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

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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.

III ~ 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.

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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 s

HPCI 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

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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.

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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

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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

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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 10

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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 ficat i on 4. 1.8 (Attachment .3) demands if 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.

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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.

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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

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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 14-

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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 Like the ECCS pipe cracking, this plant doesn't even have 400'05(i978).

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 fety. 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

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test programs will show that the utility simply hired t 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.

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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 10CFPSO. 10, "Requirement of License. "

V~ 10CFR50. 46, "Acceptance cri teria for emergency core cooling systems for light water nuclear power reactors."

4~ 10CFR50.55a, "Codes and Standards."

1OCFP50.59, "Changes, tests and experiments."

6. 10CFRS0.70, "Inspection, Pecords, Reports, Noti fications."

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.

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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:

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Part SO, App. A 10 CFR Ch. 1 (1 1~ Ed.II ) gudcot Regulat 1s ~

Ctffcr(oa 2y~oscbfscct rcacffo(fy cosftof periodic Inspection and tesUng of Important gUng system sh sysfctss CapabQ(fy. The teactlvlty control ateas snd fcaLutes to assess their structural appropriate pctlo systems shaB be designed to have a corn. and leaktight Integrity, and (2) an Spy toprI. tant components. v bined capabUlty. In coniuncUon with poison ate material sutveOlance program for the ~tot ptcssUfc addIUon by the emergency core cooUng reactor prcssutc vesseL des. Snd piping. t system. of reOably conttoOIng reactivity Crffcrfos 33-Rcacfor coo(Oaf tsakcup, h C pabOIty of the 5 changes to assure that under postulated ac- system to SUPPly reactor coolant makeup Ct(ktios 3t cident conditions and with appropriate for protection againsC small breaks In the sys(cps T 2'sp margin for studc rods the capabOIty to cool reactor coolant pressure boundary shall be ~Lcm shall bc dc core ls maintained. providecL The system safety funcUon shall .Stc periodic prem

'he CHfcrfoa 3d-Reac(fv(fy I(scffa. The reac- be to assure that specified acceptable fuel 'Lo smutc (I) thc 5 Uvity control systems shall be designed with design Omits are noL exceeded as a result of 'Legtity of its colnl

~ pptoprlate Omits on thc potentISI amount reactor coolant loss due to leakage from the 1 and performance

~ nd rate of teactlvlty Increase to assure that . Of the system, an reactor coolant pressure boundary and rup-thc cffccts of Pos'LUlatcd tcscUvI(y accidents ture of smaU piping or other smaU compo. system as ~ whol can neither (1) result ln damage to Lhe reac- ncnts which atc part of the boundary. The ;dose to design as tor coolant pressure boundary greater than system shall be designed to assure that fot .Of,U1c fUll opera>

Umlted local yielding nor (2) suffidently dis- onslte electrk power system operation (as- thi. system into

'CUtb Chc co~ lts support sttUCLUfcs or suming offslic power ls tloC svsOsblc) snd ~Uon of appUca'I Lion system, th>>

other reactor pressure vessel Internals to Impair slgnlflcanUy the cayabOlty to cool the core. These postuiated reacUvlty acci-dents shaB Indude consideration of rod 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 ind emergency p

ation of the

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efecUon (unless prevented by positive maintain coolant Inventory during norma) , h system to rett means), rod dropout, stcam Une rupture. reactor operation. . containment shs changes in reactor coolant temperature and Crffcr(os 34-Rcsfdaal heaf mnooaf. h

-'afety function pressure, and cold water addIUon. conststent with 1 Crffctfoa 29-Aefccffoa apafast astfcf system Co tcmove residual heat shaB be pro. sodated systems videcL The system safety function SMl bc 'and temperature pa(cd opcraffosaf occsttcscea The protec- to transfer fhsion product decay heat and tion and reactivity control systems shaB be other tesldual heat from the reactot core at ;int acddent ani designed to assure an exttcmdy high proba- ably low levels.

blUty ol accomplishing their safety func- a rate audi that spedfled acceptable fuel .':1Sultable redu.

