Text

Suppl to Petition to Intervene,Containing Final Contentions. Contentions Include Concerns Re Substd Containment Isolation Sys & Reactor Coolant Pressure Bouncary.Certificate of Svc Encl
	ML19260A226
Person / Time
Site:	Three Mile Island
Issue date:	10/22/1979
From:	Sholly S; AFFILIATION NOT ASSIGNED
To:	; Atomic Safety and Licensing Board Panel
References
	NUDOCS 7911080197
	Download: ML19260A226 (48)
	v • d • e

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, FRC PUBLIC DOCU.'dEWI Rgpog UNITED STATES OF A> ERICA NUCLEAR REGULATORY COMMISSION In the Matter of MEIROPOLITAN EDISON COMPANY) Dock No 5 -289 (Three Mile Island Nuclear ))

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Station, Unit No. 1)

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N $ $7NI SUPPLDIENT TO PEIITION To INIERVENE CONTAINING FINAL CONIENIIONS AND BASES SET O [!.g [

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FORTH WITH SPECIFICITY, g j STEVEN C. SHOLLY, PETITICNER w Parsuant to the Board's Memorandum and Order dated 21 September 1979 and the requirements of 10 CFR 2.714, Steven C. Sholly, Petitioner, files this supplement to the petition to intervene in the above-captioned proceeding.

Said supplement contains Petitioner's contentions in final form together with bases for each contention set forth with reasonable specificity.

Due to the unavailability of the Licensee's completed response to the NRC Staff recommended requirements for restart of Unit 1 and due t:o the uncertain nature of certain NRC rules and regulations which are subject to revision, some of the Petitioner's contentions as set forth herein are not as specific as would be possible if the aforesaid information were available before the due date for final contentions in this prcceeding. Petitioner requests leniency from the Board in this respect as these contentions were prepared without the aid of legal counsel. Petitioner will make 1296 086 1911080 h [

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every effort as information becomes available to provide more spectflctty with regards to contenttons and their bases.

Pet:ttoner potnts out to the Board that a number of very important and significant investigations into the Unit 2 accident (3/28/79) are not yet completed, and that these investigations potentially have great bearing on this proceeding because the Ltcensee is the same in both cases and there is a high degree of similartty between the design and construction of Units 1 and 2 at Three Mile Island.

Pending the availabiltty of reports of said investigations and the passage of a reasonable amount of time for the Petitioner to reviewsaid reports, Petitioner will seek permission from the Board to amend the petition to intervene pursuant to 10 CFR 2.714 (a)(3), should any of these reports contain information which provides a bssis for a contention regarding the protection of public health and safety in this proceeding.

Definitions Throughout these final contentions the following definitions shall apply

a. "Ltcensee" refers to Metropolitan Edison Company

b. " Unit 1" refers to the Three Mlle Island Nuclear Station, Unit No. 1, while " Unit 2" refers to Unit No. 2 of that same factitty;

c. "the 3/28/79 accident" refers to the loss of feedwater transient / loss of coolant acctdent which began at Unit 2 on March 28, 1979:

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d. "NEPA" refers to the National Environmental Polley Acts

e. "GDC" refers to the spectfled General Design Criteria contained in Appendix A of 10 CFR 50:

f. "NRC" or "the Commission" refers to the U.S. Nuclear Regulatory Commissions

g. "NUREG" refers to the standard code used by the NRC to number certain publications, as for example, NUREG-0600.

Final Contentions Contention # 1 It is contended that the Unit 1 containment isolation system does not meet the following requirements:

a. Conformance with the Standard Review Plan Section 6.2.4, " Containment Isolation System" :

b. Compliance with GDC 16, Containment Design:

c. Compliance with GDC 50, Containment Design Basts:

d. Compliance with GDC 54, PLptng Systems Penetrattng Containment.

It is further contended that as a result of the design and construction of the Unit I containment and the containment isolation system Unit 1 ts rendered incapable of compitance with 10 CFR 20.105,10 CFR 20.106, and Appendix I of 10 CFR 50, and that, therefore, there exists reasonable doubt that Unit 1 can te operated without endangering the health and safety of the public. Inasmuch as the Commission has the authority pursuant to 10 CFR 50.109 to require backfttting when such backfitting is required to provide substantial, additional protection of public health and safety, it is contended that compilance with SRP Section

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6.2.4, GDC 16, GDC 50, GDC 54,10 CFR 20.105,10 CFR 20.106, and Appendix I of 10 CFR 50 is required to protect the public health and safety, and that therefore backfitting of the Untt I containment and containment isolatton system is a necessary preconditton to permission to restart.

Basis for Contention # 1 The General Design Criteria of 10 CFR 50 establish design, fabrication, construction, testing, and performance requirements for components, structures, and systems in a nuclear power plant which exist to provide reasonable assurance that the plant can operate without undue risk to the health and safety of the pubite. Failure of a particular plant to meet any of the General Design Criteria is a very serious matter requiring corrective action to provide reasonable assurance that the plant can be operated without endangering the health and safety of the public.

The 3/28/79 accident at Unit 2 revealed a number of -

shortcomings of the contatnment and containment isola-ion system at Unit 2 and, because of similarttles in design and construction, also at Unit 1. The failure of the containment to isolate on diverse signals, including high radiation in the containment and initiation of high-pressure injection from the emergency core cooling system, led to the release of millions of curies of radioactive noble gases into the environment, resulting in offstte doses i296 089

accordit -

o NUREG-0600 at pages II-3-16 through 18 and pages 11-3-91 en which were in excess of radtation protection standards ccntained in 20 CFR 20.105 and 20 CFR 20.106. These exposures which are cited in NUREG-0600 are also apparently in violation of the "as low as is reasonably achievable" requirements of 10 CFR 50, Appendix I. The radiation releases cited herein were the direct result of the failure of the containment to isolate in a timely fashion following inttlation of high-pressure injection. A large quantity of highly radioactive water was pumped from the containment sump to the auxiliary building from which the radiation was subsequently released. This effectively resulted in the defeat of the containment as a safety system which is responsible for establishing a leak-tight barrter against the release of radioactivity to the environment in an uncontrolled manner.

