ML18038A721

Petition for Emergency Enforcement Action Against Facility, Which Is Operating in Violation of NRC & Federal Requirements for Availability of ECCS High Pressure Core Injection & Request for Public Hearing

ML18038A721
Person / Time
Site:	Nine Mile Point
Issue date:	10/28/1992
From:	Riding B AFFILIATION NOT ASSIGNED
To:	NRC
Shared Package
ML18038A720	List: ML18038A719 ML18038A721
References
NUDOCS 9211160402
Download: ML18038A721 (148)

Text

UNITED STATED OF AMERICA BEFORE THE NUCLEAR REGULATORY COMMISSION PETITION FOR EMERGENCY ENFORCEMENT ACTION AND REQUEST FOR PUBLIC HEARING I.INTRODUCTION I, BEN L.RIDINGS (hereinafter"Petitioner")hereby petition the Commissioners of the Nuclear Regulatory Commission

("NRC" or"Commission")for emergency enforcement action against Niagra Mohawk's Nine Mile (Unit One)Nuclear power"plant, which is operatinq in violation of both the NRC and Federal requirements for availability of Emergency Core Cooling (ECCS)high pressure core injection.

As an ECCS system, the Nine Mile plant also fails to provide the mandatory emergency backup power to the high pressure core injection (HPCI)system.Over the twenty years the Nine Mile One plant has been allowed to operate, no safety related pumps have ever been available to inject water into the vessel at reactor pressure.At the same time this plant was allowed to operate at full power, there are many postulated accidents assumed in the Final Safety Analysis Report (that are capable of draining the reactor vessel)and specifically rely on the ECCS HPCI Pumps to'aintain reactor water level.These pumps have never been installed and the current administrative controls allowed this plant to operate outside the minimum federal requirement.

This specific type of plant operation outside the known minimum federal requirements greatly endangers health and property risk to the public.As discussed in detail below, the responsible utility, its Quality Assurance group and the NRC have routinely failed in their responsibility to ensure the operation of nuclear power plants within the license agreement.

Even when problems are identified,.

documented and brought to the attention of the responsible parties, various safety concerns are routinely dismissed, ignored or 9211160402 921027*PDR ADOCK 05000220 PDR

~~~M'h' administratively eliminated.

Even issues which obviously endanger public safety have been routinely dismissed, not only by the utility but such actions authorized and approved by the independent quality assurance groups and by the NRC.Any and all of these organizations have the authority to stop the op'eration of plants outside the minimum safety requirements, and not one have come forward to ful fill its duty and protect the public.Instead, each organization has reviewed the enclosed safety concerns and contrary to any practical justification, have remained silent and allowed this manner of plant operation to take place with their approval, giving evidence that these groups have also failed to remain independent of each other.Independent review by not only the government agency but the quality assurance review groups is the basic premise which allowed congress to grant operation of commercial nuclear power plants with limited liability for damages.The current administrative controls used today failed to ensure the plant operate within the minimum federal guidelines.

It is Congress's duty to protect public safety and its current administrative controls have failed.Because the Nine Nile Point Unit One Reactor violates both federal law and the Commissions's requirements for HIGH PRESSURE CORE INJECTION, the Commission can make no finding that there is resonable assurance of no undue risk to public health and safety.Petitioner therefore request that the Commission issue immediately an effective order directing the licensee to cease power operation and place the reactor in a cold shutdown condition.

The plant should not be permitted to continue or resume operation unless and until subsequent tests and inspections are shown to provide the requisite reasonable assurance of no undue risk to public health and safety.Moreover, Petitioners seek a public hearing before the plant is allowed to operate again.

I I.DESCRIPTION OF PETI TIINER I, Ben L.Ridings,'am a technical consultant for commercial nuclear power plants.Over a span of some fifteen years, while working at some twenty four sites, I have specialized in reviewing of licensing agreement (FSAR, Technical Speci f i cat i ons, Federal Codes and Regulations, ASME Codes, etc.), establishing administrative controls to meet these requirements and test programs to ensure compliance at all times.Ny test programs and administrat ve controls established while under contract to various utilities are still in use today at many facilities.

I I I~THE COMNISSIOM SHOULD EXERCISE ITS SUPERVISORY JURISDICTION OVER THIS PETITION A.The Commission has an Inherent Supervisory Jurisdiction over the Safety of Operation of the Niagra Nohawk Nine Nile Plant.This petition is brought before the Commission pursuant to the authority granted to it in 42USC 2233(d), 2236(a),2237 and 10CFR 2.204, 2.206(c)(1), 50.54, 50.57, 50.100 and 50.109.It invokes the inherent supervisory authority of the Commission to oversee all aspects of the regulatory and licensing process and its"overriding responsibility for assuring public health and safety in the operation of nuclear power facilities." Consolidated Edison Coo.of N.Y.Inc.(Indian Point, Units 1,2 and 3).CLI-75-8, 2 NRC 173 (1975)~As the Commission has previously observed, its supervisory powers include the power to order immediate shutdown of a facility"if the public health or safety so requires." t Petition for Emer enc and Remedial Action, CLI 78 6g 7 NRC 400'05 (1978)g citing 5 USC 558(c), 42 USC 2236(b)~10CFR 2 202(f)g 2 204.

t~U

~rs~i~f The Commission has exeicised its inherent authority on a number of occasions.

In addition to the ceases cited above, see Petition for Research and Develo ment Administration (Clinch River Breeder Reactor Project)p CLI 76 137 4 NRC 677 75 76(1976)I Consumers Power Co.(Midland Units 1 and 2), CLI-73-38, 6 AEC 1084 (1973);Public Service Co.of New N~ae shire (Seahroot Nuclear Power Station, Units i and 2), CL1-77-8, S NPC 503, 515-517(1977).

B.Exercise of the Commissions's Independent Jurisdiction is, Appropriate in This Case.NRC regulations at 10CFR2.206 provide that under ordinary circumstances, enforcement petitions are to be lodged with the NRC Staff, and that the Commission may take discretionary review of Staff denials of such petitions.

However, the Commissions's reviewing power"does not limit in any way" its"supervisory power over.delegated Staff actions", 10CFR2.206(c)(1).It is appropriate for the commission to exercise its supervisory powers and take jurisdiction in this case because the NPC Staff has acquiesced to Niagra Mohawks'iolations for more than two years.In Jan 1990, Niagra Mohawk Compliance Supervisor was given written notice of HPCI s and other inadequacies which effect public safety.After no apparent action, the Nine Mile Quality Fir st Team was also given notice.Petitioner was later notified by the Quality First Team that the NRC had been contacted and made aware of the problem as well.Petitioner was later contacted by the Quality First Team and told that the NRC had exempted the plant from the HPCI requirement and its need for backup power in the event of loss of power.Petitioner has yet to hear directly from the NRC on this matter.

II IV.GROUNDS FOR ACTION A.Federal Requirements for having radioactive fuels on site In accordance with 10CFR50.10, the utility Niagra Mohawk entered into contractual agreement with the federal government under the provisions of public document 50-220, on file with the federal register.Now under the Jurisdiction of 10CFR50, App.A (General Design Criteria), establish the minimum requirements for the principal design for water cooled nuclear power plant.Criterion 33 and 35 (Attachment 2)specify the minimum need that a system to provide abundant emergency core cooling shall be provided.The system safety function shall be to transfer heat from the.reactor core and must have suitable redundancy in components and on site electric power system (assuming offsite power is not available) which will enable the safety function to be accomplished.

Also (Criterion 33), a system to supply reactor coolant makeup for protection against small breaks in the reactor coolant pressure boundary shall be provided.Criterion 37 provides c)bu" g the testing requirements of the emergency core cooling system.10CFK70 details the utility and NRC responsibility for testing and inspection of these systems and 10CFR50 App.B (Quality Assurance Criteria)details the Quality Assurance Program and the administrative requirements for Inspections, Test Control, Operating Status, Corrective Action and Records.B.A Study of Contractual Agreement (docket 50-220)In, accordance with 10CFR50.34, the technical speci fication shall per form an evaluation of the safety ef fectiveness of providing for separation of high pressure coolant in Jection (HPCI)and reactor core isolation cooling (RCIC).This investigation found the Nile Nile Point Technical Specification in compliance with'his requirement.

Technical Specification 4.1~8(Attachment 3)gives positive proof that the ECCS i C:.

~gl~requirement for the HPCI system was anticipated by the designers.

Secondly, the corresponding Limiting Condition for Operation (LCO)3.i.8.c (Attachment 3)view this system as so critical that if"the utility fails to verify HPCI operability it will demand an orderly shutdown be initiated within one hour.When only one HPCI component becomes inoperable its redundant component shall be demonstrated to be operable immediately and daily thereafter (as opposed to monthly demonstration)." In accordance with the Bases for Technical Specification 3.1.8, the HPCI system is provided to ensure adequate core cooling in the unlikely event of a reactor coolant line break (also a federal requirement-design criterion 33).The HPCI system is required for line breaks which exceed the capability of the Control Rod Drive pumps and which are not large enough to allow fast enough depressurization for core spray to be effective (core spray 350 psi as opposed to HPCI 22QQ psi).In accordance with the Final Safety Analysis Report (FSAR), Chapter VII (Attachment 4), the Design Bases for HPCI is discussed.

Although several revision have been implemented by the utility'in order to fabricate the existence of a ECCS system to satisfy the HPCI federal requirement, its primary safety function is listed;(1)provide adequate cooling of the reactor core under abnormal and accident conditions, (2)remove the heat from radioactive decay and residual heat from the reactor core at such a rate that fuel clad melting would be prevented, (3)provide for continuity of core cooling over the complete range of postulated break sizes in the primary system process barrier.Once.the safety functions are understood it becomes obvious as to why this system is a minimum requirement of the federal guidelines.

~g 4 The following paragraph of FSAR Chapter VII gives the reader an indication of the lack of proper review that exists.At Nine Nile Point, unlike every other nuclear facility,"MPCI is not an engineered safeguards system and is not considered in any Loss of Coolant Accident Analysis." As stated in the FSAR (in layman terms)this feedwater system does not pretend to meet the IOCFR50 Appendix A (Criterion 33, 35, 36, 37)requirements of the minimum federal requirements.

In fact, Nine Nile Point has no system meeting these minimum federal requirements.

Next, reviewing the Design Evaluation portion of FSAR Chapter VII, (Attachment 4)a paradox occurs in design philosophy."During a loss-of-coolant accident within the drywel 1, high drywel 1 pressure due to a line break will cause a reactor scram.The automatic'cram will cause a turbine trip after a five-second delay.In order to revent claddin tern erature from exceedin their maximum limit for the entire.spectrum of breaks, the 3800 gpm (from one train of the HPCI pumps)would have to be available Obviously, the HPCI system is absolutely necessary to ensure critical heat flux (CHF)is not exceeded.Without the coolant water to transfer the heat from the fuel to the coolant, the fuel rod would then heat up rapidly and fuel cladding would take place and cause'possible melt down unless the reactor were shutdown quickly.Further, once the critical heat flux was exceeded, the departure from nucleate boiling ratio (DNBR)would exceed its 1~25 limit.These limits are Technical Speci fication requirements as well but it gives an indication of the interdependence of the ECCS systems.To make a statement in a license that"HPCI has not been considered in any Loss of Coolant Accident Analyses" is a another indication of the lack of p~opci review that exists at Nine Nile Point.Every safety limit assumed j l'I

~~,'jest at the Nine Mile Point plant is jeopardized without the assurance that the fuel will remain covered at all times.The NRC has approved the non-safety related feedwater system as an appropriate substitute for an ECCS HPCI federal requirement.

What at first seems like a quibble about a single pump is in actuality a valid argument that every bases assumed by this license is null and void.At Nine Mile Point, standard basic thermal reactor design has been significantly altered in several ECCS systems.There are no HPCI or RCIC system to transfer heat from the reactor core.There is no way of taking steam away from the reactor and using this energy to drive a high pressure pump.Normally the HPCI pumps return the condensed steam (water)back into the vessel to maintain water level.At Nine Mile Point, there is no HPCI or RCIC systems.At Nine Mile Point, unlike normal reactor design, electrically driven, non-quality

~-related feedwater pumps are considered.

These non-quality related feedwater pumps supposedly fulfill the HPCI safety function and yet do not meet the electrical backup requirements.

It must be noted that the size of these electrical pumps make it impossible to have on-site power available in the event of loss of off-site power.On-site power availability is assumed in the bases of the FSAR.It is therefore impossible for this plant to fulfill the minimum safety obligation as dictated by federal statute of the known postulated accidents.