tions In the event ol antldpated operational design Omits and the design conditions of 'features, and su the reactor coolant ptcmute boundary are I'detection, Isolat not exceedecL i bQltles shaB be IV. ilsMSystems Suitable redundancy In components and c. iinslte electric I Crffcrfoa 3P-Qsalffy of rcacfor coolast features, and suitable IntetconnecUons, leak , sumlng offslte detecUon. and ISOISUon cayabOltles shaB be . for offsite elect prcssure bousdary. Components which are provided to assure Chat for onslte electric (assuming onsit patt of the reactor coolant pressure bounda- power system operaUon (assuming offslte

~

. System Mety f1 ry shaB be designed. fabricated, erecLed, and power 15 noL svsOablc) and foF offslic clcc assuming a sing tested to the highest quaUty standards ptac- tric power system operaUon (amumlng Crifcr(oa 39-tlcaL Means shaB be provided for deLectlng onslte power ls not avaOable) the system 'eat

~

rc1sooaf 5)

'and. to Che extent pracUcaL IdenUfylng the safety funcUon can be accompUshccL assum- removal system location of the source of reactor coolant leakage. Ing a single faOure.

Crffer(oa 39-Esccrpescy core cooifsp. h i~ ayproprlatc pe tant componen Crffcr(oa 31-Ftacfstc pretpcsffoa of reac '. spray'nozzles, for cooiaaf prcssure bousdary. The reactor system to provide abundant emercency core c'. tegrlty and cap coolant pressure boundary shaB be designed cooOng shaB be ptovideL The system safety

<<lth sufiIdent margin to assure that when function shall be Co transfer heat from the Cr(ter(oa gp-tcmooaf sysfnr i

sttc55cd under opclaUng. maintenance. tc5L- reactor core foUowing any loss of reactor inC, and postulated acddent conditions (1) coolant at rate such that (1) fuel and clad the boundary behaves ln a nonbrft tie damage thaL could Interfere with contfriued moval system

~ yproprlate pe manner snd (2) U1c ytobabOIL7 of fapidly effective cote ceoUng 15 prevented and (2) ~1 testing to ~

propagating ftsctutc is minI111IzccL Thc dsd meta)-water rcactka Is hlted Lo negU- leaktlght Inte>

the operabOlt design shaB reflect consideration of service glble amounts.

Suitable tcdundaney ln components and ~ ctlve compon temPeratures and other conditions of the boundary matetM under operaUnc, mainte- features. and suitable Interconnections. leak the operablUt:

nance, testing, and postulated acddent con- detection. ISOISUon. and containment eapa- and under con ditions and the uncertainties In determining bOitles shall be provided to assure that for as practical t (I) 1nstctial ptopcttics, (2) Ulc cffccts of It~, onsiic electrk power system operation (as ~ operational sa radlaUon on material properties. (2) residu- suming offslte power h not avaOable) and Into operation al, steady state and Cranslent stresses, and for offslte electric power system operaUon cable portions (4) SIse of QawL (assUnllng on5ltc powct !5 not avaOablc) Chc transfer betw Crffet(oa 32-laspccffoa qf reacfor coo(- system safeLy funcUon can be accompUshed. power sources Oat yrcsssre boesdary. Components whkh assursfng a singk faOure. sodated cooUr are part of the reactor coolant pressure Crffcrfoa 3g-fsspccffos Of esacrpescy Ct(tv{os boundary shaB be dcslgncd Co permit (1) core cooffsp sysfcts. The emergency cote ckasup. Syst 546

~p

~ >

)