The fact that such a large quantity of radiation was released into the environment as a direct result of the failure of the containment to isolate in a timely manner ts a violation of GDC 16 which requires that the containment establish the leak-ttght barrter to the uncontrolled release of radioactivity into the environment which is referred to above. GDC 16 further requires that the containment design conditions not te exceeded for as long as postulated accident conditions require. The same 1296 090

conditions which exist at Unit 2 with respect to GDC 16 also exist at Unit 1.

GDC 50 requires that the contatnment structure, including penetrations, be so designed so that the containment structure can accomodate the calculated pressure and temperature con-dLttons resulting from any loss-of-coolant accident without exceeding the destgn leakage rate. Design to accomodate the pressure and temperature conditions must be with " sufficient margin. " This margin, according to GDC 50, shall reflect ,

among other ttems, expertence available for defining accident phenomena and containment response. Considerable expertence in this area is now available as a result of the 3/28/79 accident at Unit 2 and must, under the provisions of GDC 50, be taken into account in evaluating the adequacy of design of containments. The radiation which was released to the envircnment during the Unit 2 accident is evidence that the design leakage rate was in fact exceeded, and that therefore the containment does not meet GDC 50 under loss-of-coolant accident (LCCA) condittons. By virtue of design and construc-tion similarities with Unit 2, Unit 1 also suffers from this non-compliance with GDC 50.

GDC 54 provides that piptng systems, including the containment sump piping, which penetra te the containment be provided with leak detection, isolation, and containment capabilities having redundancy, reliablltty, and performance b 09l

capabilities reflecting the importance to safety of tsola: Lng these piping systems. The containment sump piping, by virtue of its functLon in conveying water from the contatnment sump ptt to the auxiliary building, should be provided with the leak detection, isolation, and contatnment capablltties discussed in GDC 54. This is because in the event of an LOCA, radioacttvely contaminated watet wculd be expected to collect in the containment sump as it is vented frcm the primary reactor cooling system. It is clear from the results of the 3/28/79 accident at Unit 2 that the containment sump piping is not properly provided as per GDC 54 with necessary leak detection, tsolation, and containment capabt1Lttes , because it was through this pipe system that the radioactive water from the containment was sent to the auxillary butiding from which the radioactivity subsequently escaped into the environment. By virtue of the design and construction similarittes of Unit 2 to Unit 1, the same noncompliance with GDC 54 exists with Unit 1 as exists with Unit 2.

The radiation exposure and doses resultant from the releases from Unit 2 during the course of the 3/28/79 accident were specifically described in NUREG .00. At page II-3-18, NUREG-0600 cites releases from Unit 2 of noble gases which were at least eleven times the concentrations permited under the limits in 10 CFR 20.106.

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At pages II-3-91 and II-3-92, NURE3-0600 cites exposure and dose rates on Kohr Island for the perted from 0700 hours0.0081 days 0.194 hours 0.00116 weeks 2.6635e-4 months on 3/28/79 to 1200 hours0.0139 days 0.333 hours 0.00198 weeks 4.566e-4 months on 3/29/79 of 30 mrem /hr and 14 mrem /hr. These exposures are in excess of the 2 mrem /hr limit contained in 10 CFR 20.105. The exposures are also in excess of the Appendix I, 10 CFR 50, as low as is reason-ably achievable ILmits wb' c" limits the annual atr dose due to gamna radiation at any location near ground level which could be occupied by individuals at or beycnd the boundary of the site to ten milltrads. This limte was eastly exceeded by the 30 mR/hr and 14 mR/hr reading s taken on Kohr Island as etted in NUREG-0600 at pages II-3-91 and II-3-92.

Since the Commission has authority pursuant to 10 CFR 50.109 to require backfttting, and since violations of General Design Criteria directly relates to the ability of the plant to operate without endangering the health and safety of the pubitc, the Petitioner seeks to require that backfitting of the Unit I containment and containment isolation system te made a precondition to restart.

Contention # 2 It ts contended that the reactor coolant pressure boundary for Unit 1, as designed and constructed, does not meet the requirements of GDC 14, GDC 15, and GDC 30. It is further contended titat because of noncompliance with GDC 14, GDC 15, and GDC 30, the operatton of Unit 1 poses an 1296 093

undue risk to the health and safety of the public. Inasmuch as the Commission is empowered pursuant to 10 CFR 50.109 to require backfitting of Unit 1 to provide compliance with GDC 14, GDC 15, and GDC 30, and because compliance with the General Design Criteria is required to provide assurance that Unit 1 can operate without endangering the public health and safety, it is contended that the reactor coolant pressure boundary of Unit 1 must be backfitted to provide compilance with GDC 14, GDC 15, and GDC 30 prior to restart.

Basis for Contention # 2 The General Design Criteria of 10 CFK 50, Appendix A, establish design, fabrication, construction, testing, and performance requirements for components , structures , and systems in nuclear power plants which exist to provide reasonable assurance that the plant can te operated without undue risk to the public health and safety. Failure of a plant to meet any of the General Design Criteria is a very serious matter requiring corrective action to provide the required reasonable assurance that the plant can be operated without endangering the public health and safety.

The 3/28/79 accident at Unit 2 revealed a number of shortcomings with regard to the reactor coolant pressure boundary of Unit 2, and, because of similar design and construction, also of Unit l's reactor coolant pressure boundary. GDC 14 requires that the reactor coolant pressure i296 094

boundary be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage.

The reactor coolant pressure boundaries for Units 1 and 2 contain power-operated relief valves (PORV's) in the pressurizer system. The failure of tne pressurizer system PORV to ressut properly was, according to NUREG-0578 at page A-6 and A-7, a significant contributor to the sequence of events during the 3/28/79 accident and resulted in a direct violation of reactor c~olant pressure boundary integrity. Further, NUREG-0578 at page A-7 lists a number of instcnces, including the failure of the pressurizer system PORV to reseat, in which relief and safety valves malfunctioned. NUREG-0578 comments at page A-7, "It is not clear whether these past instances of improper operation resulted from inadequate qualification of the valve or from a basic unreliability of the valve design." In either case, use of these valves in the reactor coolant pressure boundary at Unit 1 is a violation of GDC 14 regarding design and testing of the reactor coolant pressure boundary. Inasmuch as a short-term recommendation of the Lessons Learned Task Force was to have Licensees conduct testing to qualify valves, including PORV's , in the reactor coolant pressure boundary, it is obvious that such valves have not yet been appropriately tested under anticipated operating conditions, including design basis transients and accidents. This is a direct i296 095

violatton of the requirements of GDC 14.

