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 Docket No. 50-400 Municipal Power Agency (10 C.F.R. 8 2.206) 5i (Shearon Harris Nuclear Power Plant)

REQUEST FOR INSTITUTION OF PROCEEDINGS PURSUANT TO 10 CFR 2.206 pgo7 koMI~ EDO 001906

OUTLINE CONTENTS I. Introduction II. Statement of Authorization to Represent Persons, Organizations and Interests III. Standard Pursuant to 10 CFR 2.206/2.202.

IV. Emergency Plannning/Preparedness Arguments V. Quality Assurance Program Arguments VI. NEPA Psychological Stress Arguments VII. Summary of Relief Sought VIII. 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, bystandards acts or ommission, has required by 10 failed to meet the applicable 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 in pursuant to 10 the form of a CFR 2.714 (a) and 2.715 (a). A document 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. Initiate Standard Under 10 CFR 2.206 to 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 that by the NRC that there be reasonable assurance 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 exercises are part of the operational preparedness 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 and set sirens off at will, the underlying reliability 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 social will insure the integrated use of arts", inandorder natural to sciences and environmental design 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 National Environmental Policy Act, Act,

'2 consider a novel health and safety issue, in light of the USC S4321, 2133.

et.

The seq., and the Atomic Energy 42 USC s 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 towithoutassure that nuclear power plants will operate 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 the 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

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 significant new circumstances. Furthermore, in reviewing of 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 NUREG-CR-0956.

a 'hard look't NUREG-CR-2239 and 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. Steven P. Katz 04 W. Chapel Hil St. 604 W. Chapel Hill St.

Durham< N.C. Durham, N.C.

(919) 98 3818 (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, s June 7,t 1986, I contacted the following 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.

'in Chernobyl USSR has power ply tt accident on April P6~

1986 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 of Commissioners hereby reaoinds a11 prior approvala .'oard 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.

alarm because she had heard She thought it before.

it was the Shearon Harris 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 plannedsince to increase security at the plant. I was quite concerned 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 isn't if CP&L was testing the quite visible from the front siren. Although the siren 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 familyLieutenant was awakened by Charles the siren. This includes her husband, 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 disaster, one of the sirens, which we rely upon in case of a can't even wake people 200 feet away.

Ms. Thomas knows Anne Wilke who was the dispatcher for the Ch atham 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.

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 theI siren there about 1.5 miles from Ms. Thomas'ouse.

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, 836-6189. I identified myself and said I lived CP&L'pokesman, 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 hear an alarm and turn on the radio and don't hear about if they 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 authorities immediately installed to notify when an alarm goes off.

some 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 any immediate answers.

first and ask questions from Virginia.

I'l be more false alarms and the authorities will not have If I hear a siren evacuate 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 it emergency alarm for Shearon Harris because she had heard 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.

Richardson called 362-8891 and reached David Dean of Mr.

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

<'lO t

Richardson had heard the shif change whistle. ~'r)rg. zend> <<<~

+i<'~ Wld Ck<V  !~CLlddSgm Ct~1)lS.KaqiOPla i~i~ Ji~p(gg L%P ~ ~~- 0 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 present at the plant. They really felt

~

nuclear fuel was 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.

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 q 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 ~~~/ Q)la < gJv ba J g<ggc<0 Ig &p.gY'p)

I Ms. Keyworth told Mr. Mhitehouse he had laughed at her.

anal P~Qc'<1'):<+<<~

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 n ~'>

and got no answer. S>)v its iw:~I 'kvSh:i9-i/ Il~d> Uuu~ ij i<E~ii Dl 0 .

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 alarm had gone was not an emergency and CPSL did not know why the 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 Ai off accidentally." Qf 4 )pl<. Eo<I <i)~+~ @ncaa Illk ~ f <l4&us~ il w~'c v.<&j chc~wcI. orleI( 'Lj ~ (cq(.4>~8 ~l'~ cPsr- eh@ n+~~ an) Rfawd> m 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~ cJ ~d~( Neo~ rupia'yean% >~ +4I~ egg.qgqg, 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 accurate statement of the events and occurrences described it is a true and 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 see was Shearon Harris emergency siren and went outside to it if the 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 it had when she had heard loud enough to it previously. awaken people. She In did not fact, feel her that husband was 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 I I 1 r l I have read the above statement and believe it is a true and events and occurrences described accurate statement of the 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 1 986 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. p p 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 Harris she learned that felt it wassafean because much less emergency siren for Shearon the alarm malfunctioned and because evacuation response. it was too faint to elicit an To the best of my knowledge this statement accurately reflects the substance of my conversation with I have read the above statement and believe accurate statement of the events and occurrences it is described a true and therein. st I I q i ~ / 3 AFFIDAVIT From .'laire and Edward Route 4, Box 638 Thomas Hatley Road Pittsboro, NC 27312 542-3637 This affidavit, taken by Dan Frazier at about 500 p.m. on was awake at June 29, 1986, is to indicate that Claire Thomas 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 something

~ J 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 r 'I "a ~ I have read the above statement and believe occurrences it is a true and described accurate statement of the events and 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 inareplace no required formal, predetermined, evacuation plans 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 of sufficient size to provide for substant educt'on in early severe health effects (injuries or deaths) in the event of the more severe Class 9 accidents." (NUREG-0396, p 17),

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 be larger EPZ upon reevaluation. Also, it should not implied that the term "low-probability accident" indicates that a long time will pass before such an event occurs. It to expect is therefore reasonable "tempered" that consideration of emergency plans be 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 type and officals are responsible size of emergency for deciding planning is acceptable ifandthis 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 methods for 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 it the 10-mile EPZ. Even with these improved techniques, is recognized and'ocumented by the NRC that the dose estimates is still limited reliability of the rish andamounts by the uncertainty of the 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 .meteorologicalgenerate and 2) phenomena The equipment is not designed to sense the 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 Also, to the important data are small-scale sensed at heights motions of the atmosphere. some 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.