This same feedwater system (being non-quality related)was purchased as a non-quality related system.In this same system;piping, valvesf instrumentation, wiring, electrical components and control systems were all purchased and installed under non-quality related contractual provisions.

HPCI automatically initiates on a Loss Coolant Accident (LOCA)signal from the NSSS logic.The NSSS logic performs the ECCS safeguard functions and E

always installed under strict contractual mandates, which include training, quality assurance reviews, certified skilled craftsmen, etc.Secondly, the piping system, welding, hanger restraints and maintenance considerations were installed and maintained under non-quality related provisions as well.Again, ECCS safeguard systems are purchased, constructed and maintained under much stricter guidelines.

The feedwater system was never designed, purchased, built, maintained nor capable of ful filling the HPCI requirements of the federal guidelines.-

At Nine Nile Point the HPCI system simply does not exist.The administrative controls which allowed acceptance of such a non-quality related system to fulfill this mandatory ECCS federal requirement is not acceptable.

C.Knowledge of Existing Concerns The need for an operable ECCS HPCI System is mandatory as evidenced from the grounds for relief in this report.At Nine Nile Pointy'the Utility, Quality Assurance personnel and the NPC were well aware of this requirement.

F'r what ever reason, this plant was licensed by the NRC and allowed to operate without this mandatory requirement installed.

Attempts by these same parties to substitute non-quality related feedwater equipment to fulfill this mandatory safeguard function supports the fact the need for requirement was understood.

Even if non-quality related equipment was.acceptable to support ECCS functions (and its not), there is no onsite electric power system that will support the safety function of a feedwater/HPCI system.This elec.ric system is another mandatory minimum requirement (Attachment 2-Criterion 35).To prove the collaboration between all parties mentioned, the licensee attempts to take credit for r onsite power availability from the Benton Dam, some 100 miles away.Obviously the reviewers are aware of these mandatory requirements but there I

'I gg~'I" resolution to the safety concerns is not acceptable.

The possibility of a tornado destroying the switchyard is a known postulated accident that can occur.Without this power availability, the HPCI function cannot possibly be assumed, as stated in the FSAR Chapter VI I (Attachment 4).Every time the feedwater procedures were revised this issue would have to be reviewed.Everytime the FSAR (Chapter VII)was revised, the Technical Specifications revised or containment integrity was questioned this issue had to be reviewed in accordance with administrative requirements set out by the federal guidelines.

Everytime the Quality Assurance groups and NRC performed their independent audits and inspections this issue had to be reviewed.Everytime this plant was operated at modes 1 or 2, the responsible Senior Reactor Operator (SRO), who is specifically trained (10CFR50 App E)on these issues would have to question the validity of the current HPCI system.Every time the HPCI surveillance (monthly)was performed to ensure operability, the responsible SPO would have to question the validity of a non quality related feedwater system fulfilling the HPCI system.Taking credit for non-quality related equipment to fulfill the requirements of a ECCS safety function is not acceptable and it would be the SPO's responsibility to question the feedwater ability to perform this HPCI safety function.Of course, that is the another problem to consider, it would be the SRO's job.Although previously aware of the problem, on Jan 18, 1990, the Utility was served notice of these and other safety concern.If the non-quality related feedwater system was to supposedly ful fill the HPCI safety function, it failed to met the onsite electrical requirements and many of

'I'I the main flow path valves had never been included in the Inservice Test Program (10CFR50.55).Some 44 out of 47 val ves were currently not identi fied in the Inservice Test Program(ECCS Surveillance violation).

With such knowledge, the Utility, Quality Assurance group and the NRC allowed'the plant to start up and continue into full operating (mode 1)condition.

No pumps, no valves yet Technical Speci fi cat i on 4.1.8 (Attachment

.3)demands i f one valve is not demonstrated operable a daily surveillance is required to be performed.

This is just another lack of administrative control in which the review groups have failed to audit or review properly.Unfortunately, this dilemma is not unique to Nine Nile Point, Other plants were also somehow licensed without this mandatory HPCI capability.

That is another indicator of the type of review that has taken place at other facilities as well but eventually these plants installed the mandatory system.The most stunning fact of this investigation shows that after literally thousands of technical reviews performed by hundreds of"qualified personnel" working in different shifts, separate departments, sites or regions, have all failed to stop this facility from operating outside the minimum federal guidelines.

Every month during full power operation, the HPCI system is verified operable by a"qualified" Senior Peactor Operator and a sworn affidavit submitted each month by the Utility to the NPC attesting that all requirements have been fulfilled.

Obviously, the current system of checks and balances cannot stop this plant from operating outside these mandatory federal guidelines, an assumption falsely made by congress.11

~I D.Pesponsibi 1 i t ies 10CFR50 App.B details the administrative requirements for Test Control, Inspections, Operating Status, Corrective Action, Pecords and-independent Audits.These requirements are addressed in both the Technical Specifications and FSAR;Site specific administrative procedures detail utility and quality assurance staff position responsibilities.

10CFR50.70 detail the NRC inspections while IOCFRS0.72 detail report notification responsibilities for all parties.The NRC have their own administrative procedures which detail staff responsibilities.

NUREG-0800 details the USNRC tandard review plan for inservice testing of pumps and valves.All parties mentioned were required to have knowledge of the HPCI requirements at the level of review for which each individual was involved.These reviews require mandatory action.Despite all mentioned reviews this requirement was not met.On Jan 18, 1990 the Niagra Mohawk, Nine Mile Point Nuclear Regulatory Compliance Group were served notice of this and many other known safety concerns.On July 31, 1990 the Niagra Mohawk Quality First Team were served written notice.The NPC was noti fied and on and the Quality First Team notified petitioner that the NRC exempted the utility'from the requirement.

V.STATEMENT OF THE LA'W 1.There is a minimum requirement for a High Pressure Core Inje'ction ECCS Safeguard System at the Nine Mile Point Unit One facility.This requirement comes from the federal guidelines, Technical Specifications and FSAR minimum mandates.2.'o High Pressure Core Injection System meeting the safeguard federal guidelines exists at Nine Mile Point, Unit One.

E~

3.If the non-quality related feedwater system was to supposedly ful fill the HPCI safety function, it failed to met the onsite electrical requirements and many of the main flow path valves had never been included in the Inservice Test Program.4.If the HPCI System is not a safeguard system and is not considered in any Loss of Coolant Accident Analyses as stated in the FSAR Chapter VII, then no assumption can be made that the fuel will remain covered by the moderator and related safety limits set in the current license are null and void.Obviously unreviewed safety questions exist.5.Congress made an assumption of the current checks and balances that would never allow a plant to operate outside the minimum safety requirements set out in federal guidelines.

On this assumption, unlike any other industry, the nuclear industry has been allowed to operate under limited liability.

The utility, Quality Assurance Groups, NRC and Chief Executive Officer have received written notice of their failure to comply with the minimum federal guidelines and have administratively failed to comply with this issue.As discussed above, the Nine Mile Unit One Plant fails to comply with both the minimum federal and NRC's requirements for HPCI ECCS System.This has been acknowledged by the NRC Staff and is demonstrated unequivocally by the evidence in the public record.Moreover, the Staff has performed no valid analysis that meets the Commission's narrow criteria for continuing to operate in the absence of compliance.

Compliance with both Federal and NPC safety regulations is a prerequisite to safe operation of a nuclear power plant.In fact, as the NRC's Appeal Board has observed, regulatory 0 r~

~~and safety." Maine Yankee Atomic Power Com an ALAB-161, 6 AEC 1003, 1009(1973).

Compliance may not be avoided by arguing that, although an applicable regulation is not me, the public health and safety will still be protected.

for, once a regulation is adopted, the standards it embodies represent the Commission's definition of what is required to protect the public health and safety.Vermont Yankee Nuclear Power Cor.ALAB-138, 6 AEC 520, 528(1973)(emphasis added).The Commission's essential safety standards must be met, without regard to the cost or inconvenience of achieving compliance.

10CF'R50.109 See also Union of Concerned Scientists v NRC, 824 f.2d 108(DC Cir 1987).VI~REQUEST FOR RELIEF t F'r the reasons enumerated above, petitioner states that the following relief is requir ed: A.Immediate Shutdown Pending Demonstration of Regul atory Compliance.

As discussed above, the Nine Mile Point nuclear plant fails to comply with an array of fundamental requirements for HPCI ECCS mandatory requirements.

No exemptions to this requirement can possibly be justified without undue risks to public safety.Consistent with the requirements of the Atomic Energy Act, F'ederal mandatory requirements and NPC regulations, Petitioner therefore seeks immediate shutdown of the Nine Mile Point unit one reactor pending full compliance with the regulations.

In seeking this relief, Petitioner notes that maintaining ECCS systems necessary to metigate loss of coolant accidents is a regulatory goal that warrants the most immediate and stringent enforcemenC action.Nine Mile Point's noncompliance with Che federal minimum design criteria and the"cover up" activities of all responsible parCies which poses a safety risk

~~

t t'f commensurate, i f not graver, dimension than the suspicion of ECCS pipe cracking that caused the commission to order 23 plant shutdowns in l975.See Petition for Emer enc and remedial Actioni CLI 78 Sg 7 NRC 400'05(i978).

Like the ECCS pipe cracking, this plant doesn't even have the pipes, valves or pumps necessary to metigate a known postulated accident that effects known safety limits of the FSAR.This system is necessary for the cooling of the core during an accident and this system (which does not exist)is tfle only means to prevent a meltdown.Again, unlike normal ECCS systems which have redundant components and can therefore withstand~a single failure, this system does not exist and cannot be compensated for by any other system.Simply put, a small break described in the FSAR bases as a postulated accident will in all likelihood meltdown the reactor for lack of cooling.Because the containment is not designed to withstand a meltdown, such'n event would probably lead to an uncontained release of radioactivity to the public environment.

This utility is not insured for such an accident.B.Public Hearing T.';e issues raised by the Nine Mile Point's noncompliance with federal requirements raises grave safety questions of tremendous public importance.

Petitioner therefore request that before allowing the Nine Mile Point plant to continue operating, the Commission provide for public hearing, with rights of discovery and cross examination, to determine whether Nine Mile Point is in full compliance with all federal minimum requirements revelant to HPCI and public safety.Secondly, congress be notified that the administrative controls relied upon to grant the nuclear industry the immunity of liability have failed to ensure public safety.After literally thousands of reviews by"quali fied

personnel" from di f ferent disciplines, departments, sites and regions completed their review, not one came forward and demand this plant operate within the law as laid out by act of congress.Should noncompliance be found, many of these reviews demand mandatory action on the part of the reviewer.The petitioner has notified all responsible parties and after two years Nine Mile Point Unit One continues to operate outside the federal gui del ines at a tremendous r i sk to publ i c sa f ety.A congr essi onal investigation of this matter be ini t i ated immediately.

The petitioner's services were contracted by Niagra Mohawk to review and ensure administrative compliance to Technical Speci fication prior to Start-Up.A qualified group of ten began a laborious review and when enormous problems began to immerge.This group was disbanded immediately.

In Jan 1990, the Niagra Mohawk's Nuclear Pegulatory Compliance Staf f was given a detailed memo (Attachment 5)giving evidence that 45%of the containment isolation valves had administrative deficiencies.

Two weeks later the review group was disbanded prior to completion of their review.Along with HPCI concerns, containment isolation valves as found in the FSAR Table VI-3 had deficiencies with corresponding Technical Specification Tables 3.3.4 h 3.2.7.This plant had operated for twenty years and yet the license failed to even correspond to itself, let alone actual plant conditions.

These valves are required by federal guidelines to protect the public yet almost half had deficiencies.

Petitioner alleges that when concerns are identified, the concerns are routinely"covered up", dismissed or administratively exempted.A proper review of the Nine Mile Point Unit One Technical Specification 4.0.5 requirements and the comliance of the l r t test programs will show that the utility simply hired another review group that (for whatever reason)failed to document the deficiencies that truly exist.Nine Nile Point Unit One resumed full power operations even after the safety concerns were identified and documented.

This type of cover up is not unique to this plant and a congressional investigation of this matter be initiated immediately.