1-l-88 Edition) Nuciaar Regulatory Commission Part 50, App A aine of bnportanL cooling system shall be designed to permit ucts hydrogen oxygen and other sub-ss their structura) appropr(ate Periodic InspecUon of stances which may be released into the reac-4 (2) an appropri- tant component@, such as spray rings In the tor containment shaD be provided as neces-program for the reactor pressure vessel, water Iniectlon nas- sary to reduce. conshtent with the functlon-zles. and piping. to assure the Integrity and Ing of other associated systems. the concen-oolanl makcu)L A capability of the system. 'tration and quality o! fhsion products re-coolant makeup Cr((cr(on 2F-Tcrtfnp of emerpency core leased to thc environment following postu-sall breaks In the cool(op system. The emergency core coollne lated accidents, and to control the concen-boundary shall be system shaD be designed to permit approprl- tration of hydrogen or axygen and other ety function shaD ate periodic Pressure and funcUonal testing substances In the containment atmosphere acceptable fuel to assure (l) the structural and leaktight In- followlne postulated accidents to assure cded as a result of tegrity of its components. (2) the operablllty that containment lnteerILy h maintained.

> leakage from the, and performance of the active components system shaD have suitable redundan.

boundary and cup. of the system. and (3) the operablllty of the cyEachIn components and features, and suitable ther smaD compo. system as a whole and. under condfUons as Interconnections. leak detection, Isolation, he boundary. The dose to desbcn as practical, the performance conLalnmenL capabilities ta assure Chat of the fuD operational sequence that brings and 4

to assure that for (or onslte electric po~er system operaUon wm operation (as- the system Into operation. Including oper. (assuming of!site power h nat available) and not available) and ation of applicable portions of the yrotec-lg Uon system. Lhe transfer beCween normal for of!site dcctric yower system operaLlon system operation and emergency power sources, and the oper- (assumtne onslte power h not available) ILs

noL available) the safety function can be accomplhhed. assum; n be accomplhhed ation of the asscclated cooling water system.

Cry(sr(on 4d-Con(a(nmcn( heal removal. Ing a single !allure.

and valves used to A sysLem to remove heat from the reactor Cr((er(oa 42-laspcc((on of con(a(nmca(

)ry during normal containment shall be prov(ded. Thc system a(morphcre cleanup syslcms. The contaln-

'eal removal. A safety lunctlon shall be to reduce rapidly. ment atmosphere cleanup systems shaD be conshtent with the lunctlonlne of other as. designed to permit apprapHate yeriodh'In-I heat shall be yro- sodated systems. the contalrunent pressure spectlon of bnyortant components. such as

~ function shaD be and temyeraLure following an)r l~fwool. filter fcames. ducts, and piping Co assure the et decay heat and ant acddent and maintain them at accept Integrity and capablDty o! the systema, the reactor core at ably low leveh. Cr(ter(on 43-Tcstfnp of confafnment a!-

cd acceptable fuel Suitable redundancy ln components and mar yhere deanup sysfema The containment eslgn conditions of features. and suitable Interconnections. leak atmosphere deanup systems shaD be de-sure boundary are detection. Isolation. and containment caya- signed to permit appropriate periodic pres.

bllltles shall be yrovided to assure thaC for sure and functional testing to assure (1) the n nants and onsite electric Power system operaUon (as- structural and leakUght lntegrlLy ol Its com-arcs . ectlons, leak. suming of!site power h not available) ind ponents. (2) the operability and perfonn-mpabIDUes shaD be for of!site electr@ power system operation ance of the'ctlre components of the sys-for onslte electric (assuming onslte power Is not available) the tems such as fans. filters, dampecs. pumps.

> (assumine offsite system safety function can be accomplhhed, ~ nd ralves and (3) the operabOIty of the sys-end for offslCe elec- assuming a slnelc !allure.

eraUon tems as a whole and. under condIUons as (assuming Crffcr(on 39-fnspcc((oa qf conte(amen( to design as pracUca). the perfonnance st)able) the system heat remelt system. The containment heat close the full operational sequence that brings xompllshed. assum- remoral system shaD be designed ta permit of the systems Inta operation. Indudlng oper

~ ppiopr(ate periodic Inspection of Impor icy core coo(lap. A tant components. such as the torus, sumps, ation of applicable yorUons af the yroteo.

ant emergency core spray nuules. and plplne ta assure the in- tlon system. the transfer between normal and emergency power sources. and the oper-L The system safety . tegrity and capability of the system.

ssfer heat from the Cr((cr(oa 40-Tcsfln p o/con(a(nmca( heal ation of assodated systems.