Roughly $ 60 million of the neev funth In addition, ncw hunch criteria will be ccording to the rcport, institutions of all s ught for this year are to bc transfcrrcd cstablishcd at thc outset, Trulv said. "When in all regions of thc countIy arc affect-'d.

it's time for the first flight, we are going to e problems of enginccring disciplines fr the Pentagon to DOE,,presumably for 'd to be most serious.

onc r morc underground tests in Nevada, do it as safely as possible. Wc are going to were launch in thc daytime from Kenned> [Space Thc ommittee was formed last May to yo thc nvo to four tests alrcad> schcd-Center in Florida], ivc'rc going to have a assess th state of undergraduate education ulcd fo this fiscal year at a cost of $ 157.8 million. fiscal year 1987, thc under- conservative flight design, [and] ive'rc going in science, mathematics, and cnginecring ground t ting account will jump to $ 226 to have repeat payload, a one that wc have and make r ommendations nn thc role thc miiiiony or nough for three to fiv explo- cxpcricnce with." No civilians willfly during National Scic cc Foundation should take in sions. (The dgct for underground testing thc first year, and all flights will occur in improring it. chairgnan was Homer A of the weapon has cxcceded that for labora- warm weather, he indicated. Ncal, provost the-State University of tory research fo scvcral years.) In addition Truly cxplaincd that thc rules arc neces- Ncw York at Sto y, Brook. Thc committee to the x-ray laser, variety of nuclear-driven sary to restore the agency's credibility in the reported to the,.'ational Science Board, weapons such as article beams, micro- wake of thc Challenger disaster (Scig;>su, 28 which is thc jIdlic) making body for thc waves, hypcrvclocit) Ilets, and optical la- March, p.1495). Thc agency's present plan foundation.>'n sers arc also under i vcstigation and may is to conduct roughly nine flights a year, its rem~mmendat ns, thc committee

~

eventually bc tested. beginning a year from now. First priority said that N SF lacks the resources to solve the nuclear power urces, ifyou want

'These will bc given to launching military satellites, problems itself, but should take a Icadcrship to consider them that wa (they arc cxplo- as well as a tracking and communicauons role in stimulating the state's and thc private sions but they act as powe sources)," may satellite dcstroycd by thc accident. 'Wc can- sector to increase their investment in under-ulumatcly bc unnecessary for a ballistic mis- not print enough mona~ to make thc flights graduate science, cnginecrin and math sile defense, Wagner testifie. ut "thc first risk-free, Truly added. "But wc certainly arc.n". 'education. Thc panel docs recommend that stages of the SDI program, wh h... may going to correct any mistakes that wc may NSF expenditures in thc field %increased last decades, I bclicvc and thc D artment have made in the past, and wc arc going to gct by $ 100 million a year in "Icvcraged" pro-soon wc can." r':. gram support. Some $ 5.5 million for college believes will have this nuclear component, going again just as as this new kind of nuclear-driven tlirccted R. JEFFREY 'SMITH instrumentation is thc only program in un-energy weapon as onc of its very im rtant dergraduate education in the NSF budget options." r R.JEEPREv SMITH this year. NSF director Erich Bloch'.is charged irith converting the committcc rec-Panel Sees Decline in ommendanons into proposals to be incor-'<

Undergraduate'Education porated in net year's NSF budget. ~

Briejg: 7oHN w~H A National Science Board committcc rc-port says that the nation's undcrgraduatc New Shuttle Director programs in science, mathematics, and engi- t Promises Emphasis on neering "havydecline'd in quality and scope Nuclear Meltdown: A y

Safety to such anI>extent that they are no longer Calculated (and .ruj meeting ~IIational needs." This poses a A ncw emphasis on safety will be'thc "gravepkong-term threat to thc nation's sci- Recalculated) Risk hallmark of thc space shuttle's operations entific and tcchnical capacity, its industrial when flights resume, according to Rear Ad- and'cono'IIIic competitiveness, and the For yearsy the nuclear indusuy has been miral Richard Truly, thc new associate ad- length of its national defense," thc panel trying to persuade the govcmment to sec a ministrator for space flight at thc National /watns. silver lining in thc cloud that gathered over Aeronautics and Space On thc basisepf evidence gathered in its Three Mile Island. Broadly, thc argument is that the 1979 nuclear accident was much Administration'NASA).

Speaking on 25 March beforcyan inquiry, the cominittce pinpointed three ar-enthusiastic crowd at thc Johnson /pace eas that require highest prioritv attention. less dangerous than oflicial risk cstimatcs Center in Houston, Texas, Truly outlined a r Laboratory Instruction was described as would have Ied pcoplc to expect. Therefore, scrics of activities that hc said are~required "often uninspired, tedious, and dull." In- thc risk studies should bc rewritten. Eventu-to establish a rcalisuc and achiyrabie hunch strumentation and facilitics werc found to ally, ifanalysis confirms what thc accident at rate that will bc safely sustairNfble." be obsolete and inadequate thc need for Three Mile Island suggested, safety regula-r Specifically, the entire budget and pro- ncw Instruments was put at $ 2 billion to $ 4 tions may be adjusted to reflect a calmer gram management "philosophy, structure, billion. view of what would happen in a meltdown.

reporting channels "'rfd decision-making r Faculty members in too many nses An exercise of this kind has begun at the process will be tho ughly rcvicwcd," hc werc seen as unablc to maintain their teach- Nuclear Regulatory Commission (NRC),

said. Allshuttfe co poncnts considered vital ing skills, currency in their disciplines, and called the "source terms" review (Scicyyu, 5 to thc safety of c orbiter and thc crew will command of ncw technology. Serious short- April 1985, p. 31). The phrase refers to bc rcasscssed, will all waivcrs of engineer- ages of qualificd faculty werc noted in some mathematical terms used to calculate leakage ing redund cy. Inspection and test rcquirc- disciplines. from radioactive sources. This project was w'c r

. ments reviewed, and the booster joints, Idely recognized to have been the bc cdcsigncd under thc direction of thc Courses and curricula werc dcscribcd as "frequently nus f thc shuttle accident in January, will nativc, poorly organized for students with out-of. date in content, uniinagi-interests, and (they) fail to rcflcct 'iflcrcnt 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 arshall Space Flight Ccntcr in Huntsville, rcccnt advances in thc understanding of thc accident than many had predicted. Rath-Alabama. tnching and lcarning.- er than venting to thc atmosphere in a pure II hPRII l986 Sc.i ee~t ~e'o'e y3ugNEws ic'. cohthIENT I33

