.The sumps are also provided with level switches to alert the operator of high level conditions in the event of sump pump maLfunction The sump discharge line may be sampled from outside of the Containment to provide additional aid in identifying the leakage source.The system is designed to permit calibration and operability tests during plant refueling.

5.2.5.3.2 Containment Airborne Particulate and Gaseous Radioactivity Monitoring 37 The containment atmosphere radiation monitor is part of the safety related portion of the Radiation Monitoring System and is designed to provide a continuous indication in the Control Room of the particulate and gaseous radioactivity levels inside the Containment.

Radioactivity in the containment atmosphere indicates the presence of fission products due to a reactor coolant system leak.s~Egpol~~The monitor draws a continuous sample of containment air through a~~.located inside the Containment.

ampled'oi s in the onta n nt are at e nort actor c sty, sou reactor c ity, above e ch of the three stea generato , above ach of t three rea or coolant p ps, and ove the pr ssurizer Normal , all po ts (except he pressuriz r)are clo d;on det tion of i h'i n a i The guidelines of ANS-13.1 have been followed to minimize biasing the particulate portion of the air sample'll sample lines are heat traced outside the Containment to prevent condensation within the LLnes up to 120 F and 100 percent humidity (non-condensing)

~zzrsmN Ct oayT~~d RGP8 The monitor uses the airborne part culate an n ble gas de ctor described in Section 11.5.2.6.5.The containment monitor is powered by the A bus.The monitor normally monitors the containment atmosphere for eakage as required by Regulatory Guide 1 45.A containment isolation signal will.isolate the monitor from the Containment.

The monitor provides a high radiation alarm when concentcations reach preset limits.The receipt of this alarm will alert the operator to the presence of low level leakage so that pppggpggh$

AcfjoV can be dane in order to locate the lea age source/i~<iw1iilE p~~~pusgut<o4fiou~pm/PREsET Jiminy AjzE Ek~za'E'd, 5.2.'5-6 Amendment No.~Jg

SHNPP FSAR TABLE 5.4.13-1~PRESSURIZER VALVES DESIGN PARAMETERS Pressurizer Safet Valves Number Haximum relieving capacity, ASHE rated flow (lb/hr)Set pressure (psig)Design temperature (F)Fluid 380,000 2485 650 Saturated steam Transient Condition (F): Non-Faulted Conditions Faulted Conditions 673 682 Backpressure Normal (psig)Expected during discharge (psig)Throat Area (in)3 to 5 500 3.64'Pressurizer Power 0 crated Relief Valves Number Design pressure (psig)Design temperature (F)2485 650 Relieving capacity at 2350 psig, per valve (lb/hr)210,000 Fluid Saturated steam Transient condition (F): Non-Faulted Conditions Faulted Conditions 673 682 Throat Area (in)Pressurizer S ra Valves Number Design Pressure, psig Design Temperature, F Design Flow, for valves full open, each, gpm'485 650 350 5.4.13-3 37 Amendment No.W SHNPP FSAR~a~~4aa~Wc Co&kcls&McP+

cLuerc qe+e~pcra+~<<be'to~

('~o F c)During normal operation, the CCS is designed t:o 1 en the service water temperature is 90 F or below, ewo o e four sa related fan cooler units will operat:e wit:h bot ans per unit operating full speed along with three non-s y fan-coil units 37 2)When service water teraperat s above 90 F, fn addition to the operation of safety and no a ety coo unfes as discussed in 1)above, both standby eey related fan cooler s wf.ll be energized to operate wi ne fan per'nf.t: running at full spe Operation of standb cooler units is an'ticipated approximately 370 s a d)Nixing the containmeht atraosphere following an accident.Design Description 6.2.2.2.1.2 The CCS consists of four safety related fan cooler units and three non-safety fan coil units.Following a design basis accident only t: he safety related fan cooler units are required to operate.During normal power operation, safety related units operate in conjunction with the non-safety units t:o maintqfn required containment temperature.

'See Table 6.2.2-1 for major system components.