GDC 15 requires that the reactor coolant pressure boundary systems shall be destgned with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any conditton of normal operation, includins anticipated operattonal occurrences.

During the first eighteen minutes of the 3/28/79 accident at Unit 2, the design pressure for the system was exceeded in that the reactor coolant system pressure dropped below the safety injection actuation set-potnt of 1640 psig to 1100 psig (see NUREG-0600 at page I-2-15). Since the design and constructLon of Units 1 and 2 are very similar, Unit I suffers from the same problems under the same conditions; therefore, GDC 15 ls violated because the design condition of the system relating to pressure is in fact exceeded durtng an anticipated operational occurrence (a loss of feedwater transtent with reactor trip).

GDC 30 requires that components which are part of the reactor coolant pressure boundary be designed, fabricated, erected, and tested to the highest quality standards practical.

It has already been noted above that NUREG-0578 at pages A-7 and A-8 *ecommends a Licensee program to qualify valves, including P0h"'s which are in the Unit i reactor ccolant pressure boundary, under expected operating conditions including design basis transtents and accidents. This obviously means that 1296 096

such testing has not been performed fcr the PORV's in Unit l's reactor coolant pressure boundary. This is a direct violation of GDC 30. Inasmuch as violations of the General Design Criterta relate directly to the ability of the plant (Unit 1) to operate without endangering the pubile health and safety, and since the Commission has authority in this situa". ion pursuant to 10 CFR 50.109 to require backfitting, Petitt ner seeks to require backfitting of the reactor coolant pressure boundary and/or appropriate qualification testing of components of that system as a precondition for restart.

Contentton # 3 It ts contended that as a result of Licensee's Operating Procedures, the emergency core cooling system can be defeated by operator actions during the course of a transient and/or accident at Unit 1, such defeat consisting of either throttling back the high-pressure inject _on pumps or tripping these pumps. It is further contended that under the conditions of a loss-of-feedwater transient / loss of coolant accident at Unit 1, defeat of the emergency core cooling system high-pressure injectLon system by pump throttling and/or pump trip results in signiftcant cladding metal-water reaction, causing the production of amounts of hydrogen gas in excess of the amounts required by NRC regulations to be considered Ln the design and accident analysis of nuclear power plants.

It is contended further that such production of hydrogen gas 1296 097

results in the high risk of breach of containment integrity due to the explostve ccmbustion of the hydrogen gas in the contatnment. Inasmuch as the emergency core cooling system ts an engineered safety feature which is relled upon to protect the public health and safety, end because proper operation of the emergency core coo 1 Lng system is required to provide reasonable assurance that Unit 1 can be operated without endangering the public health and safety, it is contended that the emergency core cooling system operating precedures must be modified in order to ensure compliance with the GDC 35 requirement of negligible clad metal-water reactLon following a loss-of-coolant accident (LCCA). It is further contended that the emergency core cooling system operating procedtres must be appropriately modified prior to restart in order to provide for protection of the public health and safety.

Basis for Contention # 3 The emergency core cooling system is required for Unit 1 pursuant to 10 CFR 50.46. The emergency core cooling system is required by GDC 35 to prevent significant clad metal-water reaction during an LCCA. NUREG-0578, at page A-22, makes the following statement regarding the production of hydrogen during the 3/28/79 accident at Unit 2, "This amount of hydrogen generation was well in excesc of the amount required by the Cor.mtssion regulations as a design basis for any type of 1296 098

of post-accident combustible gas control system." The large production of hydrogen gas was reduced in seriousness at the Untt 2 3/28/79 accident by combustion at approxtmately 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months and 50 minutes after the initiation of the accident, resulting in a containment pressure sptke of 28 pstg (see NUREG-Oo00, at page IA-87). This combustion must be considered to have been fortuitous because Lt was not inttlated on purpose by the ope-'ators of the plante Ahcent this " lucky" occurrence,

.tgnificant quantitles of hydrogen gas would have collected in the containment, thus posing the risk of breach of contatnment integrity due to hydrogen explosion. Considering the lack of a hydrogen recombiner system at Unit 1 in a stmilar situation (which could occur given the similartty in design and construction of Units 1 and 2), presence of a dangerous amount of hydrogen gas in the containment requires the venting of the containment by plant operators (see NUREG-0578, at pages A-21 through A-25). Under accident conditions where there is a large amount of radicactivity in the containment, venting the containment wculd release this radioactivity to the environment. Had this procedure been necessary in the Unit 2 accident, off-site radiation doses would have been far in excess of those permitted under 10 CFR Part 20.

NUREG-0600 in the Operational Sequence of Events for the 3/28/79 accident (Appendix I-A of NUREG-0600) itsts several instances in whtch plant operators, followLng operattonal procedures , either throttled or tripped the ht2h-pressure 1296 099

Lnjection system pumps such events occurred at the following times after the reactor trip at 04:00:37 on 3/28/79:

a. 3 minutes, 13 seconds, operator throttles MU-V16 valves to pump 1C;

b. 4 minutes, 38 seconds, operator fully closes MU-V16 valves and trips pump IC: pump 1A still running in throttled condition and ts the only high-pressure injection (HPI) pump running at this potnt;

c. 5 minutes, 15 seconds, operator throttles pump 1A to minimum:

d. 10 minutes, 24 seconds, operator stops, restarts , and stops pump 1A within four seconds:

e. 11 minutes , 43 seconds , operator restarts pump 1A;

f. 93 minutes, operator running HPI flow at 150-200 gpm/ loop (below maximum):

g. 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months , 20 minutes , operator manually starts pump 1C; pumps 1A and 1C now running with valves wide open; ,

h. 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months , 27 mtnutes, operator resets ECCS;

i. 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months , 37 minutes , operator trips pump 1C;

j. 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months , 56 minutes, pump 1C and containment tsolation automattcally initiated by high-pressure signal;

k. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , zero minutes, ECCS and isolation defeated by operator;

1. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , 17 minutes, operator secure pumps 1A and 1C, leaving to HPI pumps running ;

m. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , 19 minutes, ECCS and isolation initiated automatically, but pump 1A is in pull-to-lock position and does not start;

n. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , 19 minutes, 18 seconds, operator defeats ECCS and isolation;

o. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , 22 minutes, operator restarts pump 13;

p. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , 27 minutes, operator starts pump 1C; HPI at 250 gpm (below maximum) until 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months and 4 minutes;

q. 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months and 4 minutes, operator trips pump IC 1296 100

r. 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months , 50 minutes, 28-psig pressure spike causes ECCS and isolation initiated; pumps 13 and 1C now running (IA still locked out):

s. 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months , 51 minutes, operator trips pump IC.

According to NUREG-0600 at page I-2-21, failure of the operators to maintain HPI flow according to design and procedural requirements resulted in apparent serious core damage and onsite and offsite exposure to radioactive materials. Failure to follow procedure 2202-1.3 in this regard ts reported by NUREG-0600 at page I-2-21 to be under consideration as a potential item of noncompilance pursuant to Technical Specification 6.8.1.a. This situation demonstrates inadequate procedures on the part of the Licensee to ensure that operators' adhere to Technical Specificattons. This situation applies equally to Unit I due to design and construction similarteles and procedural similarities between Units 1 and 2.

Because this situation involves an engineered safety feature of Unit I which is required for the protection of the pubite health and satety, and because it is beyond the capabtlittes of predictive sciences to determine that the ECCS will not be required if procedural changes are postponed untti after restart, Petitioner contends that changes are necessary prior to restart to protect the public health and safety.

Contention # 4 It is contended that the ability of the Licensee to provide radiation exposure and dose data to responsible officials having decision-making responsibilities wLth respect

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to off-site radiation releases and emergency response to such releases is significantly impatred due to the lack on the part of the Licensee of on-stte environmental TLD processing facilities. It ts further contended that the Licensee is not prepared to implement Health Physics Procedure 1670.6, "Offstte Radiological Monitoring," due to the lack of on-site TLD processing capabilities. It is contended further that this lack of preparedness to implement Health Phystes Procedure 1670.6 does not adequately protect the public health and safety under conditions of off-stte release of radioactive materials, because this limits the Ltcensee's environmental TLD data to five off-stte stations beyond five miles from the site, conditions which can permit plumes to fall in between TLD sites , thus severely limiting the ability of the Ltcensee to provide accurate and timely radiation exposure and dose estimates to responsible public officials. It is further contended that protection of public health and safety under accidental release of radioactivity conditions requires that the Ltcensee have on-site TLD processing capability prior to restart.

Basts fcr Contention # 4 I se basis for this contention comes from NUREG-0600 at pages II-3-95 and 11-3-96 where Lt is revealed that the Licensee is dependent on off-stte processtng of TLD's by Teledyne Isotopes and Radiation Management Corporation.

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It ts further revealed in NUREG-0600 at page II-3-96 that, by the admisston of an employee of the Licensee, the Ltcensee is not prepared to implement Health Phystes Procedure 1670.6 due to the lack of on-site TLD processing capability. Since the winds in the area of the Unit I stte are variable in speed and direction, the existence of a well-defined plume in the area of an existing TLD site cannot be presumed.

Therefore, the inability of the Licensee to implement the referenced Health Physics procedure effectively precludes accurate and timely radiation exposure and dose data from being available for use by officials in determining the need for emergency protective actions in the event of a release of radioactive materials to the off-stte environment.

This is clearly not in the interests of public health and safety, and the Petttioner therefore contends that on-stte TLD processtng capability is necessary prior to restart.

Centention # 5 It is contended that the Ltcensee has not provided sufficient radiation monitoring capacity in the contatnment, spaces which could contain LCCA fluids , effluent discharge paths, and the plant environs, as required by GDC 64. It ts further contended that existing radiation monttoring in said locations does not provide for sufficient monitortng for radioacttvity which may be released due to anttclpated

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operattonal occurrences and due to postulated accidents ,

capacity for which ts required in GDC 64. It is contended that lack of compliance with GDC 64 will prevent the Ltcensee from maktng accurate estimates of radioactivity releases from Unit 1 under conditions of anticipated operational occurrences and postulated acctdents, and that this lack of compilance places the public health and safety at significant risk because such information is required by public officials to provide bases for decision-making related to emergency actions which may be required to protect the public health and safety. It is further contended that untti the Ltcensee provides sufficient numbers and distributton of radiation release monitors which are capable of yielding on-scale results for all conditions of radiation release resulting from anticipated cperational occurrences and postulated accidents, permission for the Ltcensee to restart Unit I should be denied.

Basis for Contention # 5 The basis for this comes from NUREG-0578 which discusses the subject of radiation monitoring capability. Under GEC 64, the Ltcensee is required to provide means for monitoring the reactor containment atmosphere, spaces containing components for rectrculation of loss-of-coolant accident fluids, effluent discharge paths, and the plant environs for radioac-tivity that may be released from normal operations, including i296 104

anttelpated operational c',currences and postulated accidents.