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form, virtually aH of it combined with other For example, onc member of the APS Insurance Drought-chemicals and stayed in thc plant. committee, Fred Finlayson of thc Acrospacc On 26 March, NRC heard a stafF rcport Corporation, wrote to thc NRC in January Fosters Self-Help Plari for on the work done so far in the source term to explain why he considered thc task unfin- Biotechnology Firms renew. Thc NRC staffcrs said they definite- ished. The codes have not been thoroughly 'he ly could scc a glimmer in thc darkness, but peer-reviewed, Finlayson wrote, and their insurance crisis that is cu ndy af-they could not bc sure whether it was thc technical assumptions have not bccn adc- fecting a host of industries has not f>assed up glint of a silver lining or just another light- quatcly disclosed. Hc concluded that there biotechnology. Faced with exorbitant pre-ning bolt. Dcspitc their uncertainty, they werc "too many uncertainucs to provide a miums and in many instances thc mability to promised to have some ncw risk cstimatcs rcasonablc basis for revised risk analysis at obtain insurance, small biotechnology firms ready for publication this fall. this time." Nothing has changed his opinion arc turning to insuring themselves. The As-Last year, thc NRC released the first draft since January. soeiatihn of Bioteehnoloiiy /Companies of a source term document that is meant to -

Another, broader problem with the codes (ABC) plans to sct up an ofFshorc insurance serve as the new scienufic basis for work in is that they distort natural phenomena by venture'to provide liability coverage to 20 the area. The report, called NUREG.0956, simplifying them. (The codes must bc sim- member'zompanics.

docs not deal at aH with risks. (These wiH bc plified to suit thc computer.) Thus, knotty Warren, Hyer, managing director ofABC,

/

calculated in a separate.document duc in problems arc sometimes omitted. Howcvcr, says that this plan hopefully: wiH solve the V

October, designated NUREG-1150.) In- these knotty ones could be important in an member companies'mmediate insurance stead, thc scientific document provides de- accident. For example, onc such hard-to- crisis. Furtflermorc, it also Iiiay pave thc way tailed forecasts of hoiv radioactive chemicals model event is the scenario in which a for thc insurance indusuy tn provide at least might behave in 16 types of accidents and in molten core interacts with a limestone con- limited supplemental underwriting to com-six t)~ of reactors. When it is complctc in crete floor to produce volumes of gas, heat, panies for upgrading general liability cover-July, it willscrvc as the starting point for thc and a radioactive aerosoL In thc right cir- age, protecting corporate,cxecutivcs and di-risk analysis. cumstances, these fumes could burst rectors as individuals, bringing ncw prod-While the future version of this NUREG through thc containment and pose a serious ucts to market,'or sealinp'up experiments for report may bc sound, the present edition has threat to public health. field and clinical been grcctcd with skepticism. Thc nuclear is hard to "gct, says Hycr, be-trials.,'nsurance Indeed, thc codes are inadequate to cope industry, which has sponsored its own re- with fuel-concrete interactions, onc NRC cause the insuraiicc indusuy "does not know search, calls it outdated and alarmist. Thc official says, because the tcchnical issues arc much about bioteehnblogy. Thc risk right antinueiear groups sec it as underplaying unresolved. Research on this topic is now in now cannot be identified." But insurers may hazards. And a number of scientists describe progress in West Germany and at thc Sandia bc more wiHingaio jake on biotechnology it as simply unripe. In this regard, the file of National Laboratory in Ncw Mexico. Simi- concerns, hc says, after at I thc association's new public comments reveals an inhcrcnt prob- lar unccrtaintics plague thc issues of contain- insurance operation starts functioning. Dis-lem that may keep the project unripe for a ment building integrity, high-pressure ejec- cussions with two Ncw York-based interna-long ume. This is a disagreement over thc tion of fuel from the reactor vessel, hydro-uonal brokers Maarsh 8c McLennan, Inc.

credibility of some computer modeling gen production, iodine and lanthanum and Johnson Bc Higgins indicate that cov-codes that are the basis for aH thc predicuons chemisuy, and rcvaporization of deposited erage on potentiaI liability claims exceeding that will come out of NUREG-0956. fission products. AH arc being researched. Sl million might bc availablc from private There arc two levels of disagreemcnt. Ciung the code's deficiencie in dealing with insurance companies in the future, says First, some researchers chaHcngc the codes chcmisuy, R. Potter, a Briush official at the Hycr. /

on a mechanical basis. Thc codes arc so Atomic Energy Establishment at Winfrith, 'ABC's tentative plan calls for each mem-complex, tedious to rcvicw, and obscure, wrote of the trcaunent of iodine chemisuy: ber companyl'to bc in-..urcd for liability eriucs say, that they have been reviewed by "At best this is an oversimplification, and at claims up to/Sl miHior:. Each company almost no one except those paid to do so, worst, wrong." Unless this and other aspects would pay an annual prem um of $ 100,000.

that is, by N RC contractors. There may bc a were improved, hc concluded that it would Thc companies wiH rcvicii'ach other's re-hidden bug in thcsc models that no onc has bc difFIcult to have the necessary confidence search portfolios and will esebHsh "a strong detected. Furthermore, it is impossible to in thc results." risk-prevenu'on program" that sets out gen-

"validate" thc codes fully, for no one is Thc NRC stafF, induding the acting exec- eral guideli.ies for thc conduct of rcscarch.

going to stage nuclear accidents to scc how uuvc director Victor SteHo, assured thc Thc affiHati'. of thc trade assoc iation is likely well the numbers represent reality. For this commission that corrccuons and cmenda- to be loca".ed in the Bahamas or Bermuda, reason, it is important that they be thor- tions of document NUREG.0956'ill bc Hyer indi ated, to avoid U.S.'-tax laws that oughly vctted by independent scientists. finished by July. Unresolved technical is- would tre'at a surplus in thc in':urancc enti-Several commissioners stressed this point sues, such as thc interactions of thc fuel with ty's trust'funds as a taxable profit.

during thc briefing. concrete, will bc handled by setung wide Thc ir,'surancc crisis cxtcnds to biotcchno-Last year, a committee of the thc Ameri- uncertainty margins around relevant terms log)~s larger players, including pharmaceuti-can Physical Society (APS) reviewed some in the analysis. Work on the risk estimates cal and;chemical giants. "Everybody is hav-of this work, issued a rcport, and then themselves has already begun and will bc ing insurance problems," says Sus'an Racca, disbanded long before the game was over, completed within 6 months. Finally, in the an analyst at the Industrial Biota;hnofogy it turns out. These APS members werc bureaucratic tradition, a policy paper issued Assodation. Member companies of rhc IBA consulted, according to the NRC stafF, by StcHo also promised that thc stafF would arc scheduled to meet next week to discuss a

.p s bout the final version of NUREG-0956. begin to propose regulatory changes right self-Insurance plan. Thc associauon:helvcd But some of the APS group felt the consul- away, or, in any case, "as soon as the avail- thc jdea several months ago but is ta.dng it tation was perfunctory and fell far short of ab! c information warrants such changes." a up again, says Raeca, "because things have Full pccr review. ELIOT MARSHALL go cn so bad." r MAMtCRAvmoari SCIENCEs VOL. 232

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Page 1 of 2 AMENDMENTNO. 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 injection valve as normally open based on results of startup 'afety 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 2 9.1.3 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.