~IM SUNDRY There can be no justification for the operation of nuclear power plants outside the minimum requirements specified by act of congress.These are the minimum requirements deemed necessary by act of congress to grant the immunity of liability currently assumed by the utility.When public safety is Jeopardized by known postulated accidents, there can be no justi fication for the lack of action by the responsible parties in this instance.Simply put, this utility is not insured to operate in this manner.Respect fully submitted, Ben L.Ridings P.O.Box 1101 Kingston, TN 37763 P'

BIBLIOGRAPHY NODERN POWER PLANT ENGINEERING, Weisman h Eckart, 1985 Prentice-Hall Inc.20 V~4~10CFPSO.10,"Requirement of License." 10CFR50.46,"Acceptance cri teria for emergency core cooling systems for light water nuclear power reactors." 10CFR50.55a,"Codes and Standards." 1OCFP50.59,"Changes, tests and experiments." 6.10CFRS0.70,"Inspection, Noti fications." Pecords, Reports, 7.10CFRSO, Appendix A, General Design Criterion 33,"Peactor coolant makeup." 8.10CFP50, Appendix A, General Design Criterion 35,"Emergency core cooing." 9.10CFP50, Appendix A, General Design Criterion 36,"Inspection of emergency core cooing system." 10..10CFR50, Appendix A, General Design Criterion 37,"Testing of Emergency Core Cooling systems." 11.10CFR50, Appendix B, II."Quality Assurance Program" 12.10CFR50, Appendix B, III."Design'Control." 13.10CFR50, Appendix B, VI."Document Control." 14.10CFR50, Appendix B, X."Inspections." 1S.10CFR50, Appendix B, XI."Test Control." 16.10CFP50, Appendix B, XIV."Inspection, Test and Oper at in g St at us." 17.10CFR50, Appendix B, XVI."Corrective Action." 18.10CFR50, Appendix B, XVII."Quality Assurance Records." 19.10CFR50, Appendix E, F."Training." 20.Federal Register, Public Docket: 50-220, Niagra Nohawk, Unit One, Nine Nile Point Thermal Nuclear Reactor.-18"

UNITED STATED OF AMERICA BEFOPE THE NUCLEAR REGULATORY COMMISSION AFFIDAVIT OF BEN L.RIDINGS I, Ben L.Ridings do make oath and say: 1.My name is Ben L.Ridings.I am a technical consultant for commercial nuclear power plants.Over a span of some fifteen years, while working at some twenty four sites, I have specialized in reviewing of licensing agreement (FSAR, Technical Specifications, Federal Codes and Regulations, ASME Codes, etc.), establishing administrative controls to meet these requirements and test programs to ensure compliance at all times.My test programs and administrative controls established while under contract to various utilities are still in use today at many facilities.

2.I have reviewed all of the relevant publicly available correspondence between the Nuc I ear Regul at ory Commi ssi on and Ni agr a Mohawk dur ing the relvant time span.I am familar with NRC regulations and regulatory guidance governing High Pressure Core Injection.

3.The factual statement made in the attached Petition for Emergency Action and Request for public Hearing are true and correct to the best of my knowlege and belief.Ben L.Ridings Subscribed and sworn to before me this~l~day of Qf~,1992.My commision expires:

l'C Part SO, App.A Ctffcr(oa 2y~oscbfscct rcacffo(fy cosftof sysfctss CapabQ(fy.

The teactlvlty control systems shaB be designed to have a corn.bined capabUlty.

In coniuncUon with poison addIUon by the emergency core cooUng system.of reOably conttoOIng reactivity changes to assure that under postulated ac-cident conditions and with appropriate margin for studc rods the capabOIty to cool'he core ls maintained.

CHfcrfoa 3d-Reac(fv(fy I(scffa.The reac-Uvity control systems shall be designed with~pptoprlate Omits on thc potentISI amount~nd rate of teactlvlty Increase to assure that thc cffccts of Pos'LUlatcd tcscUvI(y accidents can neither (1)result ln damage to Lhe reac-tor coolant pressure boundary greater than Umlted local yielding nor (2)suffidently dis-'CUtb Chc co~lts support sttUCLUfcs or other reactor pressure vessel Internals to Impair slgnlflcanUy the cayabOlty to cool the core.These postuiated reacUvlty acci-dents shaB Indude consideration of rod efecUon (unless prevented by positive means), rod dropout, stcam Une rupture.changes in reactor coolant temperature and pressure, and cold water addIUon.Crffctfoa 29-Aefccffoa apafast astfcf pa(cd opcraffosaf occsttcscea The protec-tion and reactivity control systems shaB be designed to assure an exttcmdy high proba-blUty ol accomplishing their safety func-tions In the event ol antldpated operational IV.ilsM Systems Crffcrfoa 3P-Qsalffy of rcacfor coolast prcssure bousdary.Components which are patt of the reactor coolant pressure bounda-ry shaB be designed.fabricated, erecLed, and tested to the highest quaUty standards ptac-tlcaL Means shaB be provided for deLectlng'and.to Che extent pracUcaL IdenUfylng the location of the source of reactor coolant leakage.Crffcr(oa 31-Ftacfstc pretpcsffoa of reac for cooiaaf prcssure bousdary.The reactor coolant pressure boundary shaB be designed<<lth sufi Ident margin to assure that when sttc55cd under opclaUng.maintenance.

tc5L-inC, and postulated acddent conditions (1)the boundary behaves ln a nonbrft tie manner snd (2)U1c ytobabOIL7 of fapidly propagating ftsctutc is minI111IzccL Thc design shaB reflect consideration of service temPeratures and other conditions of the boundary matetM under operaUnc, mainte-nance, testing, and postulated acddent con-ditions and the uncertainties In determining (I)1nstctial ptopcttics, (2)Ulc cffccts of It~, radlaUon on material properties.

(2)residu-al, steady state and Cranslent stresses, and (4)SIse of QawL Crffet(oa 32-laspccffoa qf reacfor coo(-Oat yrcsssre boesdary.Components whkh are part of the reactor coolant pressure boundary shaB be dcslgncd Co permit (1)546 10 CFR Ch.1 (1 1~Ed.II)periodic Inspection and tesUng of Important ateas snd fcaLutes to assess their structural and leaktight Integrity, and (2)an Spy toprI.ate material sutveOlance program for the reactor prcssutc vesseL Crffcrfos 33-Rcacfor coo(Oaf tsakcup, h system to SUPPly reactor coolant makeup for protection againsC small breaks In the reactor coolant pressure boundary shall be providecL The system safety funcUon shall be to assure that specified acceptable fuel design Omits are noL exceeded as a result of reactor coolant loss due to leakage from the reactor coolant pressure boundary and rup-ture of smaU piping or other smaU compo.ncnts which atc part of the boundary.The system shall be designed to assure that fot onslte electrk power system operation (as-suming offslic power ls tloC svsOsblc)snd for offslte electric Power system opetatlon (assuming onslte power Is not avaOable)Lhe system safety funcUon can be accomplished U5lng U1c plyingi pumps, snd vsivc5 115cd to maintain coolant Inventory during norma)reactor operation.

Crffcr(os 34-Rcsfdaal heaf mnooaf.h system Co tcmove residual heat shaB be pro.videcL The system safety function SMl bc to transfer fhsion product decay heat and other tesldual heat from the reactot core at a rate audi that spedfled acceptable fuel design Omits and the design conditions of the reactor coolant ptcmute boundary are not exceedecL Suitable redundancy In components and features, and suitable IntetconnecUons, leak detecUon.and ISOISUon cayabOltles shaB be provided to assure Chat for onslte electric power system operaUon (assuming offslte power 15 noL svsOablc)and foF offslic clcc tric power system operaUon (amumlng onslte power ls not avaOable)the system safety funcUon can be accompUshccL assum-Ing a single faOure.Crffer(oa 39-Esccrpescy core cooifsp.h system to provide abundant emercency core cooOng shaB be ptovideL The system safety function shall be Co transfer heat from the reactor core foUowing any loss of reactor coolant at i rate such that (1)fuel and clad damage thaL could Interfere with contfriued effective cote ceoUng 15 prevented and (2)dsd meta)-water rcactka Is hlted Lo negU-glble amounts.Suitable tcdundaney ln components and features.and suitable Interconnections.

leak detection.

ISOISUon.and containment eapa-bOitles shall be provided to assure that for onsiic electrk power system operation (as~suming offslte power h not avaOable)and for offslte electric power system operaUon (assUnllng on5ltc powct!5 not avaOablc)Chc system safeLy funcUon can be accompUshed.

assursfng a singk faOure.Crffcrfoa 3g-fsspccffos Of esacrpescy core cooffsp sysfcts.The emergency cote gudcot Regulat 1s~gUng system sh appropriate pctlo tant components.

~tot ptcssUfc v des.Snd piping.t C pabOIty of the 5-Ct(ktios 3t-2'sp sys(cps T~Lcm shall bc dc.Stc periodic prem'Lo smutc (I)thc 5'Legtity of its colnl 1 and performance

.Of the system, an system as~whol;dose to design as.Of,U1c fUll opera>thi.system into~Uon of appUca'I Lion system, th>>ind emergency p;ation of the~.",.Cr(kt(os 39-(, h system to rett.containment shs-'afety function conststent with 1 sodated systems'and temperature

int acddent ani ably low levels..'

1Sultable redu.'features, and su I'detection, Isolat i bQltles shaB be c.iinslte electric I , sumlng offslte.for offsite elect~(assuming onsit.System Mety f1 assuming a sing~Crifcr(oa 39-'eat rc1sooaf 5)removal system i~ayproprlatc pe tant componen'.spray'nozzles, c'.tegrlty and cap Cr(ter(oa gp-tcmooaf sysfnr moval system~yproprlate pe~1 testing to~leaktlght Inte>the operabOlt~ctlve compon the operablUt:

and under con as practical t operational sa Into operation cable portions transfer betw power sources sodated cooUr Ct(tv{os ckasup.Syst

~p~>

)1-l-88 Edition)aine of bnportanL ss their structura) 4 (2)an appropri-program for the oolanl makcu)L A coolant makeup sall breaks In the boundary shall be ety function shaD acceptable fuel cded as a result of>leakage from the, boundary and cup.ther smaD compo.he boundary.The to assure that for wm operation (as-not available) and system operation: noL available) the , n be accomplhhed and valves used to)ry during normal'eal removal.A I heat shall be yro-~function shaD be et decay heat and the reactor core at cd acceptable fuel eslgn conditions of sure boundary are n nants and arcs.ectlons, leak.mpabIDUes shaD be for onslte electric>(assumine offsite end for offslCe elec-eraUon (assuming st)able)the system xompllshed.

assum-icy core coo(lap.A ant emergency core L The system safety ssfer heat from the~ny Ioss o!reactor sat (l)fuel and clad fere with continued r prevented and (2)n h limited to ncgll-tn components and.terconnectlons, leak I contabunent capa-d Lo assure thaL fot istem operation (as-s not available) and er system operation h not available) the can be accomplished.

.((on of cmcrpency the emergency core 4 lg Nuciaar Regulatory Commission cooling system shall be designed to permit appropr(ate Periodic InspecUon of tant component@, such as spray rings In the reactor pressure vessel, water Iniectlon nas-zles.and piping.to assure the Integrity and capability of the system.Cr((cr(on 2F-Tcrtfnp of emerpency core cool(op system.The emergency core coollne system shaD be designed to permit approprl-ate periodic Pressure and funcUonal testing to assure (l)the structural and leaktight In-tegrity of its components.

(2)the operablllty and performance of the active components of the system.and (3)the operablllty of the system as a whole and.under condfUons as dose to desbcn as practical, the performance of the fuD operational sequence that brings the system Into operation.

Including oper.ation of applicable portions of the yrotec-Uon system.Lhe transfer beCween normal and emergency power sources, and the oper-ation of the asscclated cooling water system.Cry(sr(on 4d-Con(a(nmcn(heal removal.A sysLem to remove heat from the reactor containment shall be prov(ded.Thc system safety lunctlon shall be to reduce rapidly.conshtent with the lunctlonlne of other as.sodated systems.the contalrunent pressure and temyeraLure following an)r l~fwool.ant acddent and maintain them at accept ably low leveh.Suitable redundancy ln components and features.and suitable Interconnections.

leak detection.

Isolation.

and containment caya-bllltles shall be yrovided to assure thaC for onsite electric Power system operaUon (as-suming of!site power h not available) ind for of!site electr@power system operation (assuming onslte power Is not available) the system safety function can be accomplhhed, assuming a slnelc!allure.Crffcr(on 39-fnspcc((oa qf conte(amen(heat remelt system.The containment heat remoral system shaD be designed ta permit~ppiopr(ate periodic Inspection of Impor tant components.