According to NUREG-0578 at page A-37, " A recent survey of existing gaseoui effluent monitoring capabilttles of operating plants shows that less than 20 percent of operating plants have monitors that would have stayed on scale under the conditions of the TMI accident. It can also be shown, however, that the potential releases from postulated accidents may be several orders of magnitude higher than was encountered at TMI. Undte such circumstances , none of the effluent monitors now in service at any operating plant would remain on scale." The clear tmpli-cation of these statements from NUREG-0578 is that gaseous effluent monitoring at Unit 1 ts incapable of on-scale readings during at least some postulated accident conditions. Since off-scale readings permit only lower-bound estimations to be made on the amount of radiation released, the inability of 3 Unit 1 effluent monitors to provide on-scale results under certain accident conditions introduces uncertainty into a j situation during which accurate information is a necessity, .A '

as was so vividly demonstrated during the 3/28/79 accident ,YF-at Unit 2. Accurate radiation release information is required ,

i l) for determinations of emergency actions by pubile officials .]k h, g charged with responsiblittles in this respect during accident ,

conditions. Lack of reltable and accurate information during

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such a critical time clearly is not in the interest of public health and safety, and where releases exceed the capabtlity of the monitor by orders of magnitude (as with the Unit 2 vent ;- I 1256 ios

monitor which, according to NUREG-0578 at page A-39 was off-scale at 0.5 CL/second, compared with estimated release rate calculated from NUREG-0600 at page II-3-17 for the 33-hour pertod beginning 0700 hours0.0081 days 0.194 hours 0.00116 weeks 2.6635e-4 months on 3/28, which was over 45 CL/second) the health and safety of the public can be seriously jeopardized by underestimation of radiation release rates.

Contention # 6 It is contended that Unit l's design contains the following features which make the plant unusually sensitive to off-normal transient conditions originating in the secondary systems

a. Design of the steam generators to operate with a relatively small itquid volume in the secondary sides

b. Lack of direct initiation of reactor trip upon the occurrence of off-normal conditions in the feedwater systems

c. Reliance on an integrated control system to automatically regulate feedwater flows

d. Actuation before reactor trip of a pilot-operated reitef valve on the primary system pressurtzer

e. Low steam generator elevation relative to the reactor vessel which provides a smaller driving head for natural circulation.

It is further contended that these features result in the placing of a greater burden on successful ECCS function and proper operator decisions and actions than was anticipated in design to deal with off-normal system behavior during anttcipated transient conditions. It is contended that the total effect of these factors results in the lack of reasonable assurance that Unit I can be operated without endangering the pubile health and safety, and that until the design and operating procedures at Unit 1 are acdified to provide such reasonable assurance, permission for restart must be denied.

9 06

It is further contended that the short-term actions identified in the Commissions Order and Notice of Hearing dated 9 August 1979 are insufficient to provide the requisite reasonable assurance of operation without endangering the public health and safety.

Basis for Contentton # 6 There are a number of sources for the basis of this contention. The five features of the Unit I reactor wbtch make it unusually sens ttive to off-normal transient conditions originating in the secondary system were tdentifled as factually existing in the Commisston's Order and Notice of Hearing dated 9 August 1979 at page 3. The contention of the Petitioner that these features place a greater burden on ECCS and operator performance during such transients also comes from the same Order at page 3. Inasmuch as the five identifled features relate to compliance with GDC 10, 14, 15, 30, 34, and 35 and since compliance with the General Design Criteria of 10 CFR 50, AppendLx A, ts required to provide reasonable assurance that Unit I can be operated without undue risk to the public health and safety, it is inconsistent for the Commisston to permit restart untti all Ltems relating to compliance with the General Design Criteria are complete.

The Commission proposes to permit restart (as per Order and Nottce of Hearing dated 9 August 1979 at pages 7 and 3) before the Licensee is to submit a failure mode and effects analysis of the Integrated Control System. The ICS ts used 1296 107

to regulate feedwater flow , which relates to compliance with GDC 34. If the Commisston is unsure Lf the ICS will function prcperly, Lt would appear to be inviting non-compliance with Lts own regulations by permitting restart while compliance with GDC 34 is not assured.

Similarly, the Commission proposes to permit restart before the Licensee is required to comply with Category "B" recommendattons of the NUREG-0578 report. These recommendations include the following Ltems which relate to compilance with the General Design Criteria and/or to the five features identified as contributing to the Unit's sensttivity to off-normal transtents

a. Completion of instrumentation installation for detection of inadequate core cooling (relating to GDC 13, 15, and 10):

b. Ccmpletion of installation of hydrogen gas control penetrations of the containment (relating to GDC 16, 41, 50, 54, and 56):

c. Review of basis for recombiner use (relating to GDC 41) ;

d. Completton of installation of high-range effluent monitor system (relating to GDC 64).

Again, the Commission appears to be inviting non-compliance by proposing to permit restart prior to completion of these Leems . This is inconststent with the requistte finding of " reasonable assurance" prior to restart.

Contentton # 7 It ts contended that the Licensee's analysis of loss-of-coolant (LCCA) accidents is imcomplete, that the Licen ee 1296 108

is therefore not in compliance with the requirement of 10 CFR 50, Appendix K, requiring analysis of a spectrum of postulated LOCA's , and that to the extent such analyses are incomplete, operational procedures used by the Licensee to define operator action with respect to LOCA's are defective. It is further contended that without complete analysis of LOCA's there is a lack of reasonable assurance that Unit 1 can be operated without endangering the public health and safety. It is there-fore contended that until LOCA analyses are complete and operational procedures relative to operator actions in dealing with LOCA's are revised and accurate, permission for restart must be denied to protect the public health and safety.

Basis for Contention # 7 The basis for this contention comes from a statement in the Commission's Order and Notice of Hearing dated 9 August 1979 at page 5, which states as a short-term action that the Licensee shall, " Complete analyses for potential small breaks and implement operating instructions to define operator action." This clearly implies that the LOCA analysis for Unit 1 are imcomplete, which appears to violate 10 CFR 50, Appendix K requirements . It also leaves open the possibility that a thus-far unanalyzed LCCA could occur for which there are no operating procedures. This is clearly noc in the interests of public health and safety.

There are many operational procedures used by the Licensee which could define operator actions with respect to LCCA's which could be defective as a result of incomplete LCCA analysis. .It is

~

1296 109-

difficult to specify these procedures without benefit of discovery to obtain coptes of the Licensee's operating procedures, but a preltminary review of the " Report in Response to NRC Staff Recommended Requirements for Restart of Three Mlle Island Nuclear Station, Unit 1" gives the following as a list of probable defective operational procedures:

a. Group 2 Procedures 1303-11.8 (High and Low Pressure Injection), 1303-11.9 (R.B. Emergency Cooling System), 1303-8.1 (Reactor Coolant System Test), 1302-5.10 (R.B. 4 PSIG Channels),

1302-5.11 (R.B. 30 PSIG Channels), 1303-5.4 (Emergency Feedwater Pumps), 1303-1.2 (R.C.