9.1.0 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 This section is revised to provide additional details of .'.he as-built design of the diesel generator cooling water system and suppcrt preoperational testing.

12.3.0 Editorial Change 13.1.1, These sections are revised to reflect recent management organization 13.1.2 R changes and provide consistency with Technical Specifications.

13.1.3 13.2.2 TPe description of the Licensed Operator Requalif ication Training is revised to reflect 10CFR55 requirements.

10.2.12

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

15.6.5 Typographical Error 15.7.0 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 37 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 ~

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.

37 2.4.8-1 Amendment No. P6

TABLE 3.2.1-1 (Continued)

CLASSIFICATION OF STRUCTURES STSTEMS AND COMPONENTS Desi n and Construction and 0 eratlons Remarks Quality Quality r Safety Code Seismic Quality Class Assurance Structures Class (1) Code Class C~ate or (2> Assurance (3> (23> (24>

Diesel Fuel Oil 'Storage NA See >lots (3C>

Tanks and Tank Building Containment Air Locks, Equipment 2 ASME I I I MC 3 3.33 S Hatch and Valve Chamber Note (29)

Containment Internal Structures Containment Crane Supports Cooling Tover NNS E Electr Ical Manholes tor See Note (30)

Emergency Pover and Control Cables S stems and Components Reactor Coolant S stem Reactor Vessel" I ASME ill Steam Generator (tube side)

(shel I side)

I 2

ASME ASME

'I I I III Q

Q S<<<o<<(~> I

~

Pressurizer I ASME III

TABLE 3.2.1-1 (Continued)

CLASS IF ICATION OF STRUCTURES SYSTEHS AND COHPONENTS Desi n and Construction and 0 eratlons Rsssrks

()ue((ty ()ue((ty Safety Code Se I sml c quality Class Assurance S stems and Com onents Class (I) Code Close ~Cote or (2) Sssureuue (3) (23) (24)

Reactor Coolant Hot and Cold ASHE ill I I Leg Piping, Flttlngs and Fabrication Surge Pipe, Spray Pipe, Fittings, and Fabrication ASHE ill I P, See Note (5)

Crossover Leg Piping, Fittings I ASHE I I I I and Fabrication 4J RTD Bypass Hanlfold ASHE I I I 1 I

Pressurizer Safety Valves ASHE III 1 :A Q Pressurizer Power Operated ASHE III 1 h P Relief Valves and Block Valves Valves of Safety Class I to ASHE III . 1 Safety Class 2 Interface Pressurizer Relief Tank NNS , ASHE VIII Reactor Coolant Thermowell I ASHE III 1 g'ux 5

1 I I ary Reactor Coolant Piping (Drains, etc,)

2 ASHE III 2 Pressurizer Rel lef Valve Discharge 1 ASHE III I Lines (between Pressurizer Nozzle and Relief Valve Only) 4v

TABLE 3.2 ~ I-I (Continued)

CLASSIFICATION OF STRUCTURES SVSTEMS AND COHPONENTS I

Desi n and Construction and 0 eratlons Remarks Pual(ty Quality Safety - Code Seismic puallty Class Assurance S stems and Com onents Class (I) Code Class ~Cate or t2) Assurance (3) (23) (24)

Steam Generator Forging Type A I ASME I I I B A 0 See Note (9)

Chemical K Volume Control S stem Regenerative HX 2 ASHE I I I Letdoxn HX (tube side) 2 'ASME I II (shell side) 3 ASHE III Hlxed Bed Demlneral I zer 3 ASHE ill 3 See Note (7)

Cation Bed Demlneral Izer 3. 'SHE III - 3 See Note (7) e Reactor Coolant Filter 2 ASHE I I I 2 Volume Control Tank 2 ASME I I I Charging (High Head Safety 2 ASHE III Infection) Pumps Charging Pump Hotors IE Seal Mater InJectlon Filter 2 ASHE III Seal Mater Return Filter 2 ASHE III Boric Acid Blender 3 ASHE III B Letdoxn Orlf lees 2 ASPIC III

TABLE 3,2 '-I (Continued)

CLASS IF ICATION OF STRUCTURES SYSTEHS AND COMPONENTS I

Desi n and Construction and 0 eratlons Remarks Quality ()uallty Safety Code Sel sml c ()uallty Class Assurance S stems and Com onents Class (I) Code Cl ass ~Cate or (2) Assurance (31 (23) (24)

Excess Letdown HX (tube side) 2 ASHE III (shell side) 2 ASHE III Seal Mater HX (tube side) 2 ASHE III 2 (shell side) 3 ASHE III 3 Chemical Hlxlng"Tank NNS ASME VI II Chemical Hlxlng Tank Orlflce NNS Boron Heter NNS ANSI B3I ~ I Boric Acid Tanks 3 ASME III Boric Acid FIl ter 3 ASHE II I Boric Acid Transfer Pump 3 ASME III B A Boric Acid Transfer Pump Hotors IE Boric Acid Batchlng Tank NNS ASME Vl I I Reactor Coolant Pump (RCP) NNS ASME Vl I I Standpipe RCP Standpipe Orlf Ice 37 RCP Seal Bypass Orlflce I ASHE I I I 8

TABLE 3.2 ~ l-l (Continued)

CLASSIFICATION OF STRUCTURES SYSTEMS AND COMPONENTS I

Desi n and Construction and 0 eratlons Remarks Puality Quality Safety Code Seismic Quality Class Assurance S stems and Com onents Class (l) Code Glass ~Cate or (2) Assaraaaa (3) (23) (24)

System Piping and Valves a) Part of RCPB III b) Required for reactor coolant letdown and makeup 2

I ASME ASME III I

2.