such as the torus, sumps, spray nuules.and plplne ta assure the in-.tegrity and capability of the system.Cr((cr(oa 40-Tcsfln p o/con(a(nmca(heal removal sys(cm.The containment heat re-moval system shall be designed to permit apprayrhCc perlodlc pressure and function-s l testing to assure (l)the strucLural and leaktlght Integrity of Its components.

(2)the operability and pecformance of the active components of Che system.and (3).the operablllty of the system as a whole.and under conditions as close Lo the design~s practical the performance o!Che full operational sequence that brings Ch'e system Into operatio'n.

Indudlng operaUon of appli-cable portions of the protection system.the transfer beLween normal and emergency power sources.and the.operation of the as ,: 'odated cooDn'g water system.CH(erfon ef-Con(a(nmcal a(mosphcre deans~Systems to control fhsion prod-Part 50, App A ucts hydrogen oxygen and other sub-stances which may be released into the reac-tor containment shaD be provided as neces-sary to reduce.conshtent with the functlon-Ing of other associated systems.the concen-'tration and quality o!fhsion products re-leased to thc environment following postu-lated accidents, and to control the concen-tration of hydrogen or axygen and other substances In the containment atmosphere followlne postulated accidents to assure that containment lnteerILy h maintained.

Each system shaD have suitable redundan.cy In components and features, and suitable Interconnections.

leak detection, Isolation, and conLalnmenL capabilities ta assure Chat (or onslte electric po~er system operaUon (assuming of!site power h nat available) and for of!site dcctric yower system operaLlon (assumtne onslte power h not available)

ILs safety function can be accomplhhed.

assum;Ing a single!allure.Cr((er(oa 42-laspcc((on of con(a(nmca(a(morphcre cleanup syslcms.The contaln-ment atmosphere cleanup systems shaD be designed to permit apprapHate yeriodh'In-spectlon of bnyortant components.

such as filter fcames.ducts, and piping Co assure the Integrity and capablDty o!the systema, Cr(ter(on 43-Tcstfnp of confafnment a!-mar yhere deanup sysfema The containment atmosphere deanup systems shaD be de-signed to permit appropriate periodic pres.sure and functional testing to assure (1)the structural and leakUght lntegrlLy ol Its com-ponents.(2)the operability and perfonn-ance of the'ctlre components of the sys-tems such as fans.filters, dampecs.pumps.~nd ralves and (3)the operabOIty of the sys-tems as a whole and.under condIUons as close to design as pracUca).the perfonnance of the full operational sequence that brings the systems Inta operation.

Indudlng oper ation of applicable yorUons af the yroteo.tlon system.the transfer between normal and emergency power sources.and the oper-ation of assodated systems.Cr(fer(oa 4S-Cool(np roofer.A system to transfer heat from structures, systems, and components bnportant ta safety, to an ulti-mate heaC sink shaD be provided.The system safety function shaD be ta transfer\he combined heaL load of these stcuctures.

systems.and components under normal op-erating and acddent conditiona, Suitable redundancy ln components and features.and suitable InterconnecUons.

leak detection.

and holatlon capabilities shall be provided ta assure that for onslte electric power system operation (assuming offsite power h not'available) and for of!site elec.tric power system oyeraUon (assuming onslte power h not avaDable)the system safety!uncUon can be accomplhhed.

assum-Ing a single!allure.

\

LIHI TING CONDITION FOR OPERATIOH SURVEILLANCE AE(UIAEHLHT

~A 3.1.8 IIIQI PRESSURE COOLANT I JECTIOH A icabi it;Applies to the operational status of the high pressure coolant injection syst~a.;O~bectiva:

To assure the capability of the high pressure coolant injection system to cool reactor fuel in the event of a 1 oss-o f-cool ant acci den t.~li ii)ggxf~~~e 1 a.Ouring the power operating con-dition whenever the reactor coolant pressure is greater than 110 psig and the reactor coolant temperature greater than saturation temperature, the high pressure coolant injection sys tem shall be operable except as specified in Specification"b" below.b.If a redundant component of the high pressure coolant injection system becomes inoperable the high pressure coolant injection shall be considered operable provl~led that the component is returned to an operable co>>dition wltl>>n 15 days and the additional sur-vei 1 lance required Is performed.

4.1.8 lllGII PAESSUAE COOLANT INJECTION A~il Applies to the periodic testi>>g require>>iunts for the high pressure coola>>t i>>juctiu>>systume~0b ective: To verify the operability of the high pressure coolant injection system.~Sec(fi cation: The high pressure coolant injection sur-veillance shall be performed as indicated below: a.At 1 eas t Enny 1'.-a e~aBIIQ".-cycie=.g

~Ail~<AA)AAAe~~)LM Automatic start-up of the hii)h pressure coolant injectio>>system shall bu demon-s tra ted.A.A Pump operability shall bo dutermI>>ud.

71 J I J' S~UA f JL At{C~f(}lJJJEML'N'I c.lf Specification"a" and"b" are not met, a normal orderly shutdown shall be initiated M>thin one hour and reactor coolant pressure and temp-erature shall be reduced to less than 110 psig and saturation temperature within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.c.Surveillance with I~no arable Co>>>l>>>aunt When a component becomes l>>operabio i ts redundant compo>>ant shall bu deulonstrdtud to be operable i>>mediately and dailj thereafter.

X J~7 BASES FOR 3.1.8 AKD 4.).8 HTGH'PRESSURE COOLAKT IKJB.i)OK I High Pressure Coolant In5ection System (HPCl)is provided to qnsure adequate core cooling in the unlike)y event of a))reactor coolant))he break.The HPCl System.is required.for line breaks which exceed the capability of the ntrol Rod Drive pumps and which are not large enough to allo~'fast'enough depressurization for core spray to be fective.e set of high pressure coolant in)ection pumps consists of a condensate pump, a feedwater booster pump and a motor iven feedwater pump.One set of pumps is capable of deliver)ng 3,000 gpm to the reactor vessel at reactor essure.The performance capability of HPCI alone and in con]unction with other systems to provide adequate core~oling for a spectrum of line breaks is discussed in the Fifth Supplement of the FSAR.i determining the operability of th~HPCl System, the required performance capability of various components shall be>ns)dered.

~The HPCl System shal'l be capable of meeting its pump head versus flow curve.The motor driven feedwater pump shall be capable of automatic initiation upon receipt of either an automatic turbine trip signal or reactor)ow-water-)eve) signal..~The Condenser hotwe)l~eve)shall not be less than 57 inches (75,000 gallons).e~)'he Condensate storage tanks inventory shall not be less than 105,000 gallons.-The motor-driven feedwater pump will automatlcal)y trip if reactor high~ater leve)is sustained for ten seconds and the associated pump downstream flow control valve and)ow f)ow control valve are not closed.ir)ng reactor start-up, operation and shutdown, the condensate and feedwater booster pumps are in operation.

At.actor pressures up to 450 psig, these pumps are capable of supplying the required 3,800 gpm.Above 450 pslg a)tor-dr)yen-feedwater pump is necessary tu provide the required f)ow rate.se capability of the condensate, feedwa'ooster and motor driven feedwater pumps will be demonstrated by their terat)on as part of the feedwater supply during normal station operation.

Stand-by pumps will be p)aced in service t least quarterly to supply feedwater during station operation.

An automatic system initiation test wil)be erformed at least once per operating cycle.This wl)l involve automatic starting of the motor driven feedwater pumps~d flow to the reactor vessel.Revised October 1, 1906 73 I J l l~)))y I.HIGH-PRESSURE COOLANT INJECTION 1.0 Desi n Bases The high-pressure coolant injection (HPCI)system is an operating'ode of the feedwater system available in the event of a small reactor coolant line break which e'xceeds the capability of the control rod drive pumps (0.003 ft2).HPCI along with one emergency cooling system has the capability of keeping the swollen reactor coolant level above the top of active fuel for small reactor coolant boundary breaks up to 0.07 ft2 for at least 1000 seconds.The HPCI system with one of the two emergency cooling systems and two core spray systems, will provide core cooling for the complete spectrum of break sizes up to the maximum design basis recirculation discharge line break (5.446 ft2).Its primary purpose is to: a.provide adequate cooling of the reactor core under abnormal and accident conditions.

Rev.7 2.0 b.remove the heat from radioactive decay and residual heat from the reactor core at such a rate that fuel clad melting would be prevented.

c.provide, for continuity of core cooling over the complete range of postulated"break sizes in the primary system process barrier.HPCI is not an engineered safeguards system and is not considered in any Loss of Coolant Accident Analyses.It is discussed in this section because of its capability to provide makeup water at reactor oper'ating pressure.~t.t The HPCI system utilizes the two condensate storage tanks,.the main condenser hotwell, two condensate pumps, condensate demineralizers, two feedwater booster pumps, feedwater heaters, two motor-driven feedwater pumps, an integrated control system and all associated piping and valves.The system is capable of delivering 7600 gpm into the reactor vessel at reactor pressure when using two trains of feedwater pumps.The condensate and feedwater booster pumps are capable of supplying the required 3,800 gpm at approximately reactor pressures up to 270 psig.Above 270 psig a motor-driven feedwater pump is necessary to provide the required flow rate.Rev.7 Rev.7 Rev.7 F

VII-61a The feedwater system pumps have recirculation lines with air operated flow control valves to prevent the pumps from operating against a closed system.In the event of loss of air pressure, these valves open recycling part of the HPCI flow to the hotwell.HPCI flow would be reduced to approximately 3,000 gpm at a reactor pressure of 1,150 psig and 3,800 gpm at a reactor pressure of 940 psig.Condensate inventory is maintained at an available minimum volume of 180,000 gallons.Rev.7'.0 Oesi n Evaluation During a loss-of-coolant accident within the drywell, high drywell pressure due to a line break will cause a reactor scram.This automatic scram will cause a turbine trip afte-a five-second delay.In order to prevent cladding temperatures from exceeding their maximum limit for the entire spectrum of breaks, the 3800 gpm<from one train of HPCI/feedwater pumps)would have to be available immediately.

Feedwater flow would be available for considerable time from the shaft-driven feedwater pump.The shaft-driven feedwater pump should coast down awhile the electric motor-driven condensate pumps and feedwater booster pumps would continue to operate.The coast down time to reach 3,800 gpm delivery to the core is approximately 3;2 minutes (Figure VII-17), since both the condensate and feedwater booster pumps will continue to operate on off-site power.The turbine trip will signal the motor-driven feedwater pump to start.The signal will be simultaneous with the start of the shaft pump coast down.The motor-driven feedwater pump will be up to'peed and capable of supplying 3,800 gpm in about ten seconds.As a backup, low reactor water level will also signal the motor-driven pump to start.The initiati'on signal transfers control from the normal feedwater to the HPCI instrumentation and controller which has been continuously tracking the normal feedwater control signal.Thus there will be a continuous supply of feedwater to the reactor.The HPCI single element control system will attempt to maintain reactor vessel water level at 65 inches or 72 inches (depending upon which pump, 11 or 12 respectively, is in service)with a maximum feedwater flow limit of 3800 gpm.Rev.7 P't VII-62 A sustained high reactor water level reactor protection system signal coincident with an open feedwater flow control valve will selectively trip the associated feedwater pump.The clutch of the shaft-driven pump will also be disengaged immediately upon high reactor water level.Should the reactor water level reach the lo'w level scram setpoint the motor driven pump that tripped on high reactor water level wi 11 restart.Hecessary feedwater pump recirculation is provided to allow for continued pump operation with the flow control valve closed.As feedwater is pumped out of the condenser hotwell, through'he selected equipment of the condensate and feedwater systems and into the reactor, the condenser hotwe'll level will fall.Since condensed steam from the turbine no longer replenishes the condenser hotwell, condensate will be transferred from the condensate storage tanks to the hotwell for makeup.The feedwater system pumps operate on 4160 v.Hhen the plant.is in operation, the power is supplied from the main generator through the station service transformer when the generator is on-line and connected to the grid.Hhen the main generator is off-line, the feedwater pumps are supplied with normal off-site power from the 115 KV system through the reserve transformers.

If a HPCI initiation signal should, occur, all HPCI/feedwater system pumps would start immediately with two feedwater pump trains available for HPCI injection using the single element feedwater control system for reactor vessel level control.If a major po~er disturbance were to occur that resulted in loss of the 115 KV power supply to the Nine Hile Point 115 KV bus, power would be restored from a generator located at the Bennetts Bridge Hydro Station.This generator would have the capacity of supplying approximately 6,000 KVA which is sufficient to operate one train of HPCI/feedwater.