Flow Surveillance), 1203-28 (Post Accident H, Purge), 1104-55 (Reactor Bldg. Atmosphere Cteanup System), 1101-1 (Plant Ltmits and Precauttons),1101-2 (Plant Setpc'_ats),

1102-14 (Reactor 1104-1 (Core FloodBuilding)

System Purging and Venting),

, and 1104-5 (Reactor Building Spray System):

b. Group 1 Procedures AB 1203-15 (Loss of Reactor Coolant Makeup), EP 1202-37 (Cooldown from Out-side Control Room), EP 1202-26A (Loss of Feedwater to EPOTSG),

1202-12OP 1102-4 (Power (Excessive Operation Radiation Levels)),

EP 1202-35 (Loss of Decav Heat Removal),

EP 1202-14 (Loss of Flcw), EP 1202-6 (Loss of RC Coolant), SP 1303-5.4 (EFW Pumps),

AP 1004 (Emergency Plan and Procedures), and 1106-3 (Feedwater System).

This Itst is not meant to be definitive but is points out procedures which might require revision upon completton of LOCA analysis in order to adequately protect public health and safety. Failure of the Commission to require reviston of appropriate procedures with respect to LOCA's risks tragic consequences, and therefore Petittoner contends that the procedures must be revised prior to restart.

1296 110

Contentton # 8 It is contended that the Ltcensee's emergency plan for Unit 1 ts defective because Lt does not provide sufficiently for the protection of pubite health and safety under all con-dittons of operation, including anticipated operattonal occurrences and postulated accidents. It ts further contended that the proposal by the Commisston to extend the Licensee's emergency action capabllttles out to a distance of ten' miles from the site does not adequately address the protectLon of pubite health and safety under accident conditions, especially when highly variable factors such as changing weather conditions, strikes affecting transportation factilties , the presence of significant numbers of extra persons in the area during certain time of year, and the nature of the geography and transportation routes in the surrounding area are considered. It ts contended that the Licensee's emergency plan does not contain sufficient detati to permit the finding of reasonable assurance that accrocriate measures gan and gill be taken in the event of an emergency to protect the public health and safety and prevent damage to property, as required by Appendix E of 10 CFR Part 50.

It ts further contended that the Ltcensee's emergency plan does not provide for means for determining the magnitude of the release of radioactive materials, including sufficient criteria for determining the need for nottftcation of local, state and federal agencies, and criterta for determtning when protective measures should be considered within and outside the sLte 1296 11i

.w .

boundary to protect public health and safety and prevent damage to property, which are also required under Appendix E of 10 CFR 50.

It is also contended that the emergency procedures of the Licensee do not assure prompt notiftcation of the public and local offlctals of the need fnr pubite evacuatton or other protective measures which may be necessary in the event of an emergency at Unit 1, assurance of which is requtred under Appendix E of 10 CFR 50. It ts contended that until the Licensee's emergency plans are suitably revised and sufficiently detailed to assure that appropriate measures can and will be taken in the event of an emergency to protect the pubite health and safety and prevent damage to property, permission for restart must be cenied.

Basts for Contentton # 8 There are several aspects to this contention dealtnB with the adequacy of emergency planning by the Licensee. The Commission proposed that the Licensee be required to extend the capability to take appropriate emergency actions for the population around the site to a distance of ten miles (Order and Nottce of Hearing, 9 August 1979, at page 8). The Commission placed this requirement in the context of permitting a f tnding of reasonable assurance of safety in the long-term operation of Unit 1. Inasmuch as there is no assurance that the emergency plan would not be needed in the short-term, and since the Commission obviously ftncs the Licensee's plan lacking Ln certain respects (see Cr, =r and Notice of Hearing, 9 August 1979, at page 6 and 8), for the Commission to permit 1296 ii2

restart in the absence of workable and sufficient emergency plan would be an act of irresponstbtlity and would appear to violate the Commission's own regulations (Appendix E of 10 CFR 50). It is not acceptable for the Ltcensee to be permitted to restart Unit I without a workable and sufficient emergency plan. Further, the Petitioner finds numerous additional problems with the Licensee's emergency plan in addttion to those addressed in the Commission's Order.

Apperdix E of 10 CFR 50 requires demonstration of reasonable assurance that the Licensee's plans provide that the Licensee not only can but in fact g(ll take appropriate emergency actions to protect pubile health and safety and prevent property damage in the event of an emergency at Unit 1. The performance of the Licensee during the 3/28/79 accident at Unit 2 certainly throws considerable doubt on whether the Lleensee Elll in fact act by taking the necessary emergency steps required. In the light of the Licensee's actions during the Unit 2 accident, the Petitioner believes that it should be imcumbent upon the Licensee to demonstrate beyond reasonable doubt that Lt not only has the capability to take emergency actions when they are needed, but that it also will in fact take those actions in a timely manner. In the Unit 2 incident, the incident began at 0400 hours0.00463 days 0.111 hours 6.613757e-4 weeks 1.522e-4 months . By 0600, the Ltcensee clearly new the reactor trip was not normal (NUREG-0600, page IA-38 and 39). Both alarm printer and utility printer were out of service from 0513 to at least 0648--no alarm data was printed during this time period. At 0656, a site emergency was declared. The i296 113

Pennsylvania Emergency Management Agency was not notified for six additional minutes (until 0702). By this time the auxiliary building had already been evacuated and the containment dome monitor was reading 600 R/hr. A general emergency was not called until 0724. The fact that the lead state emergency agency was not contacted about the incident until three hours after its start is evidence that the Licensee was not at that time prepared to prcmptly notify local, state, and federal agencies in time of emergency. A minimal step would have been to put the state on alert when the first radiation monitors began registering alarm set-points or when it was realized that the trip had not been normal, yet the Licensee did nothing until there had been a formal declaration of site emergency.