~ ny Ioss o! reactor removal sys(cm. The containment heat re- Cr(fer(oa 4S-Cool(np roofer. A system to sat (l) fuel and clad moval system shall be designed to permit transfer heat from structures, systems, and fere with continued apprayrhCc perlodlc pressure and function- components bnportant ta safety, to an ulti-r prevented and (2) s l testing to assure (l) the strucLural and mate heaC sink shaD be provided. The n h limited to ncgll- leaktlght Integrity of Its components. (2) system safety function shaD be ta transfer the operability and pecformance of the \he combined heaL load of these stcuctures.

tn components and active components of Che system. and (3) systems. and components under normal op-

.terconnectlons, leak . the operablllty of the system as a whole.

erating and acddent conditiona, I contabunent capa- and under conditions as close Lo the design Suitable redundancy ln components and d Lo assure thaL fot ~ s practical the performance o! Che full features. and suitable InterconnecUons. leak istem operation (as- operational sequence that brings Ch'e system detection. and holatlon capabilities shall be s not available) and Into operatio'n. Indudlng operaUon of appli- provided ta assure that for onslte electric er system operation cable portions of the protection system. the power system operation (assuming offsite h not available) the transfer beLween normal and emergency power h not'available) and for of!site elec.

can be accomplished. power sources. and the. operation of the as tric power system oyeraUon (assuming

,: 'odated cooDn'g water system.

CH(erfon ef-Con(a(nmcal a(mosphcre onslte power h not avaDable) the system safety !uncUon can be accomplhhed. assum-

.((on of cmcrpency the emergency core deans~ Systems to control fhsion prod- Ing a single !allure.

\

LIHITING CONDITION FOR OPERATIOH SURVEILLANCE AE(UIAEHLHT ~A 3.1.8 IIIQI PRESSURE COOLANT I JECTIOH 4.1.8 lllGII PAESSUAE COOLANT INJECTION A icabi it; A~il Applies to the operational status of the Applies to the periodic testi>>g require>>iunts high pressure coolant injection syst~a.; for the high pressure coola>>t i>>juctiu>> systume O~bectiva: ~0b ective:

To assure the capability of the high To verify the operability of the high pressure pressure coolant injection system to coolant injection system.

cool reactor fuel in the event of a 1 oss-o f-cool ant acci den t.

~li ii ~ ~ ~e 1

) ggxf ~Sec( fi cation:

a. Ouring the power operating con- The high pressure coolant injection sur-dition whenever the reactor veillance shall be performed as indicated coolant pressure is greater than below:

110 psig and the reactor coolant temperature greater than saturation a. At 1 eas t Enny 1'.- a e~aBIIQ".-cycie=.g

~ < AA)AAAe~~)LM temperature, the high pressure ~Ail coolant injection sys tem shall be Automatic start-up of the hii)h pressure operable except as specified in coolant injectio>> system shall bu demon-Specification "b" below. s tra ted.

b. If a redundant component of the high A. A pressure coolant injection system becomes inoperable the high pressure Pump operability shall bo dutermI>>ud.

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.

71

J I

J'

S~UA f JL At{C~f(}lJJJEML'N'I

c. lf Specification "a" and "b" are not c. Surveillance with I~no arable Co>>>l>>>aunt met, a normal orderly shutdown shall be initiated M>thin one hour and When a component becomes l>>operabio i ts reactor coolant pressure and temp- redundant compo>>ant shall bu deulonstrdtud erature shall be reduced to less than to be operable i>>mediately and dailj 110 psig and saturation temperature thereafter.

within 24 hours.

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.

October 1, 1906 73 Revised

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 Rev. 7 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 Rev. 7 its capability to provide makeup water at reactor oper'ating pressure.

2.0

.t

~t The HPCI system utilizes the two condensate storage Rev. 7 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 Rev. 7 necessary to provide the required flow rate.