Figure 6.2.2-3 describes the extent of essential portions of t: he ductwork and equipment for the CCS.v~~<<<~"g" pTwo of the four safety related fan cooler units are located at Elevation 236', the remaining two safety related units are located at Elevation 286'.37 Two separate trains are provided, each"conqfsefng of two.fan cooler units with~each unit supplying ai,r to an independent, veref.cal concrete afr shaft.Train A Com onents'rai.n B Corn onents Fan Cooler Fan Cooler Service Water Emergency Power AH-2 AH-3 Loop A'Diesel A Fan Cooler--'Farl Cooler Service Water Emergency Power AH-1 AH-4 Loop B Diesel B Train selection of each fan cooler with" fts respective water supply is under administrative control.Each fan cooler is served by water from the Service Water System.A detailed descrfptfon of the Service Water Sys e s in o 9 Um'4 er0or~wcc 3a4c iS S4o~m i laic (Z2")Each safety related fan cooler consises of coo in coi.l sect ons and two direct driven vane axial flow fans 37 Each fan is equipped with a two.speed motor enabling half speed operation a prevent ai.r flow t.n the reverse direction when only one fan per unit is required to operate.Both fans of ehe unf.t dfschar e f.nto a comraon CD~&~'p~o Q~mP t~kegr~,red Lea.4 ra,Qg+e5$+~Mdk abKo$.G.Z.2-<Me~r SlM'P FSAR TABLE 6.2.2-1 CONTAINMENT COOLING SYSTEM COMPONENTS NOTE: All air quantities=are actual cfm.CONTAINMENT FAN COOLER SAFETY CLASS 2 UNITS No.of Units Normal Operating Conditions 2 fans per unit and 2 units operating Design Basis Accident Conditions 1 fan per unit half speed, 4 units starting and 2 units operating Fan Cooler Unit Operating Capacity.ACFM Actual Air Mixture Flow (ACFM)at Fan Inlet 125,000 62,500 31,250 31,250 Design Ambient Pressure, psig Ambient Temp, F Total Pressure, in.WG Fan RPM Outlet Velocity, FPM Brake HP Motor HP J Cooling Water Flow-GPM 120 7'1770 5800 101.2 125 1500 45.0/39.1 ())258 5'870 2560 32.8 62.5 Entering Water Temp, F 95 NOTE: (1)39.1 psig-steam line break pressure 45.0 psig" maximum containment design pressure 6.2.2-16 37 Amendment No.M

SlWPP FSAR+4rd~4 joh..h~pe~id A branch duct connection has~~pprovided to serve as a post accident discharge nozzle and is normally isolated by means of a separate pneumatically operated, fail open damper.~nserg 4'ro~Y~e 4 2-2"5 6.2.2.2.1.F 1 Post Accident Operation During post-accident operation, four Ean cooLer units operate with one Ean per unit running at half speed.The system can operate in this mode as long as both.emergency diesel generators and both service water system trains are available.

In the event of failure of one of the emergency diesel generators or one~The damper in the post-accident discharge branch duct will be opened.The post-accident discharge duct is provided with high velocity nozzles to diffuse air to accelerate the temperature mixing inside containment.

These nozzles are directed to selected areas of heat release, to achieve thorough mixing oE containment atmosphere'he high velocity nozzles direct turbulent air jets from discharge points at two levels inside containment where two separate trains of containment fan coolers are located.Two'ets of nozzles are located at Elevation 286 Et's shown on Figure 6.2.2"14, Sections C-14-1 and C-12-1, and'he other two nozzles are shown'on Figure 6.2.2"10 (plan at Elevation 221.00 ft.)as post accident discharge nozzles.Seismic Category I ductwor'k is used from the fan coolers to the discharge outlets.As the post-accident containment atmosphere steam-air mixture passes through the system cooling coils, it is cooLed and a portion of the steam is condensed.

The combined cooling capacity of all four cool.er units is adequate to prevent excursions beyond the peak design pressure and temperature of the Containment; however, in the event of a single active failure in one train, one containment spray pump and two containment fan coolers will provide the adequate cooling capacity.The fan cooler units receive electric power from the diesel generators 15 seconds after a LOCA through a timer-sequencer.

An additional 10 seconds are requir'ed to bring the fans to the operational speed."-The'containment.fan cooler performance data, showing the energy removal rate as a function of containment atmosphere temper'ature, is shown on Figures 6.2.2-4 and 6.2.2-S and Tables 6.2.2-2 and 6'2.2-3.6.2.2.2.1.2.2

+taPog d S d.l(PORC'F During normal power operation, three non-safety fan coil units are in co&tvuous operation along with~safety-related fan cooler units.The following describes.

their operation'.

CO~)catsshma,~p Otu et'Cay 8 l IS a)When'emperature is 40 p or below: Only two Ean cooler units)i7 will operate with both fans of the unit running at full speed.Each of the two vertical concrete air shafts is served by an operating fan cooler unit.In this mode of operation the idle trainzserving as standby.6.2 2 4 E>c4~ad%+si-hyphen air dct~per is~cycReh dpchD oi&$goch~o+Q~c d c~~e~is choked~37 Amendment?lo.PZ SliNPP FSAR units wx a total of 4.683x10'tu/hr With 90 F service water'emperature, heat removal capacity and is rated two operating fan s will supply at a to x10 Btu/hr.heat generated in heat generated in a~nment.eac er has 2.83x10 Btu/hr.During this mode of operation, a total o m and will remove the Containment.