NUREG-0600 at page II-2-7 evaluates the emergency conditions relative to the Licensee's existing plan at the time of the 3/28/79 accident at Unit 2. The investigators from the Office of Inspection and Enforcement found that a site emergency should have been called at 0430 hours0.00498 days 0.119 hours 7.109788e-4 weeks 1.63615e-4 months on 3/28 rather than at 0656 because a condition of site emergency involving loss of primary coolant pressure coincident with either high reactor building pressure or high reactor building sump level was met by 0430. Upon a projection of a 40 R/hr dose rate for Goldsboro, a general emergency should have been called at 0710; instead, the calculated dese was discounted even though there was not actual data available for evaluating the correctness of the calculated radiation exposure rate. The Petitioner, as a result of these factors and a well-documented i296 ii4

lack of forthrightness in releasing information in the early days of the 3/28/79 accident on the part of the Licensee, believes that there is ample evidence to suggest that the Licensee will not act to take appropriate emergency measures on a timely basis, and that therefore the Licensee should be required to demonstrate that this attitude will not continue in the future should Unit 1 be permitted to restart.

The Licensee is required by various Technical Spect-fications, NRC regulations, and Operating Procedures to make numerous radiation surveys in the event of an accidental release of radiation. Despite the fact that a site emergency was declared at 0654 and a general emergency was declared at 0724 on 3/28/79 during the Unit 2 accident, the first environ-mental radiation survey was not performed until 0748 (see NUREG-0600 at page II-3-77). In addition, during the period from 0700 on 3/28 and 1600 on 3/29, when 6.6 milllon curies of noble gases were released from the plant, there were two time periods of more than five hours and two hours during which stable plume conditions were known to have existed and during which the Licensee failed to conduct radiation surveys. A third survey made in Goldsboro revealed radiatton levels of 20 mR/hr gamma and 30 mR/hr beta / gamma, in excess of permitted levels. It is clear that the Licensee does not provide the means for determining the magnitude of a radiation release in the emergency plan, means for which are required under Appendix E of 10 CFR 50.

1296 i15

Emergency actions under the Licensee's plans are not divided into what can be done under differing weather conditions.

Heavy snowfalls, fog, heavy rain, and temperature inversions are all meteorological phenomena which can and do occur near the site of Unit 1, and all of these can drastically impact on the success of emergency actions and the time frame within which such emergency actions must be executed. During certain times of the year (for instance, during Farm Shcw week in Harrisburg), there are many more persons in the area than would normally be the case; this is not addressed sufficiently in the Licensee's emergency plans. There are drastic differences in terrain and the availability of transportation routes out of the area depending upon which compass direction a radiation plume may be headed; this can impact significantly on decisions regarding emergency actions and the time frame within which such actions must be taken, yet this is not addressed in the Licensee's emergency plans.

Licensee's emergency plans are seriously defective, requiring far more in the way of revision than the largely cosmetic revisions suggested as sufficient by the Commission.

Such revisions must be accomplished prior to restart to protect the public health and safety.

Contention # 9 It is contended that the Licensee's environmental radiation monitoring program contains an insufficient number 1296 116

of monitoring sites and an inadequate distribution of monitoring sites within twenty miles of the Unit 1 site to provide sufficient protection of the public health and safety. It is further contended that there is in the Licensee's environmental radiation monitoring program an unwarranted reliance on the use of thermoluminescent dosimeters (TLD's) for providing information used to calculate radiation exposure data. and that this unwarranted reliance on TLD's seriously underestimates radiation doses to the public.

It is also contended that the Licensee does not possess adequate portable radiation monitors to provide additional information in the event of an offsite radiation release, and that the Licensee does not exercise adequate administrative control over the maintenance of these units , nor the training of personnel in their use. It is contended that the radiation monitoring program of the Licensee must be greatly upgraded prior to restart to ensure adequate protection of the public health and safety .

Basis for Contention # 9 It is clear from correspondence between the Licensee and the Commission and from NUREG-0600 (at pages II-3-92 and II-3-95) that there are only five TLD locations outside a five-mile radius from the Unit I site and only 15 TLD locations within the five-mile radius at off-site locations. Plumes from off-site radiation releases could easily fall in between such a ilmited unmber of radiation monitoring sites. In fact, i296 i17

during the first 68 hours7.87037e-4 days 0.0189 hours 1.124339e-4 weeks 2.5874e-5 months of the 3/28/79 accident at Unit 2, when most of the radioactivity was released, the wind blew in a given sector for several hours at a time only 30% of that time period, during the major part of the 68-hours period, the plume was either not well defined or tended to meander, causing actual exposure rates at a given point to fluctuate considerably (see NUREG-0600 at page II-3-95). Even though the Licensee's own procedure (Health Physics Procedure 1670.6) required the deployment of additional TLD's to assist in compiling accurate dose estimates, the Licensee failed to accomplish this because the Licensee was not able to implement the procedure (see NUREG-0600 at pages II-3-95 and 96).

There are other examples of inadequate management administrative control over radiation monitoring involved with the Unit 2 accident. NUREG-0600 at page II-3-97 lists two time periods totalling 12.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months (time periods when significant amounts of radiation were being released) when inadequate assessments of radiological conditicns were made.

In addition, NUREG-0600 goes on to say that the original radiation survey records were discarded, leav' ; only Emergency Control Station records from results which aere radioed in.

The ECS log does not contain complete records of the specific instruments used in the surveys , the mode in which the instruments were used, the _rientation of the instrument with respect to the source being measured, and the duration of the actual measurement.

1296 i18

In addition, NUREG-0600 lists in page II-F-4 as a potential item of noncompliance the fact that eight environmental air samplars used in the Licensee's radiation monitoring program had not been calibrated for a period beginning in 1974 up to March 1979, in violation of the Environmental Technical Specification 5.7 and Surveillance Procedure 1302-5.24.

In addition, actions of the plant personnel as evaluated by NUREG-0600 at pages II-3-68 and 69 revealed that the employees of the Licensee, including chemistry and radiation protection staffs, did not have adequate comprehension of:

a. The methods of evaluating internal and skin doses of radiations

b. The limitations of portable survey instruments, including range, energy response, and the significance of open and closed window readings;

c. The plant systems and components;

d. Concepts and techniques used to minimize radiation doses;

e. Regulations, license conditions, and plant procedures:

f. The emergency organization.