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 Rev.

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.

7'.0 Condensate inventory is maintained at an available minimum volume of 180,000 gallons.

Oesi n Evaluation During a loss-of-coolant accident within the drywell, Rev. 7 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 and capable of supplying 3,800 gpm in about to'peed 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.

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 Rev. 7 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

4 A

~ tg

VII-62a use of a Bennetts Bridge hydro generator, while not equivalent to an on-site emergency power source, ev 7 provides a highly reliable alternate off-site power supply for the HPCI function of the feedwater system.

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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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, C

g Ben L. Ridings

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

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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.

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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 existed requirements of the federal guidelines. At Nine Nile Point the HPCI system simply does not 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 tion. No pumps, no val ves yet Technical Speci f i cat i on 4. i. 8 (Attachment 3) demands if 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.

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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

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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.

18

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:

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'i Part 50, App. A loCFR Ch I (1 1~8 Edltlon) Nuclear ReSuiator

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27-Combfrrcd reactivity control

'riterion periodic insPcctlon and testing of ImportanL "-'ing system shall systnns capability. The reactivity control areas and features to assess their structural appropriate components. su perlodt'ant systems shall bc designed to have a corn. and lcaktlght integrity, and (2) an approprl.

blned capability, In coniunctlon with poison ate material surveillance program for the tor pressure ves addition by the emergency core cooling reactor pressure vessel. ae and piping Lor system, of reliably controlling reactivity Crifcrfon 33-Reactor coolant makeup. A >Pabtiity oi the sys changes to assure that under postulated ac- system to supply reactor coolant makeup '-.Crftcriorr 37-Tcs cident conditions and with appropriate for protection against small breaks In thc Ifnp sysicm. 'The margin for stuck rods the capability to cool reactor coolant pressure boundary shall bc s~tem shall be desb thc core ls maintained. provided. The system safety function shall -.'ate periodic press<<

Criterion 28-Reactfvffy limits. The reac- be to assure that specified acceptable fuel >~;p assure(1) the str tivity control systems shall be designed wlLh design limits are not exceeded as a result of 'cgrity of ILs comPo>

appropriate limits on thc potential amount reactor coolant loss due to leakage from the r(and performance or and rate of reactivity increase to assure that , of the system, and (

reactor coolant pressure boundary and rup-the effects of postulated reactivity accidents ture of small piping or other small compo. 'system as a whole r can neither (1) result In damage to thc reac- nents which are part of the boundary. The .dose to design as pr tor coolant pressure boundary greater than :of.the full operatlo

, the, system into ot, system shall be designed to assure that for limited local yielding nor (2) sufficiently dis- onslte electric power system operation (as.

turb the core. Its support structures or sumlng offslte power ts not available) and ation of applicable other reactor pressure vessel internals to for'ffslte electric power system operation ,tlon system, the L Impair significantly the capability to cool ~"and emergency poa (assuming onslte power ts not available) the of the assocla the core. These postulated reactivity acci- system safety function can be accompltshcd .;,ation dents shall Include consideration of rod ~ .~.."<Cr(tcrion 38-Co:

using the plplnlr. Dumps, and valves used to '.h system to rcmov election (unless prevented by positive maintain coolant Inventory during normal means), rod dropout, steam line rupture, ~,". containment shall reactor operation. safety function sh changes ln reactor coolant temperature and Criterion 36-Residual heat rcmovaL A pressure, and cold water addition. consistent with thr Criterion 29-Prefect(orr agafrrst anticf- system to remove residual heat shall be pro. soclated systems, t patcd opcratfonal occurrences. The protec- vlded. The system safety function shall be -and temperature f tion and reactivity control systems shall be to transfer fission product decay heat and ,ant accident and r designed to assure an extremely high proba- other residual heat from the reactor core at ~r ably Iow levels.