B 8

Q Q s~~~ 31 c) Required lor providing boric 3 ASME III A acid for the letdown and makeup loop d) Normally or automatically NNS ANSI B31 ~ I 37 lA Isolated from parts of system covered by a, b or c I Instrumentatlon IE A Q See Note (15)

Operators for Safety-Related IE A Q See Note (31)

Active Valves Boron Thermal Re eneratlon Subs stem Moderating HX (tube side) 3 ASME III 3' See Note (7) P Sa~~(2P)

(shell side) ASME III 3 See Note (7) Q Sem8&~9 Letdown Chiller HX (tube side) 3 ASME III See Note (7) A

' ~4ccto=~

(shell side) NNS ASME V I I I E II" Letdown Reheat HX (tube side) 2 ASME I I I I Q ~eke=~

(shell side) 3 ASME III See Note (7)

Thermal Regeneration 3 ASME III See Note (7)

Demlnerallzer 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) Coda Code Class

'e I sml c

"~Cats or (2)

Quality Assurance (3)

(Puallty Class (23)

'uality Assurance (24)

Chiller Surge Tank NNS ASME Vl I I Chiller Unit NNS E a) Evaporator NNS ASHE 'Vl I I E b) Condenser NNS ASHE Vill E c) Compressor NNS E System Piping and Valves Ll a) Not normally or 3 ASHE III automatically Isolated from safety class I

components c b) Other 37 NNS ANSI B3l ~ I Boron Rec cle S stem Recycle Hold Up Tank 3 ASHE III 31 Recycle Honltor Tank NNS AWA D-l00 Recycle Honltor Tank Pump Casing NNS ASHE Vill' b Recycle Evap, Feed Pump 3 ASHE III 3 See Note (7) 8 Recycle Evap Feed 3 ASME I I I 3 See Note (7)

Demlnerallzer ft 0

5 TABLE 3.2 '-1 (Continued)

CLASS IF ICATION DF STRUCTURES SYSTEMS AND COHPONENTS Desi n and Construction and Operations Reearka I 24 Quality Quality Safety Code Seismic Quality Class Assurance lm S stems and Com onents Class (I) Code Class Rate(ear (2) Assurance (3) (23) (24)

Recycle Evap. Feed Filter 3 ASME III See Note (7) B A Q Recycle Evap. Condensate NNS ASHE VIII Deminerallzer 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 E VNI Recycle Evaporator Package a) Feed Preheater I) Feed Side 3 ASME III 3 . See Note (7)

2) Steam Side NNS ASME VIII b) Gas Stripper ASHE III 3 See Note (7) c) Submerged Tube Evap, I) Feed Side 3 ASHE III 3 See Note (7)

2) Steam Side NNS ASHE VIII d) Evaporator Condenser I) Distillate Side 3 ASHE III 3 See'Note (7)

2) Cooling Water Side 3 ASME ill 3 I Q 4ee=Hc4e~+

TABLE 3,2,1-1 (Continued)

CLASSIFICATION OF STRUCTURES SYSTEMS AND COMPONENTS Desi n and Construction and 0 eratlons Raaarks pballty Quality Safety Code Seismic Quality Class Assurance S stems and Components Class (I) Code Class ~Cata or >2> Assurance (3> >2>> tra>

e) Distillate Cooler I) Distillate Water Side 3 ASHE III 3 See Note (7)

2) Cooling Water Side 3 ASHE III 3 I e Note (

f) Absorption Tor)er 3 ASHE III 3 See Hote (7) Note (

g) Vent Condenser I) Gas Side 3 ASHE III See Note (7) 3'A ee Note (4

2) Cooling 'Water Side 3 ASHE III I ee Note (4 h) Distillate Pump 3 ASHE III See NCTe (7) 4 /I ee ot (4 I) Concentrate Pump 3 ASME III See Note (7) 4'l ee T (4 J) Piping I) Feed 3 ASHE III 3 See Note (7) 8 ~A ~

echo e (4

2) Distillate 3 ASHE III 3 See Note (7) 8 p A eeh a I-'a

3) Concentrate 3 ASHE III 3 See Note (7) 8 T

4) Cool lng 3 ASHE III 3 I 8 ee cT (4

5) Steam NNS ANSI 831 '

k) Valves I) Feed 3 ASHE III 3 See Note (7) B.

PA QQ e Nota 4

2) Distillate ASHE III See Note (7) 8 P'A ~ ()I) e Note (

3) Concentrate 3 ASHE III 3 See Note (7) 8 P'A . P' e Nota (

4) Cooling 3 ASHE Ill 3 I 8

5) Steam NNS ANSI 831 '

TABLE 3,2.l-i (Continued)

CLASS IF ICATION OF STRUCTURES SYSTEHS PAD COMPONENTS Desi n and Construction and Operations Reearks Quality Quality Safety Code Seismic Quality Class Assurance S stems and Components Class (I) Code C(ess ~Cate or (2) Assurance (3) (23) (24)

System Piping and Valves a) Not normally or= 3 ASHE III 3 See Note (7) Q automatically Isolated from safety class components b) Other NNS '. ANSI B31 ~ I Safet In ection S stem Accumulators 2 ASHE III Boron Infection Tank (BIT) 2 ASHE III B Hydro Test Pump Q

System Piping and Valves a) Part of RCPB I ASME III Q

TABLE 3.2.1-1 (Continued )

CLASS IF ICATION OF STRUCTURES SYSTEHS AND COHPONENTS Desi n and Construction and Operations Remarks Quality Quality Safety Code Seismic . Quality Class Assurance S stems and Com onents Class (I) Code Class ~Cate or (2> Assurance (3> (23> (24>

c) Piping and valves required for 2 ASHE III 2 I 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 Q See Note (31) 43 Active Valves Reactor Coolant Drain Tunk I

t4 Ht ~ Exchanger (shell side) 2 ASHE III 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

-Quality,'uality Safety Code Se I sml c Quality - 'lass . Assurance '-

S stems and Com onents Class (I) Code Class ~Cere or (2> Assurance (3) .(23) (24)

Fuel Pool Coolln and Cleanu S stem Fuel Pool Heat Exchanger (tube side) 3 ASME Ill 3 8 A ~

(shell side) 3 'ASHE I II 3 8 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 VI I I

.8

. ASME E Fuel Pool Demlnerallzer NNS ASME Vl I I Fuel Pool Refueling Water, NNS ASME Vl I I., " E ~

Purl f ication FIi ter Fuel Pool Stralners 3 ASHE I I I Fuel Pool Sklmmer Filters L -

ASME Vll I E Fuel Pool Sklmmer Pumps'NS: ~ 'NS E P-r Fuel Pool and Refuel lng Water . E Pump NNS'urification 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 . 3 ASHE III 8 makeup to the fuel pools b) Hakeup from RWST 3 ASHE III 8

TABLE 3 2,1-1 (Continued)

.CLASS IF ICATION OF STRUCTURES SYSTEHS AND COHPONENTS I

Desi n and Construction and 0 eratlons Remarks Quality Quality Safety Code Se I sml c Quality Class Assurance S stems and Com onents Class (I) Class ~Cate or (2) Assurance (3) (23) (24) c) Required for fuel pool cleanup NNS ANSI 831 '

and normally Isolated from a)

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 See Note (10)

Q .