system pumps.If HPCI initiation were to occur, the preferred feedwater train pumps (feedwater pump 12, feedwater booster pump 13, condensate pump 13)would start.The non-preferred train pumps, would be locked out on loss of off-site power and not start until the operator manually reset the lock out.If a preferred train pump had been locked out prior to the loss of off-site power, it would remain locked out and the non-preferred train backup pump would automatically start on HPCI initiation.

If both the preferred and backup pumps are running, the preferred pump would remain in service and the backup pump will trip.The Rev.7 4 A~tg VII-62a use of a Bennetts Bridge hydro generator, while not equivalent to an on-site emergency power source, provides a highly reliable alternate off-site power supply for the HPCI function of the feedwater system.4.0 Tests and Ins ections ev 7 Tests and inspections of the various components are described in Section XI-Steam to Power Conversion.

a l~

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~d'TS.Vglgle, a.v.~.~(~~a.~~.cv(ko)u~s(.~(~em.~~v-......FIP~MH:..~..~kJ AC~VA rnid&/p.A E k)h f cg~r~~-W I n y t'I MEMO FOR YOUR FILES Oct 27, 1992 TO: U.S.Nuclear Regulatory Commisson Executive Director for Operations Public Document Room i7i7 H Street Washington, DC 20555 FROM: Ben L.Ridings P.O.Box ii0i Kingston, TN 37763 Ref: Petition pursuant iOCFR2.206

Dear Sirs:

Enclosed for filing PETITION FOR EMERGENCY ENFORCEMENT ACTION AND REQUEST FOR PUBLIC HEARING.Respectfully submitted, Ben L.Ridings C g

UNITED STATED OF AMERICA BEFORE THE NUCLEAR REGULATORY COMMISSION PETITION FOR EMERGENCY ENFORCEMENT ACTION AND REQUEST FOR PUBLIC HEARING I.INTRODUCTION I, BEN L.RIDINGS (hereinafter"Petitioner")hereby petition the Commissioners of the Nuclear Regulatory Commission

("NRC" or"Commission")for emergency enforcement action against Niagra Mohawk's Nine Mile (Unit One)Nuclear power plant, which is operating in violation of both the NRC and Federal requirements for availability of Emergency Core Cooling (ECCS)high pressure core injection.

As an ECCS system, the Nine Mile plant also fails to provide the mandatory emergency backup power to the high pressure core injection (HPCI)system.Over the twenty years the Nine Mile One plant has been allowed to operate, no safety related pumps have ever been available to inject water into the vessel at reactor pressure.At the same time this plant was allowed to operate at full power, there are many postulated accidents assumed in the Final Safety Analysis Report (that are capable of draining the reactor vessel)and specifically rely on the ECCS HPCI Pumps to maintain reactor water level.These pumps have never been installed and the current administrative controls allowed this plant to operate outside the minimum federal requirement.

This specific type of plant operation outside the known minimum federal requirements greatly endangers health and property risk to the public.As discussed in detail below, the responsible utility, its Quality Assurance group and the NRC have routinely failed in their responsibility to ensure the operation of nuclear power plants within the license agreements Even when problems are identified, documented and brought to the attention of the responsible parties, various safety concerns are routinely dismissed, ignored or

'f'f 7 administratively eliminated.

Even issues which obviously endanger public safety have been routinely dismissed, not only by the utility but such actions authorized and approved by the independent quality assurance groups and by the NRC.Any and all of these organizations have the authority to stop the operation of plants outside the minimum safety requirements, and not one have come forward to fulfill its duty, and protect the public.Instead, each organization has reviewed the enclosed safety concerns and contrary to any practical justification, have remained silent and allowed this manner of plant operation to take place with their approval, givinq evidence that these groups have also failed to remain independent of each other.Independent review by not only the government agency but the quality assurance review groups is the basic premise which allowed congress to grant operation of commercial nuclear power plants with limited liability for damages.The current administrative controls used today failed to ensure the plant operate within the minimum federal guidelines.

It is Congress's duty to protect public safety and its current administrative controls have failed.Because the Nine Nile Point Unit One Reactor violates both federal law and the Commissions's requirements for HIGH PRESSURE CORE INJECTION, the Commission can make no finding that there is resonable assurance of no undue risk to public health and safety.Petitioner therefore request that the Commission issue immediately an effective order directing the licensee to cease power operation and place the reactor in a cold shutdown condition.

The plant should not be permitted to continue or resume operation unless and until subsequent tests and inspections are shown to provide the requisite reasonable assurance of no undue risk to public health and safety.Moreover, Petitioners seek a public hearing before the plant is allowed to operate again.

I I.DESCRIPTION OF PETITIONER I, Ben L.Ridings, am a technical consultant for commercial nuclear power plants.Over a span of some fifteen years, while working at some twenty four sites, I have specialized in reviewing of licensing agreement (FSAR, Technical Speci fications, Federal Codes and Regulations, ASME Codes, etc.), establishing administrative controls to meet these requirements and test programs to ensure compliance at all times.My test programs and administrative controls established while under contract to various utilities are still in use today at many facilities.

I I I.THE COMMISSION SHOULD EXERCISE ITS SUPERVISORY JURISDICTION OVER THIS PETITION A.The Commission has an Inherent Supervisory Jurisdiction over the Safety of Operation of the Niagra Mohawk Nine Mile Plant.This petition is brought before the Commission pursuant to the authority granted to it in 42USC 2233(d), 2236(a),2237 and 10CFR 2.204, 2.206(c)(1), 50.54, 50.57, 50.100 and 50.109.It invokes the inherent supervisory authority of the Commission to oversee all aspects of the regulatory and licensing process and its"overriding responsibility for assuring public health and safety in the operation of nuclear power facilities." Consolidated Edison Coo.of N.Y.Inc.(Indian Point, Units 1,2 and 3).CLI-75-8, 2 NRC 173 (1975).As the Commission has previously observed, its supervisory powers include the power to order immediate shutdown of a facility"if the public health or safety so requires." Petition for Emer enc and Remedial Action, CLI-78-6, 7 NRC 400, 405 (1978), citing 5 USC 558(c), 42 USC 2236(b), 10CFR 2.202(f), 2.204.

"4 i q(,)'

The Commission has exercised its inherent authority on a number of occasions.

In addition to the ceases cited above, see Petition for Research and Develo ment Administration (Clinch River Breeder Reactor Project), CLI-76-i3, 4 NRC 67, 75-76(i976);

Consumers Power Co.(Nidland Units i and 2), CLI-73-38, 6 AEC i084 (i973);Public Service Co.of New H~am shire (Seabrook Nuclear Poser Station, Units 1 and 2), CLI-77-S, 5 NRC 503, 5i5-5i7(i977).

B.Exercise of the Commissions's Independent Jurisdiction is Appropriate in This Case.NRC regulations at iOCFR2.206 provide that under ordinary circumstances, enforcement petitions are to be lodged with the NRC Staff, and that the Commission may take discretionary review of Staff denials of such petitions.

However, the Commissions's reviewing power"does not limit in any way" its"supervisory power over delegated Staff actions", iOCFR2.206(c)(i)

~It is appropriate for the commission to exercise its supervisory powers and take jurisdiction in this case because the NRC Staff has acquiesced to Niagra Mohawks'iolations for more than two years.In Jan i990, Niagra Mohawk Compliance Supervisor was given written notice of HPCI and other inadequacies which effect public safety.After no apparent action, the Nine Nile Quality First Team was also given notice.Petitioner was later notified by the Quality First Team that the NRC had been contacted and made aware of the problem as well.Petitioner was later contacted by the Quality First Team and told that the NRC had exempted the plant from the HPCI requirement and its need for backup power in the event of loss of power.Petitioner has yet to hear directly from the NRC on this matter.

IV.GROUNDS FOR ENFORCEMENT ACTION A.Federal Requirements for having radioactive fuels on s'ite In accordance with 10CFR50.10, the utility Niagra Mohawk entered into contractual agreement with the federal government under the provisions of public document 50-220, on file with the federal register.Now under the jurisdiction of 10CFR50, App.A (General Design Criteria), establish the minimum requirements for the principal design for water cooled nuclear power plant.Criterion 33 and 35 (Attachment 2)specify the minimum need that a system to provide abundant emergency core cooling shall be provided.The system safety function shall be to transfer heat from the reactor core and must have suitable redundancy in components and on site electric power system (assuming offsite power is not available) which will enable the safety function to be accomplished.

Also (Criterion 33), a system to supply reactor coolant makeup for protection against small breaks in the reactor coolant pressure boundary shall be provided.Criterion 37 provides the testing requirements of the emergency core cooling system.10CFR70 details the utility and NRC responsibility for testing and inspection of these systems and 10CFR50 App.B (Quality Assurance Criteria)details the Quality Assurance Program and the administrative requirements for Inspections, Test Control, Operating Status, Corrective Action and Records.B.A Study of Contractual Agreement (docket 50-220)In accordance with 10CFR50.34, the technical specification shall perform an evaluation of the safety effectiveness of providing for separation of high pressure coolant injection (HPCI)and reactor core isolation cooling (RCIC).This investigation found the Nile Mile Point Technical Specification in compliance with this requirement.

Technical Specification 4.1.8(Attachment 3)gives positive proof that the ECCS

requirement for the HPCI system was anticipated by the designers.

Secondly, the corresponding Limiting Condition for Operation (LCO)3.1.8.c (Attachment 3)view this system as so critical that if"the utility fails to verify HPCI operability it will demand an orderly shutdown be initiated within one hour.When only one HPCI component becomes inoperable its redundant component shall be demonstrated to be operable immediately and daily thereafter (as opposed to monthly demonstration)." In accordance with the Bases for Technical Specification 3.1.8, the HPCI system is provided to ensure adequate core cooling in the unlikely event of a reactor coolant line break (also a federal requirement-design criterion 33).The HPCI system is required for line breaks which exceed the capability of the Control Rod Drive pumps and which are not large enough to allow fast enough depressurization for core spray to be effective (core spray 350 psi as opposed to HPCI 2200 psi)~In accordance with the Final Safety Analysis Report (FSAR), Chapter VII (Attachment 4), the Design Bases for HPCI is discussed.

Although several revision have been implemented by the utility in order to fabricate the existence of a ECCS system to satisfy the HPCI federal requirement, its primary safety function is listed;(1)provide adequate cooling of the reactor core under abnormal and accident'onditions, (2)remove the heat from radioactive decay and residual heat from the reactor core at such a rate that fuel clad melting would be prevented, (3)provide for continuity of core cooling over the complete range of postulated break sizes in the primary system process barrier.Once the safety functions are understood it becomes obvious as to why this system is a minimum requirement of the federal guidelines.

The following paragraph of FSAR Chapter VII gives the reader an indication of the lack of proper review that exists.At Nine Mile Point, unlike every other nuclear facility,"HPCI is not an engineered safeguards system and is not considered in any Loss of Coolant Accident Analysis." As stated in the FSAR (in layman terms)this feedwater system does not pretend to meet the 10CFR50 Appendix A (Criterion 33, 35, 36, 37)requirements of the minimum federal requirements.

In fact, Nine Mile Point has no system meeting these minimum federal requirements.

Next, reviewing the Design Evaluation portion of FSAR Chapter VII, (Attachment 4)a paradox occurs in design philosophy."During a loss-of-coolant accident within the drywell, high drywell pressure due to a line break will cause a reactor scram.The automatic scram will cause a turbine trip after a five-second delay.In order to revent claddin tern erature from exceedin their maximum limit for the entire spectrum of breaks, the 3800 gpm (from one train of the HPCI pumps)would have to be available Obviously, the HPCI system is absolutely necessary to ensure critical heat flux (CHF)is not exceeded.Without the coolant water to transfer the heat from the fuel to the coolant, the fuel rod would then heat up rapidly and fuel cladding would take place and cause a possible melt down unless the reactor were shutdown quickly.Further, once the critical heat flux was exceeded, the departure from nucleate boiling ratio (DNBR)would exceed its 1.25 limit.These limits are Technical Specification requirements as well but it gives an indication of the interdependence of the ECCS systems.To make a statement in a license that"HPCI has not been considered in any Loss of Coolant Accident Analyses" is a another indication of the lack of prope: review that exists at Nine Mile Point.Every safety limit assumed

at the Nine Mile Point plant is jeopardized without the assurance that the fuel will remain covered at all times.The NRC has approved the non-safety related feedwater system as an appropriate substitute for an ECCB HPCI federal requirement.