bility of accomplishing their safety func- a rate such that specified acceptable fuel 'W -;:jSultable redund:

tions In the event of anticipated operational design Ilrnlts and the design conditions of 'features and sulta occurrences. the reactor coolant pressure boundary are 'r',detection, Isolatlor not exceeded. ~ bllltles shall be pr IV. Fluid Sysfnrrs Suitable redundancy In components and ~<lonslte electric pov Crffcrion 30-Quality of reactor coolant features, and suitable Interconnections, leak < sumlng offslte pressure b'oundary. Components which are detection, and Isolation capabilities shall be offslte electric po':for part of the reactor coolant pressure bounda- provided to assure that for onsltc electric (assuming onslte g ry shall be designed, fabricated, erected. and power system operation (assuming offslte  ; system safety fun(

tested to the highest quality standards prac- power ls not available) and for offslte elec- .'ssuming a single:

tical. Means shall be provided for detecting 'ric power system operation (assuming ~ Criterion 39-Ir.

and, to the extent practical, Identifying the onslte power h not available) the system  ;$ heat removal systr location of the source of reactor coolant safety function can be accomplished. assum- ,, removal system sl leakage. Ing a single failure. appropriate perte Crftcr(on 3I-Fracture prevent(on of reac- Crffcrforr 35-Emergnrcy core cooHng. A " tant components, tor coolant prcssure boundary. The reactor system to provide abundant emergency core .:. spray'nozzles, anr coolant pressure boundary shall be designed cooling shall be provided. The system safety ,'S,c,'. tegrlty and capabl with sufficient margin to assure that when function shall be to transfer heat from the Cri tcrion 40-Tc stressed under operating, maintenance, test- reactor core following any loss of reactor removal system.

ing, and postulated accident conditions (1) coolant at a rate such that (1) fuel and clad system sh: 'oval damage thaL could Interfere with continued appropriate perio the boundary behaves In a nonbrlttle manner and (2) the probability o! rapidly propagating fracture Is minimized. Thc effective core cooling Is prevented and (2) clad metal-water reaction Is limited to negli-

~,,'lleaktlght testing to assr Intcgrlr design shall reflect consideration of service gible amounts. the operability temperatures and other conditions of the Suitable redundancy In components and active componcn'he boundary material under operating, mainte- features, and suitable Interconnections, leak operablllLy c nance, testing, and postulated accident con. detection. Isolation, and containment capa- and under condlt dltlons and the uncertainties In determining bilities shall be provided to assure that for as practical the (I) material properties, (2) the effects of Ir- ons!Le electric power system operation (as- operational scour radiation on material properties, (3) residu- suming offslte power Is not available) and Into operation, Ir.

al, steady state and transient stresses. and for offslte electric Power system operation =cable portions of

<<') sm of naws. (assuming onslte power Is not avallablc) the transfer betweer Crffcrion 32-Irrspccfforr of reactor cool- system safety function can be accomplished. ': f- power sources, ar ant pressure boundary. Components which assuming a single failure. soclatcd cooling s are part of the reactor coolant pressure Crfter(on 36-Irrspccliorr of cmcrgnrcy 4";,'riterionSystem 4I-boundary shall be designed to permit (1) core coolirrp system. The emergency core cleanup.

Q((

546

'I

.'1-1-88 Edition) Nuclear Regulatory Commission Port 50, App. A

>ting of Important cooling system shall be designed to permit ucts, hydrogen, oxygen, and other sub-ss Lhclr strucLural appropriate periodic Inspection of Impor- stances which may be released Into the reac-id (2) an approprl- tant componenLs, such as spray rings In thc tor contslnmenC shall be provided as neces-program for the reactor pressure vessel, water Infection noz- sary to reduce, consistent with the function-stcs. snd piping. to assure the Integrity snd ing of other associated systems, thc concen.

volant makeup. A capability of the system. tratlon and quality of fission products re-

'oolant rnakcup Crifcrion 37-Tcsf(np of emergency core leased Lo the environment following postu.