Fuel Transfer System a) Fuel Transfer Tube and Flange 2 ASHE I I I B h Q . See Note (ll) b) Portions of Conveyor System and 3 .B A Q See Note (12)

Controls ln Fuel Handling Building c) Remainder of System NN 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 Quality Quality Safety Code Sel sml c Quality Class Assurance S stems and Com onents Class (I) Code Class ~Cote or ttt Assurance (3) (23) (24)

New Fuel Assembly Handling Fixture NNS I 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 p'- sot 6'oto CeQ (37 Reactor Coolant Drain Tank Heat Exchanger (shell side) 2 ASME I I I B.

(tube side) NNS ASME V I I'I 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 Desi n and Construction and 0 erations Rasarks Quality Quality Safety Code Seismic Quality Class Assurance S stems and Com onents Class (1) Code Glass ~Cate or (2) Assaraaoa (3) (23) (24)

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

Containment Ventilation S stem Airborne Rad I oact I v I ty Removal NNS E P- See Note (IS)

System

TABLE 3,2 l-l (Continued)

CLASS IF ICATION OF STRUCTURES SVSTEMS AND COMPONENTS 4

Desi n and Construction an3 0 eratlons Remarks Quality Qual Ity Safety Code Selsmlc Quality Class Assurance S stems and Com onents Class (I) Code C(ass ~Cate or (2) Assurance (3) (23) (24)

~es+

CROM Cool lng Systems /'B Containment Combustible Gas Control S stem Electric Hydrogen Recomblner '2 B A Q Instrumantatlon (In part) IE B A. Q See Note (l5)

Hydrogen Monitoring System (0-IOS range capability) a) Piping and Valves 2 ASME I I I A Q b) Hydrogen Analyzer Cabinet IE A Q See Note (l5) c) Remote Control Panel ' A IE See Note (I5) d) Remote Sample Dilution Panel NNS E Containment Vacuum Relief 2 ASME III (except blind flanges and valves for leak testing)

ZivCET 8eDWe27iI 3 jl~Ill Instrumentation IE Primer Shield Cool ln S stem Instrumentatlon Reactor Su orts Coolln S stem 3 Instrumentatlon

TABLE 3.2,1-1. (Cont inued)

CLASS IF ICATION OF STRUCTURES SYSTEMS ANO COMPONENTS s

Desi n and Construction and Operations Raaarks Quality Quality Safety Code . Se I smic Quality Class Assurance S stems and Com onents Class (1) Code Class '~ete or (2) Assurance (3) (23) (24)

Reactor Auxlliar Bulldin (RAB) ..

Ventilation S ste neo foR~ RAB Normal Ventilation System AHg a) Isolation dampers A hJ b) All other components 'NS E I RAB Steam Tunnel Ventilation lus2 RAB Emergency Exhaust System - 3 A RAB ESF Equipment Cooling Syste s 3 s

ESF RAB Batter y R xhaust Fbns 3 RAB Computer d Communications ~ A 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 RAB Electric Equipment Protection 3 0 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 Desi n and Construction and Operations Reearas Puallty Puality Safety Code Seismic puallty Class Assurance S stems and Components Claaa (1> COde Cleea ~Cate Or (2> eseaeraeee (3) (23> (24)

Maste Processin Buildin NNS Sffivy% C+)

Ventilation S stems MGC ~ct T~~R~~u lfACk 4>ieFW Coo~a l~l Control stems porn HVAC S pd(d Normal Supply Subsystem e'X A(AS j a) Supply Fans d Casings 3 b) Cooling Coils 3 ASHE III c) Electric Heating Coils IE d) Ducts and Dempers 3 e) Valves for Outside 3 ASHE ill 3 Air Intakes f) Chlorine d Radiation IE Detectors g) Smoke Detectors NNS

~up Control b)

Room Smoke Pur e and Exhaust a) Boundary Isolation Valves Other 3

NNS ASHE ill p ~~

OFF Ikon (/Sg i37 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 fKTHgNAIIFAf& r<< SSy uorS(S8$

AFAT&e QoICs sar A61i(J8$

i) &AnW~ JAILS,OuCT> ~d'Op~pF+S iVNS A

TABLE 3.2.l-l (Continued)

CLASS IF ICATION OF STRUCTURES SYSTEMS AND COHPONENTS Desi n and Construction and Operations Remarks Puality Puality Safety Code Seismic Quality Class Assurance S stems and Com onents Class (I) Code Cl ass Sateraor (2) Assurance (3) . (23) (24 )

b) Exhaust Fans E c) Ductwork and Dampers I) Isolation,Dampers .3 A 0

3) Other NNS E smzN Ti(IBJ Ouch Unl gen y x aust System for the Operating Floor Normal Ventilation System for Areas Below Operating Floor a) Air Handling Unit NNS E b) Exhaust Fans NNS E c) Ductwork and Dampers I) Isolation Dampers 3 8 A ..

~ .

P s(5 R( Ii(IBJ

2) Other NNS 2 Spent Fuel Pump'oom Ventl I ation A P System Instrumentatlon .- IE S . () See Rote ((5) I~

Fuel Ol I Transfer Pum House Ventilation S stem -.3 A P Diesel Generator Bulldin I a A ~

P 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

TABLE 3,2 '-I (Continued )

CLASSIFICATION OF STRUCTURES SYSTEMS AND COMPONENTS Desi n and Construction and Operations Remarks Quality Quality Safety Code Se I sml c Quality Class Assurance S stems and Components Class (I) Code Class ~Cate or (2) Assurance (3) (23) (24)

Chilled Mater Piping and Valves a) Required to provide chilled 3 ASME III uater to safety related air handling units b) Required only for RAB NNS Ventilation Systems stXPo7E (I8$ ) ~ 3/

and automatlcaliy isolated from a)