37 E~.4 h,Z+~'Pt Q~~mper<s kec4e4 mph~(k eac4 Mope(e MG Mper 55 o e~, Com+.i~~e S+o r<s e<~~~a<~~em 4 a.v t ro+~Q<EM Qf~+lAJ C p)e+~o Skc~hg Cob(erg a$g(~y'pg CV Ca L&ng++Hpp<cLkc~+e'c<ualI~l~e).j

L e~P~~'is c)~S%~~h'4g.~~gM" C Wro~P~e C,.Z.2-'L pi r unit operating at: full speed.xcess teat: generate y t: e Rtm ventilation system Cooling air fr'hese t.wo lers will be directed to the operating floor by auto'c closing, on SES, eumatically operated dampers at th crete air shaft and by opening dampers at-accident dischar z es.During this mode of operaL'ion bot:h Trains A and B be ating.With 95-F service water enL'e ring tempera ture, each f an c, ope with two fans at full speed, has 2.28x10 Btu/hr.hea oval capacity and is at 125,000 cfm.During this mode peration, all four operating fan~oo will supply a t:otal of, 00 cfm and will remove a total of 7.3x10 Btu/hr.he nerated e Cont:ainment. t=a.cg pic%+su.p~'Lq etc pcI'ic)oc40 oPea~~e~~e,s c(o.4.SHNPP FSAR Q.$l~l~g(coolers m'iLL toe 4 era"e'o Cou4o iu~em>ex@'croye, ttg~uc~~+~o h)When the temperatu e is above 80'W~Fan coo er units located at: floor Elevation 236 ft: will operat:e with Caser.fans of the<<nitSrunni.ng aL'ull speed The other t:wo fan cooler units located at Elevation 286: will operate with aaeAfanS , Air is supplied to the steam generator an pressurizer subcompartments, the operating floor, the ground floor and the mezzanine floor.Figures 6.2;2-10 through 6-2.2-.16 describe the p1an and g,'st ductwork.A portion of supply air is tapped to serve the Reactor Support Cooling System and Primary Shield Cooling System described in Section 6.2.2.2.3. +L'is ckircefch+o ~4rCco~~+V~ e~<+e are.The t: ree non-nuclear safety fan-coil unitsxhall located at L'he same elevat:ion. These units are required to operate during normal planL operating conditions only>The fan-coil units are served by the Service Water Syst;em.A detailed description of Service Water System is given in Section 9.2.1.Each unit has cooling coil section and two one hundred percent capacity, direct driven, vane axial fans.4 Vu'it per grwcvucc is~4o~u'~RL L,c t'.2.'2-I.5+5cft'+o P~c Wjth 50 F service water entering temperature, each fan coil unit has 2.082x Bt:u/hr heat: removal capacit:y.at: 80 F entering air temperatur l>uring th.eration all three operating fan coil units will rem a total of 6.246x10 Bt r heat: generated in the Containment. h)(1th 90 F service ter entering temperature ach fan-coil unit has 2.19x10 Btu/hr.heat: remova apacity.Durin is operation all three operating Fan coIL1 units will sup a tot of 273,000 cfm and will remove a t:otal of 6.57x10 Btu/hr.heat: genera in the'Containment. c)W)th 95 F service wat entering tempera e, each fan-coil unit has 1.866x10 Btu/hr.heat: r oval capacity.During thx eration, all three operating fan co 1 ts will supply a total of 273,000 and will remove a tot:al of 5.59x Btu/hr.heat generated in the Containment. Air f the fan-coil units is directed to the RCP and steam generator compar L'ment:s.6.2.2-5 37 Amendment No. SHVLPP FSAR+o+1 Qp fb Y L MAC(g G)With (2)safety related fan cooler units and (3)non-safety related fan coil units op>>rating at a service water temperature of 50 F, their eat removal capacity is 1l.lwlo i and between 48 F and 67 F WB.1'2 7 V-lo The containment heat gain is~~0'tu/hr. This includes heat contributed from equipment, lighting, pi.ping, motors as well as fan motors'.37 Since heat gain is greater than the heat removal rate the temperature in the Containment cannot fall below 80 F.6.2.2.2.2 Containment Spray System (CSS)6.2.2.2.2.1 Functional Description Th>>purpose of the CSS is to spray borated sodium hydroxide solution into the Containment to cool the atmosphere and to remove the fission products that may be released into the containment atmosphere following a LOCA or MSLB.'A summary of the design and performance data for the CSS is presen(ed in Section 6.2.1.The fission product removal effectiveness and the pH control of the containment sump water of the CSS is described in Section 6.5.2.P 6.2.2.2.2.2 Design Description The CSS consists of two independent and redundant loops each containing a spray pump, piping, valves, spray headers, and spray valves.Figure 6.2.2-1 prov, ides the process flow'and instrumentation details of the system'.C I The operation of the CSS is automatically initiated by the containment spray'ctuation signal (CSAS)which occurs when a containment pressure of 12.0 psig (HI-3 signal)is reached.Section 7.3 describes the design bases criteria for the CSAS.Upon receipt of a CSAS, the containment spray pumps start operation and the containment spray isolation valves open.The CSS has two principal modes of operation which are: a)The initial injection mode, during which time the system sprays borated water which is taken from the refueling water storage tank (RWST).Section 6.2.2.3.2.3 describes the criteria used for sizing the RWST-b)The recirculation mode, which is initiated when low-low level is reached in the RWST.Pump suction is transferred from the RWST to the containment sump by opening the recirculation line valves and closing the vaLves at the outlet of the refueling water storage tank.This switch over is accomplished automatically. See Section 7.3 for further details.6.2.2-6 37 Amendment iVo.~