>all breaks ln the coolinp system. The emergency core cooling lated accidents, and to conLrol the concen.

boundary shall be system shall be designed to permit spproprl. tratlon of hydrogen or oxygen and other cty function shall ate periodic pressure and functional testing substances ln the containment atmosphcrc

>d scccptablc fuel to assure (1) the structural snd leaktlght In- following postulated accidents to assure cded as a result of tegrity of Its components, (2) the operabglty thsC containment lnLegrlty Is maintained.

> leakage from the and performance of Lhe active components Each system shall have suitable redundan-

>oundary and rup- of the system, and (3) the operability of thc system as a whole and, under conditions ss cy In components and features. snd suitable ther small compo- Interconnections, leak detection, isolation.

.he boundary. The close to design as practical, the performance of the full operational sequence that brings and containment capabilities to assure that, to assure that for J for onsltc electric power system opcratlon wm operation (as- the system Into operation, Including oper-ation of spp)lcsble portions of the protec- (assuming oifslte power Is not available) and not available) and for offsite electric power system operaC!on

'ystem operation tion system, the transfer between normal snd emergency power sources, and the oper- (assuming onslte power Ls not available) lis

not available) the ~ t ation of the associated cooling water system. safety function can be accomplished, sssum.

a bc accomplished Crifcrion 38-Confainmen( heat removal. Ing s single failure.

and valves used to Criterion 42-lnspec(ion of confainmenl

>ry during normal A system to remove hest from Lhe reactor contslrunent shall be provided. The system almosphcrc cleanup sysfcms. The contain-

heal removal. A safety function shall bc to reduce rapidly, mcnL atmosphere cleanup systems shall be consistent with the functioning of other as. designed Lo permlL appropr'late periodic'In-I heat shall be pro- soclated systems, the containment pressure spection of Important components. such as

> function shall be and temperature following any toss.ofwool- filter frames. ducts, and piping to assure the

<ct decay heat and ant accident and maintain them at accept- Integrity and cspsblHty of the systems.

the reactor core at ably low levels. Crfferia>> 43-Tcrffnp of confafnmcnl at.

.ed acceptable fuel Sultab)e redundancy In components and cslgn conditions of mosphcrc cleanup sysfcms. The containment features. and suitable Interconnections, leak atmosphere cleanup systems shall be de-sure boundary are detection. Isolation, and containment caps. signed to permit appropriate periodic pres-bllltles shall be provided to assure that for sure and functional testing to assure (1) thc n nents and onslte electric power system operation (as- structural and leaktlght Integrity of Its corn.

>cree. cctlons, leak suming offsite power Is not available) and

>apabllltlcs shall be ponents, (2) the operability and perform.

for offslte electric power system operation ance of the active components of the sys.

for onslte electric (assuming onsite power Is not available) the terns such as Isns, filters, dsmpers. pumps,

> (assuming offslte system safety function can be accomplished, for offslte elec- and valves and (3) the operability of the sys-u>d assuming a single failure. tems as a whole and, under conditions ss

>eration (assuming Criferion 39-lnspcctfo>> of conlainmcnl

>liable) thc system heal removal sysfcm. The contalnmcnt heat close to design as practical. thc performance xompilshed, assurn- removal system shall be designed to permit of the full operational sequence that brings appropriate periodic Inspection of impor- thc systems into operation, Including oper-

>cy core coolfnp. A tant components, such as the torus, sumps. ation of applicable portions of the protec.

snt emergency core spray nozzles, and piping to assure the In- Lion system, the transfer between normal L The system safeLy tegrity and capability of the system. and ernergcncy power sources, and the oper-

>sfer heat from the Criterion 40-Tcsfinp of confafnmcn( heal ation of associated systems.

any loss of reactor removal system. The containment heaC re- Cr(ferion 44-Coohnp u>afer. A system to

>ai, (1) fuel and clad moval system shall be designed to permit transfer heat from structures, systems, and fere wlCh continued appropriate periodic pressure and function- components important to safety, to an ulti-

> prevented and (2) al testing to assure (1) the structural and mate heat sink shall be provided. The n Is limited to negll- leaktlght Integrity of Its components. (2) system safety function shall be to trsnsler the opcrablllty and performance of the Che combined hest load of these structures, In components and active components of the system. and (3) systems, and components under normal op.