What at first seems like a quibble about a single pump is in actuality a valid argument that every bases assumed by this license is null and void.At Nine Mile Point, standard basic thermal reactor design has been significantly altered in several ECCS systems.There are no HPCI or RCIC system to transfer heat from the reactor core.There is no way of taking steam away from the reactor and using this energy to drive a high pressure pump.Normally the HPCI pumps return the condensed steam (water)back into the vessel to maintain water level.At Nine Mile Point, there is no HPCI or RCIC systems.At Nine Mile Point, unlike normal reactor design, electrically driven, non-quality related feedwater pumps are considered.

These non-quality related feedwater pumps supposedly fulfill the HPCI safety function and yet do not meet the electrical backup requirements.

It must be noted that the size of these electrical pumps make it impossible to have on-site power available in the event of loss of off-site power.On-site power availability is assumed in the bases of the FSAR.It is therefore impossible for this plant to fulfill the minimum safety obligation as dictated by federal statute of the known postulated accidents.

This same feedwater system (being non-quality related)was purchased as a non-quality related system.In this same system;pipingf valves, instrumentation, wiring, electrical components and control systems were all purchased and installed under non-quality related contractual provisions.

HPCI automatically initiates on a Loss Coolant Accident (LOCA)signal from the NSSS logic.The NBBS logic performs the ECCB safeguard functions and

always installed under strict contractual mandates, which include training, quality assurance reviews, certified skilled craftsmen, etc.Secondly, the piping system, welding, hanger restraints and maintenance considerations were installed and maintained under non-quality related provisions as well~Again, ECCS safeguard systems are purchased, constructed and maintained under much stricter guidelines.

The feedwater system was never designed, purchased, built, maintained nor capable of fulfilling the MPCI requirements of the federal guidelines.

At Nine Nile Point the HPCI system simply does not existed The administrative controls which allowed acceptance of such a non-quality related system to fulfill this mandatory ECCS federal requirement is not acceptable.

C.Knowledge of Existing Concerns The need for an operable ECCS HPCI System is mandatory as evidenced from the grounds for relief in this report.At Nine Nile Point, the Utility, Quality Assurance personnel and the NRC were well aware of this requirements For what ever reason, this plant was licensed by the NRC and allowed to operate without this mandatory requirement installed.

Attempts by these same parties to substitute non-quality related feedwater equipment to fulfill this mandatory safeguard function supports the fact the need for requirement was understood.

Even if non-quality related equipment was acceptable to support ECCS functions (and its not), there is no onsite electric power system that will support the safety function of a feedwater/HPCI system.This electric system is another mandatory minimum requirement (Attachment 2-Criterion 35)~To prove the collaboration between all parties mentioned, the licensee attempts to take credit for onsite power availability from the Benton Dam, some 100 miles away.Obviously the reviewers are aware of these mandatory requirements but there

0 resolution to the safety concerns is not acceptable.

The possibility of a tornado destroying the switchyard is a known postulated accident that can occurs Without this power availability, the HPCI function cannot possibly be assumed, as stated in the FSAR Chapter VII (Attachment 4)~Every time the feedwater procedures were revised this issue would have to be reviewed.Everytime the FSAR (Chapter VII)was revised, the Technical Specifications revised or containment integrity was questioned this issue had to be reviewed in accordance with administrative requirements set out by the federal guidelines.

Everytime the Quality Assurance groups and NRC performed their independent audits and inspections this issue had to be reviewed.Everytime this plant was operated at modes i or 2, the responsible Senior Reactor Operator (SRO), who is specifically trained (10CFR50 App E)on these issues would have to question the validity of the current HPCI system.Every time the HPCI surveillance (monthly)was performed to ensure operability, the responsible SRO would have to question the validity of a non quality related feedwater system fulfilling the HPCI system.Taking credit for non-quality related equipment to fulfill the requirements of a ECCS safety function is not acceptable and it would be the SRO's responsibility to question the feedwater ability to perform this HPCI safety function.Of course, that is the another problem to consider, it would be the SRO's job.Although previously aware of the problem, on Jan 18, 1990, the Utility was served notice of these and other safety concern.If the non-quality related feedwater system was to supposedly fulfill the HPCI safety function, it failed to met the onsite electrical requirements and many of

the main flow path valves had never been included in the Inser vice Test Program (iOCFR50.55)

~Some 44 out of 47 valves were currently not identi fied in the Inservice Test Program(ECCS Surveillance violation).

With such knowledge, the Utility, Quality Assurance group and the NRC allowed the plant to start up and continue into full operating (mode i)condi ti on.No pumps, no val ves yet Technical Speci f i cat i on 4.i.8 (Attachment 3)demands i f one valve is not demonstrated operable a daily surveillance is required to be performed.

This is just another lack of administrative control in which the review groups have failed to audit or review properly.Unfortunately, this dilemma is not unique to Nine Nile Point.Other plants were also somehow licensed without this mandatory HPCI capability.

That is another indicator of the type of review that has taken place at other facilities as well but eventually these plants installed the mandatory system.The most stunning fact of this investigation shows that after literally thousands of technical reviews performed by hundreds of"quali fied personnel" working in different shi fts, separate departments, sites or regions, have all failed to stop this facility from operating outside the minimum federal guidelines.

Every month during full power operation, the HPCI system is verified operable by a"qualified" Senior Reactor Operator and a sworn affidavit submitted each month by the Utility to the NRC attesting that all requirements have been fulfilled.

Obviously, the current system of checks and balances cannot stop this plant from operating outside these mandatory federal guidelines, an assumption falsely made by congress.

D.Responsibilities 10CFR50 App.B details the administrative requirements for Test Control, Inspections, Operating Status, Corrective Action, Records and independent Audits.These requirements are addressed in both the Technical Specifications and FSAR.Site specific administrative procedures detail utility and quality assurance staff position responsibilities.

10CFR50.70 detail the NRC inspections while 10CFR50.72 detail report notification responsibilities for all parties.The NRC have their own administrative procedures which detail staff responsibilities.

NUREG-0800 details the UBNRC standard review plan for inservice testing of pumps and valves.All parties mentioned were required to have knowledge of the HPCI requirements at the level of review for which each individual was involved.These reviews require mandatory action.Despite all mentioned reviews this requirement was not met.On Jan 18, 1990 the Niagra Nohawk, Nine Nile Point Nuclear Regulatory Compliance Group were served notice of this and many other known safety concerns.On July 31, 1990 the Niagra Mohawk Quality First Team were served written notice.The NRC was notified and on and the Quality First Team notified petitioner that the NRC exempted the utility from the r equi r ement.V.BTATENENT OF THE LAW i.There is a minimum requirement for a High Pressure Core Injection ECCS Safeguard System at the Nine Nile Point Unit One facility.This requirement comes from the federal guidelines, Technical Specifications and FSAR minimum mandates.2.No High Pressure Core Injection System meeting the safeguard federal guidelines exists at Nine Nile Point, Unit One.

I A , yA<<-n>mwq 1='I 3.If the non-quality related feedwater system was to supposedly fulfill the HPCI safety function, it failed to met the onsite electrical requirements and many of the main flow path valves had never been included in the Inservice Test Program.4.If the HPCI System is not a safeguard system and is not considered in any Loss of Coolant Accident Analyses as stated in the FSAR Chapter VII, then no assumption can be made that the fuel will remain covered by the moderator and related safety limits set in the current license are null and void.Obviously unreviewed safety questions exist.5.Congress made an assumption of the current checks and balances that would never allow a plant to operate outside the minimum safety requirements set out in federal guidelines.

On this assumption, unlike any other industry, the nuclear industry has been allowed to operate under limited liability.

The utility, Quality Assurance Groups, NRC and Chief Executive Officer have received written notice of their failure to comply with the minimum federal guidelines and have administratively failed to comply with this issue.As discussed above, the Nine Mile Unit One Plant fails to comply with both the minimum federal and NRC'requirements for HPCI ECCS System.This has been acknowledged by the NRC Staff and is demonstrated unequivocally by the evidence in the public record.Moreover, the Staff has performed no valid analysis that meets the Commission's narrow criteria for continuing to operate in the absence of compliance.

Compliance with both Federal and NRC safety regulations is a prerequisite to safe operation of a nuclear power plant.In fact, as the NRC's Appeal Board has observed, regulatory

and safety." Naine Yankee Atomic Power Com an ALAB-161, 6 AEC 1003, 1009(19?3).

Compliance may not be avoided by arguing that, although an applicable regulation is not me, the public health and safety will still be protected.

For, once a regulation is adopted, the standards it embodies represent the Commission's definition of what is required to protect the public health and safety.Vermont Yankee Nuclear Power Cor.ALAB-138, 6 AEC 520, 528(1973)(emphasis added).The Commission's essential safety standards must be met, without regard to the cost or inconvenience of achieving compliance.

10CFR50.109 See also Union of Concerned Scientists v NRC, 824 F.2d 108(DC Cir 1987)~VI.REQUEST FOR RELIEF For the reasons enumerated above, petitioner states that the following relief is required: A.Immediate Shutdown Pending Demonstration of Regulatory Compliance.

As discussed above, the Nine Nile Point nuclear plant fails to comply with an array of fundamental requirements for HPCI ECCS mandatory requirements.

No exemptions to this requirement can possibly be justified without undue risks to public safety.Consistent with the requirements of the Atomic Energy Act, Federal mandatory requirements and NRC regulations, Petitioner therefore seeks immediate shutdown of the Nine Nile Point unit one reactor pending full compliance with the regulations.

In seeking this relief, Petitioner notes that maintaining ECCS systems necessary to metigate loss of coolant accidents is a regulatory goal that warrants the most immediate and strinqent enforcement action.Nine Nile Point's noncompliance with the federal minimum design criteria and the"cover up" activities of all responsible parties which poses a safety risk I

of commensurate, if not graver, dimension than the suspicion of ECCS pipe cracking that caused the commission to order 23 plant shutdowns in 1975.See Petition for Emer enc and remedial Action, CLI-78-6, 7 NRC 400, 405(1978).

Like the ECCS pipe cracking, this plant doesn't even have the pipes, valves or pumps necessary to metigate a known postulated accident that effects known safety limits of the FSAR.This system is necessary for the cooling of the core during an accident and this system (which does not exist)is the only means to prevent a meltdown.Again, unlike normal ECCS systems which have redundant components and can therefore withstand a single failure, this system does not exist and cannot be compensated for by any other system.Simply put, a small break described in the FSAR bases as a postulated accident will in all likelihood meltdown the reactor for lack of cooling.Because the containment is not designed to withstand a meltdown, such an event would probably lead to an uncontained release of radioactivity to the public environment.

This utility is not insured for such an accident.B.Public Hearing The issues raised by the Nine Mile Point's noncompliance with federal requirements raises grave safety questions of tremendous public importance.

Petitioner therefore request that before allowing the Nine Mile Point plant to continue operating, the Commission provide for public hearing, with rights of discovery and cross examination, to determine whether Nine Mile Point is in full compliance with all federal minimum requirements revelant to HPCI and public safety.Secondly, congress be notified that the administrative controls relied upon to grant the nuclear industry the immunity of liability have failed to ensure public safety.After literally thousands of reviews by"qualified mz personnel" from di fferent disciplines, departments, sites and regions completed their review, not one came forward and demand this plant operate within the law as laid out by act of congress.Should noncompliance be found, many of these reviews demand mandatory action on the part of the reviewer.The petitioner has notified all responsible parties and after two years Nine Mile Point Unit One continues to operate outside the federal guidelines at a tremendous risk to public safety.A congressional investigation of this matter be initiated immediately.

The petitioner's services were contracted by Niagra Mohawk to review and ensure administrative compliance to Technical Specification prior to Start-Up.A qualified group of ten began a laborious review and when enormous problems began to immerge.This group was disbanded immediately.

In Jan l990, the Niagra Mohawk's Nuclear Regulatory Compliance Staff was given a detailed memo (Attachment 5)giving evidence that 45/of the containment isolation valves had administrative deficiencies.

Two weeks later the review group was disbanded prior to completion of their review.Along with HPCI concerns, containment isolation valves as found in the FSAR Table VI-3 had deficiencies with corresponding Technical Specification Tables 3.3.4 8<3.2.7.This plant had operated for twenty years and yet the license failed to even correspond to itself, let alone actual plant conditions.

These valves are required by federal guidelines to protect the public yet almost half had deficiencies.