IO c) Operators for Safety-Related IE g

K Active Valves I

4J Ln Instrumentation IE A See Note (l5) Vl Non-Essential Services Chilled NNS spy /Vole'(18/

~aster S stem Containment Atmos here Pur e and M~akeop S stem Ductwork Inside Containment P 3. jB pA /e )M 37

+~F to the isolation valves Containment isolation valves 2 ASME III A '

and piping From I sol at Ion va I ves outs I de A

Containment to floor pene-tration at RAB Elevation 286 ft (puagEN4KEuP)

~i d RAB H PQ a; Cr ~r S A ~Kg 3i 0 s umen ion (iso a n E Q See Note (l5) valves only)

Other ~ NNS sa P,f~ggs

TABLE 3,2.1-1 (Continued)

CLASSIF ICATION OF STRUCTURES SYSTEHS AND COHPONENTS Desi n and Construction and Operations Qudllty Puallty fiemarks lm S stems and Components Safety Class (I) Code Code Cleat Se I sml c

~Cate ar (2)

Ouallty Assurance (3)

Class (23)

Assurance (24 )

lm Operators for Safety-Related - IE Q See Note (31)

Active Valves apl A AT Containment N dro en Pur e and Containment Isolation valves 2 ASHE III and piping F rom I so I at i on va I ve outs I de 37 A. Q Containment to floor pene-tration at RAB Elevation ft.

Instrumentatlon (isolation IE q See Note (15) valves only)

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 s g C

ToR n/E 80Jldi~ NdS 0

ate SySggRIS

TABLE 3.2.1-1 (Continued)

CLASS IF ICATION OF STRUCTURES SVSTEMS AND COMPONENTS Desi n and Construction and 0 eratlons Remarks Quality Quality Safety Code Seismic Quality Class Assurance S stems and Com onents Class (I) Code Class ~Cate or (2> Assurance (3) (23) (24)

See Note See Note b) From the MSIV up to and (16) (16) including the last seismic restraint In the Turbine Building c) Downstream of last seismic NNS . ANSI 831 ~ I restraint in Turbine Building d) Operators for Safety-Related IE A Q See Note (31)

Active Valves e) Turbine Gland Sealing System B31 ~ I C Instrumentatlon IE A Q See Note (15)

Steam Generator Slowdown S stem System Piping and Valves a) From steam generator to 2 ASME II I

.and including containment isolation valves b) From containment isolation 3 ASME III valves to RAB wall Condensate and Feedwater S stem Condensate and Feedwater Pumps NNS h4 0

E I ectromagnet I c F I I ter NNS ASME Vl I I (27) C R See Note (27)

Condenser Evacuation System NNS 831 ~I

+hf IntFIVcjffoM'I/AJVE'AGg ~/V5 Sff ATE SFFIVO1F iI 4l47 gfigiekiC gE5~AIRT) iW (gg (it J fu~62~F ger) LCh~g d) up7f,rW7; I4ZS. 7 AWE AMBI 631 I i'uabuK 8w~cfieg TABL 3.2. I-I (Cont lnu CLASSIFICATION OF STRUCTURES .SYSTEMS AND COHPONENTS Desi n and Construction and 0 eratlons Reearka Quality Quality Safety Code Se I sml c Quality Class Assurance S stems and Components Class (I) Code Class C~ate or <2) Assurance (3) (23) (24)

System Piping and 2'SHE -III 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 3 b) HFW control valves and ~S . ASHE I I I ,See Note (4) bypass control valves; -44ew-c) t)~ Operators for Safety-Related Active Valves, IE 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 IE Condensate Storage Tank 3 ASHE III AFW Pump Turbine Driver 3 ASHE III Q See Note (2S)

System Piping and Valves a) From steam generator up to 2 ASME I I I and Including the containment isolation valves

TABLE 3.2 ~ I-l (Cont.lnued)

CLASSIFICATION OF STRUCTURES SYSTEHS AND COMPONENTS Desi n and Construction and Operations Remarks Pual Ity Puality Safety 'ode Se I sml c Puality Class Assurance S stems and Com onents Class (I) Code Clless ~Ceto or (2) Assurance (3) (23) (24) b) Other 3 ASHE III A c) Operators for Safety-Related IE A P See Note (31)

Active Valves 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 NNS E Tank Hake-up Water Storage 3 ASHE III 'eactor 37 Tank 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 Desi 'n and Construction and Operations Rssarks Quality Quality Safety Code Seismic Quality Class Assurance S stems and Com onents Class (ll Cods Class C~sts or (2> Assurance (3) (23) (24)

Piping and valves up to and ASHE III 2 including second isolation.

valve

$ 31.'1 All other piping NNS X Instrumentatlon IE Q See Note (15)

Inadequate Core Cooling System . IE Q See Note (l5)

(in part)

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 although designed and fabricated to the applicable portions of it is not classiEied as ANS Safety Class 1, 2, or 3.

ASME III, (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)

- QA requirements will meet 10CFR50 Appendix B criteria.

Q R - QA requirements will meet ETSB ll-l QA requirements as a minimum.

"Q" requirements may be imposed.

F - Optionally QA requirements will meet Fire Protection QA requirements as a minimum. Optionally "Q" requirements may be requirements of 10CFR50 Appendix B are not mandatory.

imposed'A (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 System can be Eound on Table 11.4.2-4. 'rocessing (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 Valve Design System

. Env, Safety Rating Design Size

~ta llueber ~tutee Looatloa gual. ~T Oe Operator uaaufaaturer Class (ANSI S) Conditions (Inches-ID) Function I CS-V711SN CS RCB (4) Check hP Rockwell 152 I 2485 pslg RCPB Boundary 8 650 F ICS-V70SN RCB (4) Check hp Rockwell I 1521 2485 pslg RCPB Boundary 8 650 F 2CS-V129SH CS RAB (3) Check Rockwell 2 1500 220 pslg Safe Shutdown 8 200 F tA I

Vl

'tet 0

3CS-V222SN'S RAB (3) Check hp Rockwell 3 1500 150 pslg Safe Shutdown 8 250 F 3CS-V223SN CS RAB (3) Check Rockwell 3 1500 150 pslg Safe Shutdown 8 250 F ISI-V39SA V45SB SI RCB (4) Check Rockwell I 1521 2485 pslg RCPB Boundary V51SA (I 650'F Q

I S I -V63SA V69SB SI RCB (4 ) Check Rockwell I 152 I 2485 ps lg RCPB Boundary V75SA ii 650 F 0

g~-u~2, SZ RA& Cl~L < Copes- Jwlcc a Q (5'oo Zoo ps ig CO~4Cal&MCMW Mg p 2ooF WS OL O $ I ~