>terconnectlons. leak the operability of the system as a whole, crating and accident conditions.

I containment capa- and under conditions as close to the design Suitable redundancy In components and d to assure that for as practica) the performance of the full features, and suitable Interconnections, leak rstem operation (as- operational sequence that brings thc system detection, and Isolation capabilities shall be

> not available) snd Into operaLlon, Including operation of sppll- provided to assure that for onslte electric er system operation csblc portions of the protection system, the power system operation (assuming offslte Is not available) the transfer between normal and emergency power Is noL available) and for offslte elec-can be accompltshed. power sources. and Che operation of the as- tric power system operation (assuming e sociated cooling water system. onslte power Is nol, available) the system

lion of cmcrpency Criteria>> 41-Confainment almosphcrc safety function can be accomplished, sssum-fhe emergency core cleanup. Systems to control fission prod- Ing a single failure.

54'7

rr LIHITING,CONDITION FOR OPERATIOH SURVEILLAHCE AEJUIR)'.HLHT 3.1.8 Illa) PRESSuRE COOLANT IHJECTION 4.1.8 lllGII PRESSURE COOLANT INJECTION AJJ11 Applies to the operational status of the Applies to the periodic testi>>g requirum<.nts high pressure coolant injection syste<n. for the high pressure coolant i>>juctiu>> cyst<.<ll.

O~b<.c ti ve: ~Ob ective:

To assure the capability of the high To verify the operability of the high )iressuru pressure coolant injection system to coolant injection system.

cool reactor fuel in the event of a loss-of-coolant accident.

h 111<<1 ) ~Sect ficat<on:

a. During the power operating con-" The high pressure coolant i>>jectio>> sur-dition, whenever the reactor veillance shall be performed as indicated coolant pressure is greater than below:

110 psig and the reactor coolant temperature greater than saturation a. At 1 eas t &nay el8 er'at)'hg'cycle.

temperature, the high pressure coolant injection systen> shall be Auto<natlc start-up of the lii<)h p) essur< ~

operable except as specified in coolant injection system shall be demon-Speci fi cation "b" below. s trated.

b. If a redundant component of the hinh pressure coolant injection system becomes inoperable the high pressure Pump operabili ty shall be deter<)i>>ud, 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.

71

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

If Specification "a" and "b" are not c. Surveillance wi th I~no crab le Coiigionvnt met, a normal orderly shutdown shall be initiated w>thin one hour and reactor coolant pressure and temp-When a component becomes I>>operable its redundant component shall hu demonstrated erature shall be reduced to less than to be operable iiuiedlately and daily 110 psig and saturation temperature thereafter.

within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

'12

ll BASES FOR 3.1.8 AND 4.1.8 HIGH'PRESSURE COOLANT IHJEL))OM Injection is. provided to ensure adequate 'core cooling in the unlikely event of rea k ..

a High Pressure Coolant System (KPCI) xl) reactor coo 1 an t li"

'ne ' Th e HPCI System.is required. for line breaks wh')ch exceed the capab) lity of the ntrol Rod Drive pumps and which are not large enough to allo~ 'fast enough depressur)zat on for core s p ra y to be feet ive.

e se t o f hi g h press ure r 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 reactor r c 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.

T.he motor driven feeduater pump shall be capable of automatic initiation upon receipt of either an automatic g turbine trip signal or reactor low-water-level signal.

.'he e 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.

he capa bility oof 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)11 w)l bee p laced in service 1

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.

October 1, 1986 73 Revised

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 Rev. 7 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.

3.0 Desi n Evaluation During a loss-of-'coolant accident within the drywell, Rev. 7 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.

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 Rev. 7 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

0 r

VI I-62 a use of a Bennetts Bridge hydro generator, while not equivalent to an on-site emergency power source, ev. 7 provides a highly reliable alternate off-site power supply for the HPCI function of the feedwater system.

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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