Petitioner alleges that when concerns are identified, the concerns are routinely"covered up", dismissed or administratively exempted.A proper review of the Nine Mile Point Unit One Technical Specification 4.0.5 requirements and the comliance of the

test programs will show that the utility simply hired another review group that (for whatever reason)failed to document the deficiencies that truly exist.Nine Nile Point Unit One resumed full power operations even after the safety concerns were identified and documented.

This type of cover up is not unique to this plant and a congressional investigation of this matter be initiated immediately.

IN SUGARY There can be no justification for the operation of nuclear power plants outside the minimum requirements specified by act of congress.These are the minimum requirements deemed necessary by act of congress to grant the immunity of liability currently assumed by the utility.When public safety is jeopardized by known postulated accidents, there can be no justification for the lack of action by the responsible parties in this instance.Simply put, this utility is not insured to operate in this manner.Respectfully submitted, Ben L.Ridings P.O.Box ii01 Kingston, TN 37763 4 C<t~A<"l lL.1>a-t BIBLIOGRAPHY 1.NODERN POWER PLANT ENGINEERING, Weisman 5 Eckart, 1985 Prentice-Hall Inc.2.10CFR50.10,"Requirement of License." 3.10CFR50.46,"Acceptance criteria for emergency core cooling systems for light water nuclear power reactors." 4.10CFRS0.55a,"Codes and Standards." 5.10CFR50.59,"Changes, tests and experiments." 6.10CFR50.70,"Inspection, Records, Reports, Noti fi cations." 7.10CFR50, Appendix A, General Design Criterion 33,"Reactor coolant makeup." 8~10CFR50, Appendix A, General Design Criterion 35,"Emergency core cooing." 9.10CFR50, Appendi x A, General Design Cr i ter ion 36,"Inspection of emergency core cooing system." 10.10CFR50, Appendi x A, Gener al Desi gn Cr i t er i on 37,"Testing of Emerqency Core Cooling systems." 11.10CFRSO, Appendix B, II'Quality Assurance Program" 12.10CFR50, Appendix B, III."Design Control." 13.10CFR50, Appendix B, VI."Document Control," 14.10CFR50, Appendix B, X."Inspections." 15.10CFR50, Appendix B, XI'Test Control." 16.10CFR50, Appendix B, X I V."Inspection,-

Test and Operating Status." 17.10CFR50, Appendi x B, XVI."Corrective Action." 18.10CFR50, Appendix B, XVII."Quality Assurance Records." 19.10CFR50;Appendix E, F."Training." 20.Federal Register, Public Docket: 50-220, Niagra mohawk, Unit One, Nine Nile Point Thermal Nuclear Reactor.

UNITED STATED OF AMERICA BEFORE THE NUCLEAR REGULATORY COMMISSION AFFIDAVIT OF BEN L~RIDINGS I, Ben L.Ridings do make oath and say: 1.My name is Ben L.Ridings.I am a technical consultant for commercial nuclear power plants.Over a span of some fifteen years, while working at some twenty four sites, I have specialized in reviewing of licensing agreement (FSAR, Technical Specifications, Federal Codes and Regulations, ASME Codes, etc.), establishing administrative controls to meet these requirements and test programs to ensure compliance at al)times.My test programs and administrative controls established while under contract to various utilities are still in use today at many facilities.

2.I have reviewed all of the relevant publicly available correspondence between the Nuclear Regulatory Commission and Niagra Mohawk during the relvant time span.I am familar with NRC regulations and regulatory guidance governing High Pressure Core Injections 3.The factual statement made in the attached Petition for Emergency Action and Request for public Hearing are true and correct to the best of my knowlege and belief.Ben L.Ridings Subscribed and sworn to before me this~~day of Q~, 1992.My commision expires: ((<'I I g"gy((~I f, f,"., e c C'P~g,<cg<t'c"/p>>f)~~lA>>,

Cp r, f'b Part 50, App.A'riterion 27-Combfrrcd reactivity control systnns capability.

The reactivity control systems shall bc designed to have a corn.blned capability, In coniunctlon with poison addition by the emergency core cooling system, of reliably controlling reactivity changes to assure that under postulated ac-cident conditions and with appropriate margin for stuck rods the capability to cool thc core ls maintained.

Criterion 28-Reactfvffy limits.The reac-tivity control systems shall be designed wlLh appropriate limits on thc potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (1)result In damage to thc reac-tor coolant pressure boundary greater than limited local yielding nor (2)sufficiently dis-turb the core.Its support structures or other reactor pressure vessel internals to Impair significantly the capability to cool the core.These postulated reactivity acci-dents shall Include consideration of rod election (unless prevented by positive means), rod dropout, steam line rupture, changes ln reactor coolant temperature and pressure, and cold water addition.Criterion 29-Prefect(orr agafrrst anticf-patcd opcratfonal occurrences.

The protec-tion and reactivity control systems shall be designed to assure an extremely high proba-bility of accomplishing their safety func-tions In the event of anticipated operational occurrences.

I V.Fluid Sysfnrrs Crffcrion 30-Quality of reactor coolant pressure b'oundary.

Components which are part of the reactor coolant pressure bounda-ry shall be designed, fabricated, erected.and tested to the highest quality standards prac-tical.Means shall be provided for detecting and, to the extent practical, Identifying the location of the source of reactor coolant leakage.Crftcr(on 3I-Fracture prevent(on of reac-tor coolant prcssure boundary.The reactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, test-ing, and postulated accident conditions (1)the boundary behaves In a nonbrlttle manner and (2)the probability o!rapidly propagating fracture Is minimized.

Thc design shall reflect consideration of service temperatures and other conditions of the boundary material under operating, mainte-nance, testing, and postulated accident con.dltlons and the uncertainties In determining (I)material properties, (2)the effects of Ir-radiation on material properties, (3)residu-al, steady state and transient stresses.and<<')sm of naws.Crffcrion 32-Irrspccfforr of reactor cool-ant pressure boundary.Components which are part of the reactor coolant pressure boundary shall be designed to permit (1)loCFR Ch I (1 1~8 Edltlon)546 periodic insPcctlon and testing of ImportanL areas and features to assess their structural and lcaktlght integrity, and (2)an approprl.ate material surveillance program for the reactor pressure vessel.Crifcrfon 33-Reactor coolant makeup.A system to supply reactor coolant makeup for protection against small breaks In thc reactor coolant pressure boundary shall bc provided.The system safety function shall be to assure that specified acceptable fuel design limits are not exceeded as a result of reactor coolant loss due to leakage from the reactor coolant pressure boundary and rup-ture of small piping or other small compo.nents which are part of the boundary.The system shall be designed to assure that for onslte electric power system operation (as.sumlng offslte power ts not available) and for'ffslte electric power system operation (assuming onslte power ts not available) the system safety function can be accompltshcd using the plplnlr.Dumps, and valves used to maintain coolant Inventory during normal reactor operation.

Criterion 36-Residual heat rcmovaL A system to remove residual heat shall be pro.vlded.The system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design Ilrnlts and the design conditions of the reactor coolant pressure boundary are not exceeded.Suitable redundancy In components and features, and suitable Interconnections, leak detection, and Isolation capabilities shall be provided to assure that for onsltc electric power system operation (assuming offslte power ls not available) and for offslte elec-'ric power system operation (assuming onslte power h not available) the system safety function can be accomplished.

assum-Ing a single failure.Crffcrforr 35-Emergnrcy core cooHng.A system to provide abundant emergency core cooling shall be provided.The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1)fuel and clad damage thaL could Interfere with continued effective core cooling Is prevented and (2)clad metal-water reaction Is limited to negli-gible amounts.Suitable redundancy In components and features, and suitable Interconnections, leak detection.

Isolation, and containment capa-bilities shall be provided to assure that for ons!Le electric power system operation (as-suming offslte power Is not available) and for offslte electric Power system operation (assuming onslte power Is not avallablc) the system safety function can be accomplished.

assuming a single failure.Crf ter(on 36-Irrspccliorr of cmcrgnrcy core coolirrp system.The emergency core vcr'i Nuclear ReSuiator-4~"-'ing system shall appropriate perlodt'ant components.

su tor pressure ves ae and piping Lor>Pabtiity oi the sys'-.Crftcriorr 37-Tcs Ifnp sysicm.'The s~tem shall be desb-.'ate periodic press<<>~;p assure(1)the str'cgrity of ILs comPo>r(and performance or , of the system, and ('system as a whole r.dose to design as pr:of.the full operatlo , the, system into ot, ation of applicable

,tlon system, the L~"and emergency poa.;,ation of the assocla~.~.."<Cr(tcrion 38-Co: '.h system to rcmov~,".containment shall safety function sh consistent with thr soclated systems, t-and temperature f ,ant accident and r~r ably Iow levels.'W-;:jSultable redund: 'features and sulta'r',detection, Isolatlor~bllltles shall be pr~<lonslte electric pov<sumlng offslte po':for offslte electric (assuming onslte g;system safety fun(.'ssuming a single:~Criterion 39-Ir.;$heat removal systr ,, removal system sl appropriate perte" tant components,.:.spray'nozzles, anr ,'S,c,'.tegrlty and capabl Cri tcrion 40-Tc removal system.'oval system sh: appropriate perio~,,'l testing to assr leaktlght Intcgrlr the operability active componcn'he operablllLy c and under condlt as practical the operational scour Into operation, Ir.=cable portions of transfer betweer': f-power sources, ar soclatcd cooling s 4";,'riterion 4I-cleanup.System Q(('I

.'1-1-88 Edition)>ting of Important ss Lhclr strucLural id (2)an approprl-program for the volant makeup.A'oolant rnakcup>all breaks ln the boundary shall be cty function shall>d scccptablc fuel cded as a result of>leakage from the>oundary and rup-ther small compo-.he boundary.The to assure that for wm operation (as-not available) and'ystem operation: not available) the a bc accomplished and valves used to>ry during normal;heal removal.A I heat shall be pro->function shall be<ct decay heat and the reactor core at.ed acceptable fuel cslgn conditions of sure boundary are n nents and>cree.cctlons, leak>apabllltlcs shall be for onslte electric>(assuming offslte u>d for offslte elec->eration (assuming>liable)thc system xompilshed, assurn->cy core coolfnp.A snt emergency core L The system safeLy>sfer heat from the any loss of reactor>ai, (1)fuel and clad fere wlCh continued>prevented and (2)n Is limited to negll-In components and>terconnectlons.

leak I containment capa-d to assure that for rstem operation (as->not available) snd er system operation Is not available) the can be accompltshed.

e;lion of cmcrpency fhe emergency core J~t Nuclear Regulatory Commission cooling system shall be designed to permit appropriate periodic Inspection of Impor-tant componenLs, such as spray rings In thc reactor pressure vessel, water Infection noz-stcs.snd piping.to assure the Integrity snd capability of the system.Crifcrion 37-Tcsf(np of emergency core coolinp system.The emergency core cooling system shall be designed to permit spproprl.ate periodic pressure and functional testing to assure (1)the structural snd leaktlght In-tegrity of Its components, (2)the operabglty and performance of Lhe active components of the system, and (3)the operability of thc system as a whole and, under conditions ss close to design as practical, the performance of the full operational sequence that brings the system Into operation, Including oper-ation of spp)lcsble portions of the protec-tion system, the transfer between normal snd emergency power sources, and the oper-ation of the associated cooling water system.Crifcrion 38-Confainmen(heat removal.A system to remove hest from Lhe reactor contslrunent shall be provided.The system safety function shall bc to reduce rapidly, consistent with the functioning of other as.soclated systems, the containment pressure and temperature following any toss.ofwool-ant accident and maintain them at accept-ably low levels.Sultab)e redundancy In components and features.and suitable Interconnections, leak detection.

Isolation, and containment caps.bllltles shall be provided to assure that for onslte electric power system operation (as-suming offsite power Is not available) and for offslte electric power system operation (assuming onsite power Is not available) the system safety function can be accomplished, assuming a single failure.Criferion 39-lnspcctfo>>

of conlainmcnl heal removal sysfcm.The contalnmcnt heat removal system shall be designed to permit appropriate periodic Inspection of impor-tant components, such as the torus, sumps.spray nozzles, and piping to assure the In-tegrity and capability of the system.Criterion 40-Tcsfinp of confafnmcn(heal removal system.The containment heaC re-moval system shall be designed to permit appropriate periodic pressure and function-al testing to assure (1)the structural and leaktlght Integrity of Its components.