TABLE 3.9.3-14 (continued)

NON-NSSS SUPPLIED CLASS 1 2 AND 3 ACTIVE VALVES Valve Design System Env, Safety Rating Design Size

~ta N eaer ~eatee Loaatlou Qual. ~T e 0~orator Maoufaoturar Class (ANSI I) Conditions (Inches-IO) Function 3SM-V870SA-I SM RAS (3) Check hP Rockwell 3 600 150 pslg I ESF Operation 8 140F 3SM-V871SB-I SM RAB (3) Check hp Rockwell 3 600 150 ps ig ESF Operation

& 140F 2CS-V136SN CS RAS (3) Check Rockwell 2 1500 2735 pslg ESF Operation 8 200 F lA Rockwell K

CS RAB (3). Check 2 1500 2735 pslg ESF Operation pc'CS-V137SN 8 200 F I

Ln 2CS-V138SN CS RAB (3) Check hp Rockwell 2 1500 ,2735 pslg ESF Operation 8 200 F RcqsLe 4'b t.

8 lgtSA RM. Atty %ttr 25P-IC308SS- I SP RCB (5 ) G I obe Solenoid Target-Rock 2 600 90 pslg 8 400 F

%31 RCPT. Lc 4b V'W'ISQ gc.h. Ho~itpr 2SP-~&I SP RAB (3) Globe Solenoid Target-Rock 2 600 90 pslg 8 400 F 'b cf RC'PB Lac.k v NSOSh gcuh Ho l4r 2SP-~OSS-I SP RCB (5) G lobe Solenoid Target-Rock 2 600 90 pslg 8 400'F gceB L V.be VMSISB QQ. Ho~t4r 2SP-~NB I SP RAS (3) G lobe Solenoid Target-Rock 2 600 90 ps lg A~cchptcs 8 400 F o~ 3CH B2SA I ESCMS RAB (3) Sutter f I y Ol aphragm ITT/Hamme I Dahl 3 150 150 ps I g Isolation Supply 8 125 F 3CH-B4SS-I ESCMS Supply RAB '3) Butterfly Diaphragm ITT/Hammel Oahl 3 150 150 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 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.

The monitor draws a continuous sample of containment air through a located inside the Containment.

~~.s~Egpol~~

ampled'oi s in the

'i reactor c ity, above e ch onta n nt are at e nort actor c sty, sou 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 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 37 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 380,000 (lb/hr)

Set pressure (psig) 2485 Design temperature (F) 650 Fluid Saturated steam Transient Condition (F)  : Non-Faulted Conditions 673 Faulted Conditions 682 Backpressure Normal (psig) 3 to 5 Expected during discharge (psig) 500 Throat Area (in ) 3.64

'Pressurizer Power 0 crated Relief Valves Number Design pressure (psig) 2485 Design temperature (F) 650 Relieving capacity at 2350 psig, per valve 210,000 (lb/hr)

Fluid Saturated steam Transient condition (F): Non-Faulted Conditions 673 Faulted Conditions 682 Throat Area (in )

Pressurizer S ra Valves Number Design Pressure, psig Design Temperature, Design Flow, F

for valves full open, each, gpm'485 650 350 37 5.4.13-3 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.

6.2.2.2.1.2 Design Description 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" p 37 Two of the four safety related fan cooler units are located at Elevation 236',

the remaining two safety related units are located at Elevation 286'.

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.

onents'rai.n

~

Train A Com B Corn onents Fan Cooler AH-2 Fan Cooler AH-1 Fan Cooler AH-3 - -'Farl Cooler AH-4 Service Water Loop A Service Water Loop B Emergency Power 'Diesel A Emergency Power 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 Normal Design Basis Operating Accident Conditions Conditions No. of Units 2 fans per 1 fan per unit unit and half speed, 4 2 units units starting operating and 2 units operating Fan Cooler 125,000 31,250 Unit Operating Capacity.ACFM Actual Air Mixture 62,500 31,250 Flow (ACFM) at Fan Inlet Design Ambient 45.0/39.1 ())

Pressure, psig Ambient Temp, F 120 258 Total Pressure, in. WG 7 ' 5 '

Fan RPM 1770 870 Outlet Velocity, FPM 5800 2560 Brake HP 101.2 32.8 Motor HP 125 62.5 J

Cooling Water 1500 Flow - GPM Entering Water 95 Temp, F NOTE: (1) 39.1 psig steam line break pressure 45.0 psig " maximum containment design pressure 37 6.2.2-16 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 d.l( PORC' F

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S 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'.

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'emperature Otu et'Cay 8 l IS is 40 or below: Only two p Ean cooler units) i7 a) When 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.

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37 6.2 2 4 Amendment ?lo. PZ

SliNPP FSAR units wx a total of 4.683x10 'tu/hr heat generated in a~nment.

With 90 F service water 'emperature, eac er has 2.83x10 Btu/hr.

heat removal capacity and is rated During this mode of operation, 37 two operating fan s will supply at a total o m and will remove a to x10 Btu/hr. heat generated in the Containment.

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L e~ c)~ S%~~h'4g. P~~ SHNPP FSAR e,s c(o.4. $ l~l ~g( coolers m'iLL toe 4 era "e'o h) When the Cou4o iu~em> ex@'croye, Q. temperatu e is above 80'W ttg ~ ~uc ~ Fan coo er ~ +~o 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 ~~gM"C located at Elevation 286: will operate with aaeAfanS ~ pi r unit operating at: full speed. Wro P~e C,.Z.2-'L 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 of operaL'ion bot:h Trains A and B enL'e ring tempera has 2.28x10 Btu/hr. hea ture, each fan c, -accident dischar be ope oval capacity and is z es. During this mode ating. With 95-F service water with two fans at full speed, at 125,000 cfm. t:otal e of, During this mode peration, all four operating fan ~oo 00 cfm and will remove a total of 7.3x10 Btu/hr. he Cont:ainment. will supply a nerated , 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. 37

6. 2. 2-5 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. 37 The containment equipment, heat gain is lighting, pi.ping, ~~0'tu/hr. 1'2 7 V-lo This includes heat contributed from motors as well as fan motors'. 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 signal (CSAS) which occurs when a containment pressure of 12.0 psig 'ctuation (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. 37

6. 2. 2-6 Amendment iVo. ~