(2)the opcrablllty and performance of the active components of the system.and (3)the operability of the system as a whole, and under conditions as close to the design as practica)the performance of the full operational sequence that brings thc system Into operaLlon, Including operation of sppll-csblc portions of the protection system, the transfer between normal and emergency power sources.and Che operation of the as-sociated cooling water system.Criteria>>41-Confainment almosphcrc cleanup.Systems to control fission prod-Port 50, App.A ucts, hydrogen, oxygen, and other sub-stances which may be released Into the reac-tor contslnmenC shall be provided as neces-sary to reduce, consistent with the function-ing of other associated systems, thc concen.tratlon and quality of fission products re-leased Lo the environment following postu.lated accidents, and to conLrol the concen.tratlon of hydrogen or oxygen and other substances ln the containment atmosphcrc following postulated accidents to assure thsC containment lnLegrlty Is maintained.

Each system shall have suitable redundan-cy In components and features.snd suitable Interconnections, leak detection, isolation.

and containment capabilities to assure that, for onsltc electric power system opcratlon (assuming oifslte power Is not available) and for offsite electric power system operaC!on (assuming onslte power Ls not available) lis safety function can be accomplished, sssum.Ing s single failure.Criterion 42-lnspec(ion of confainmenl almosphcrc cleanup sysfcms.The contain-mcnL atmosphere cleanup systems shall be designed Lo permlL appropr'late periodic'In-spection of Important components.

such as filter frames.ducts, and piping to assure the Integrity and cspsblHty of the systems.Crf feria>>43-Tcrffnp of confafnmcnl at.mosphcrc cleanup sysfcms.The containment atmosphere cleanup systems shall be de-signed to permit appropriate periodic pres-sure and functional testing to assure (1)thc structural and leaktlght Integrity of Its corn.ponents, (2)the operability and perform.ance of the active components of the sys.terns such as Isns, filters, dsmpers.pumps, and valves and (3)the operability of the sys-tems as a whole and, under conditions ss close to design as practical.

thc performance of the full operational sequence that brings thc systems into operation, Including oper-ation of applicable portions of the protec.Lion system, the transfer between normal and ernergcncy power sources, and the oper-ation of associated systems.Cr(ferion 44-Coohnp u>afer.A system to transfer heat from structures, systems, and components important to safety, to an ulti-mate heat sink shall be provided.The system safety function shall be to trsnsler Che combined hest load of these structures, systems, and components under normal op.crating and accident conditions.

Suitable redundancy In components and features, and suitable Interconnections, leak detection, and Isolation capabilities shall be provided to assure that for onslte electric power system operation (assuming offslte power Is noL available) and for offslte elec-tric power system operation (assuming onslte power Is nol, available) the system safety function can be accomplished, sssum-Ing a single failure.54'7 rr LIHI TING,CONDITION FOR OPERATIOH SURVEILLAHCE AEJUIR)'.HLHT 3.1.8 Illa)PRESSuRE COOLANT IHJECTION AJJ11 Applies to the operational status of the high pressure coolant injection syste<n.O~b<.c ti ve: To assure the capability of the high pressure coolant injection system to cool reactor fuel in the event of a loss-of-coolant accident.4.1.8 lllGII PRESSURE COOLANT INJECTION Applies to the periodic testi>>g requirum<.nts for the high pressure coolant i>>juctiu>>cyst<.<ll.

~Ob ective: To verify the operability of the high)iressuru coolant injection system.h 111<<1)a.During the power operating con-" dition, whenever the reactor coolant pressure is greater than 110 psig and the reactor coolant temperature greater than saturation temperature, the high pressure coolant injection systen>shall be operable except as specified in Speci fi cation"b" below.b.If a redundant component of the hinh pressure coolant injection system becomes inoperable the high pressure coolant injection shall be considered operable provided that the component is returned t.o an operable condition within 15 days and the additional sur-veillance rec)uired is performed.

~Sect ficat<on: The high pressure coolant i>>jectio>>sur-veillance shall be performed as indicated below: a.At 1 eas t&nay el8 er'at)'hgŽcycle.

Auto<natlc start-up of the lii<)h p)essur<~coolant injection system shall be demon-s trated.Pump operabili ty shall be deter<)i>>ud, 71 r

Lie r C S~UAVf LLAHCE A~F.IJI AI'.HLH'I'.=-

If Specification"a" and"b" are not met, a normal orderly shutdown shall be initiated w>thin one hour and reactor coolant pressure and temp-erature shall be reduced to less than 110 psig and saturation temperature within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.c.Surveillance wi th I~no crab le Coiigionvnt When a component becomes I>>operable its redundant component shall hu demonstrated to be operable iiuiedlately and daily thereafter.

'12 ll BASES FOR 3.1.8 AND 4.1.8 HIGH'PRESSURE COOLANT IHJEL))OM a High Pressure Coolant Injection System (KPCI)is.provided to ensure adequate'core cooling in the unlikely event of ntrol Rod Drive pumps and which are not large enough to allo~'fast enough depressur)zat on xl)reactor coo an'ne rea..e 1 t li"'k Th HPCI System.is required.for line breaks wh')ch exceed the capab)lity of the for core s ra to be p y feet ive.e se o g press r t f hi h ure coolant injection pumps consists of a condensate pump, a feedwater booster pump and a motor iven feedwater pump.One set of pumps)s capable of del)ver.ing 3,800 gpm to the r c reactor vessel at reactor essure.The performance capability of HPCI alone and in conjunction with other systems to prov)de adequate core so))ng for a spectrum of 1)ne breaks)s discussed)n the Fifth Supplement of the FSAR.t determ)ning the operability of the HPCI System, the required performance capab)l)ty of various components shall be>ns)dered.

~The KPCI System shall be capable of meeting its pump head versus flow curve.g T.he motor driven feeduater pump shall be capable of automatic initiation upon receipt of either an automatic turbine trip signal or reactor low-water-level signal.e.'he Condenser hotwell~eve)shall not be less than 57)nches (75,000 gallons).The Condensate storage tanks inventory shall not be less than 105,000 gallons.-The motor-driven feedwater pump will automatically tr)p)f reactor h)gh water level is susta)ned for ten seconds and the associated pump downstream flow control valve and low flow control valve are not closed.rr)ng reactor start-up, operat)on and shutdown, the condensate and feedwater booster pumps are in operation.

At eactor pressures up to 450 pslg, these pumps are capable of supply)ng the required 3,800 gpm.Above 450 ps)g a vtor-driven-feedwater pump is necessary to provide the required flow rate.h bilit of the condensate, feedwa-'booster and motor dr)ven feedwater pumps w)1)be demonstrated by their peration as part of the feedwater supply during normal station operation.

Stand-by pumps w)l 1 e p e capa y o w)11 be laced in service t least quarterly to supp)y feedwater during station operation.

An automatic system initiation test will be erformed at least once per operating cycle.Th)s w)ll involve automatic starting of the motor dr)ven feedwater pumps nd flow to the reactor vessel.Revised October 1, 1986 73 r

I.HIGH-PRESSURE COOLANT INJECTION 1.0 Desi n Sases The high-pressure coolant injection (HPCI)system is an operating mode of the feedwater system available in the event of a small reactor coolant line break which exceeds the capability of the control rod drive pumps (0.003 ft2).HPCI along with one emergency cooling system has the capability of keeping the swollen reactor coolant level above the top of active fuel for small reactor coolant boundary breaks up to 0.07 ft2 for at least 1000 seconds.The HPCI system with one of the two emergency cooling systems and two core spray systems, will provide core cooling for the complete spectrum of break sizes up to the maximum design basis recirculation discharge line break (5.446 ft2).Its primary purpose is to: a.provide adequate cooling of the reactor core under abnormal and accident conditions.

b.remove the heat from radioactive decay and residual heat from the reactor core at such a rate that fuel clad melting would be prevented.

c.provide for continuity of core cooling over the complete range of postulated"break sizes in the primary system process barrier.HPCI is not an engineered safeguards system and is not considered in any Loss of Coolant Accident Analyses.It is discussed in this section because of its capability to provide makeup water at reactor operating pressure.2.0 S stem Desi n The HPCI system utilizes the two condensate storage tanks, the main condenser hotwell, two condensate pumps, condensate demineralizers, two feedwater booster pumps, feedwater heaters, two motor-driven feedwater pumps, an integrated control system and all associated piping and valves.The system is capable of delivering 7600 gpm into the reactor vessel at reactor pressure when using two trains of feedwater pumps.The condensate and feedwater booster pumps are capable of supplying the required 3,800 gpm at approximately reactor pressures up to 270 psig.Above 270 psig a motor-driven feedwater pump is necessary to provide the required flow rate.

VII-6la The feedwater system pumps have recirculation lines with air operated flow control valves to prevent the pumps from operating against a closed system.In the event of loss of air pressure, these valves open recycling part of the HPCI flow to the hotwell.HPCI flow would be reduced to approximately 3,000 gpm at a reactor pressure of 1,150 psig and 3,800 gpm at a reactor pressure of 940 psig.Rev.7 Condensate inventory is maintained at an available minimum volume of 180,000 gallons.3.0 Desi n Evaluation During a loss-of-'coolant accident within the drywell, high drywell pressure due to a line break will cause a reactor scram.This automatic scram will cause a turbine trip afte" a five-second delay.In order to prevent cladding temperatures from exceeding their maximum limit for the entire spectrum of breaks, the 3800 gpm (from one train of HPCI/feedwater pumps).would have to be available immediately.

Feedwater flow would be available for considerable time from the shaft-driven feedwater pump.The shaft-driven feedwater pump would coast down while the electric motor-driven condensate pumps and feedwater booster pumps would continue to operate.The coast down time to reach 3,800 gpm delivery to the core is approximately 3.2 minutes (Figure VII-17), since both the condensate and feedwater booster pumps will continue to operate on off-site power.The turbine trip will s i gna1 the motor-driven feedwater pump to start.The signal will be simultaneous with the start of the shaft pump coast down.The motor-driven feedwater pump will be up to speed and capable of supplying 3,800 gpm in about: ten seconds.As a backup, low reactor water level wi 11 also signal the motor-driven pump to start.The initiation signal transfers control from the normal feedwater to the HPCI instrumentation and controller which has been continuously tracking the normal feedwater control signal.Thus there will be a continuous supply of feedwater to the reactor.The HPCI single element control system will attempt to maintain reactor vessel water level at 65 inches or 72 inches (depending upon which pump, ll or 12 respectively, is in service)with a maximum feedwater flow limit of 3800 gpm.Rev.7 r

VII-62 A sustained high reactor water level reactor protection system signal coincident with an open feedwater flow control valve will selectively trip the associated feedwater pump.The clutch of the shaft-driven pump will also be disengaged immediately upon high reactor water level.Should the reactor water level reach the low level scram setpoint the motor driven pump that tripped on high reactor water level will restart.Necessary feedwater pump recirculation is provided to allow for continued pump operation with the flow control valve closed.As feedwater is pumped out of the condenser hotwell, through the selected equipment of the condensate and feedwater systems and into the reactor, the condenser hotwell level will fall.Since condensed steam from the turbine no longer replenishes the condenser hotwell, condensate will be transferred from the condensate storage tanks to the hotwell for makeup.The feedwater system pumps operate on 4160 v.When the plant is in operation, the power is supplied from the main generator through the station service transformer when the generator is on-line and connected to the grid.When the main generator is off-line, the feedwater pumps are supplied with normal off-site power from the 115 KV system through the reserve transformers.

If a HPCI initiation signal should occur, all HPCI/feedwater system pumps would start immediately with two feedwater pump trains available for HPCI injection using the single element feedwater control system for reactor vessel level control.If a major power disturbance were to occur that resulted in loss of the 115 KV power supply to the Nine Nile Point 115 KV bus, power would be restored from a generator located at the Bennetts Bridge Hydro Station.This generator would have the capacity of supplying approximately 6,000 KVA which is sufficient to operate one train of HPCI/feedwater system pumps.If HPCI initiation were to occur, the preferred feedwater train pumps (feedwater pump 12, feedwater booster pump 13, condensate pump 13)'would start.The non-preferred tra'in pumps would be locked out on loss of off-site power and not start until the operator manually reset the lock out.If a preferred train pump had been locked out prior to the loss of off-site power, it would remain locked out and the non-preferred train backup pump would automatically start on HPCI initiation.

If both the preferred and backup pumps are running, the preferred pump would remain in service and the backup pump will trip.The Rev.7 0 r VI I-62 a use of a Bennetts Bridge hydro generator, while not equivalent to an on-site emergency power source, provides a highly reliable alternate off-site power supply for the HPCI function of the feedwater system.ev.7 4.0 Tests and Ins ections" Tests and inspections of the various components are described in Section XI-Steam to Power Conversion.

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