It is clear from a reading of NUREG-0600 that the Licensee did not exercise proper managment administrative controls over the radiation monitoring program, and therfore the Petitioner seeks to require proof of such proper controls prior to restart.

Contention # 10 It is contended that until a method for decontaminating i296 iI9

and restoring Unit 2 has received NRC approval and the environmental impact of that method and the impact of that method on the waste handling and storage capacity on Unit I have been evaluated, to proceed. with restart would place an unnecessary and unreasonable risk on the public health and safety. It is further contended that until the Licensee can provide reasonable assurance that it can safely operate Unit I while decontaminating and restoring Unit 2, restart must be postponed.

Basis for Contention # 10 This contention is based on GDC 5 which prohibits the sharing of structures , systems , and components important to safety among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions , including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units. The petitioner desires the Licensee to make a showing of insignificant impairment with tespect to Units 1 and 2, considering accidents at either unit.

Assuming that the Licensee plans to decontaminate Unit 2.

Petitioner feels that the Licensee must be able to show that an accident at Unit I will not so affect activities at Unit 2 as to pose a significant risk to pubile health and safety.

Restart of Unit I without knowing how the activities at Unit 2 will impact on Unit 1 places an unnecessary risk 1296 120

on the public health and safety because if Unit I waste handling and storage capacity is used to assist in the Unit 2 activities and an accident occurs at Unit I which requires the use of Unit l's waste handling and storage capabilities, it is unclear what would occur. It is possible that an uncontrolled release of Ilquid wastes would result in such a situation, and that could easily result in significant risk to public health and safety, especially via the drinking water pathway. It is therefore necessary that the impact of Unit 2 activities be known in advance to the extent possible prior to the restart of Unit 1.

Contention # 11 It is contended that the production of hydrogen in the reactor core from clad metal-water reactions following an LOCA poses an unacceptably high risk of catastrophic failure of the reactor pressure vessel and the reactor containment, with the subsequent release of a substantial portion of the core inventory into the environment. It is further contended that until a safe and reliable means for eliminating hydrogen gas from the containment is installed at Unit 1, and is provided with suitable redundancy as required by GDC 41, restart of Unit 1 poses a risk to public health and safety and must be denied.

Basis for Contention # 11 There was, according to NUREG-0578, the production of amounts of hydrogen in the 3/28/79 accident at Unit 2 which are above the amount required to be considered in design bases under Commission regulations. A 28-psig pressure spike in the 1296 121

containment at Unit 2 during J,e accident caused by hydrogen combustion. According to NURB3-0578 at page A-21, had there not been provisions for recombiners at Unit 2, it would have been conceivable that containment atmosphere venting would have been required to control hydrogen gas concentrations.

Such a venting, in view of the grave amounts of radiation in the containment, would have resulted in off-site doses far in excess of anything permitted under NRC regulations.

It is clear that in the case of Unit 1, which currently has not hydrogen recombiner capability, venting of containment atmosphere following an LOCA of similar magnitude to the accident at Unit 2 would also result in greatly excessive radiation doses at off-site locations. Due to the large population at risk and the large amount of radioactivity involved in such an instance, venting is clearly not in the interests of public health and safety. In view of the fact, according to NUREG-0578 at page A-22, that hydrogen recombiner systems are not very costly, and pursuant to backfitting authority of the Commission under 10 CFR 50.109, a requirement for installation of a hydrogen gas control system at Unit 1 prior to restart would be a prudent measure which is well in line with ests;'ished policy regarding defense-in-depth and protection of pubile health and safety.

Collection of high concentrations of hydrogen gas in the containment at Unit 1 following a postulate LCCA under current conditions would present the choice between venting i296 122

a large amount of radiation to the environment or risking the chance of catastrophic pressure vessel / containment failure in the event of a massive hydrogen explosion. Providing hydrogen control systems 't Unit 1 will eliminate this choice.

Contention # 12 It is contended that regardless of the substance of the Final Order in this proceeding, the decision of the Board and the Commission will, in view of the extraordinary circumstances of this proceeding, constitute a major federal action which could significantly affect the quality of the human environment, and that as such, the decision of the Board and the Commission in this case falls under the requirements of Section 10Z of NEPA.

It is further contended that the environmental impact statement required under Section 102 of NEPA must be issued before the NRC's decision can take effect in this case. It is further contended that as a result of the similarities in design and construction of Units 1 and 2, the results of the 3/28/79 accident at Unit 2 have rendered invalid major sections of the Final Environmental Impact Statement on Three Mile Island Unit 1 (NL1UDG-0552), and that such sectione as have bca rendered invalid must be addressed in the new environmental impact statement which is required to be preoared. It is also contended that the psychological impact of Unit I restart must be evaluated in the NEPA statement required under Section 102 for this proceeding.

1296 123

Basis for Contention # 12 Petitioner has thus far been unable to find any other instance of a hearing conducted for the sole purpose of determining whether a Licensee should be permitted to restart a nuclear power plan: following the suspension of the Licensee's operating license. In view of the unprecedented nature of these proceedings , and in view of the unprecedented accident at Unit 2, a unit with similar design and construction and run by the same Licensee, Petitioner finds that an environmeatal imecct statement is clearly called for under Section 102 of NEPA which rer,uires such a statertant for any " major Federal action."

The accident at Unit 2 has substantially rendered invalid major portions of the FES for Units 1 and 2, NUREL-0552.

The sections rendered invalid are as follows:

a.Section III, D, 2, which covers radioactive wastes released during operation and the solid wastes expected to generated during operations;

b.Section V, which covers the environmental impact of plant operations, especially part D, covering the radiological impact of routine operations;

c.Section VI, which covers the environmental L1 pact of postulated accidents , especially as it discusses Class 9 accidents (Unit 2 accident on 3/28/79 was classified aa Class 9 by the Commission), and LCCA's;

d.Section XI, which covers the cost-benefit ratto, which should be recalculated to include the costs associated with the Unit 2 accident and cleanup and the costs of modifying Unit 1 for restart.

Evaluation of psychological impacts under NEPA is required, inasmuch as Paragraph 3 of Section 102 of NEPA requires all federal agencies to develop methods and procedures to insure the evaluation i296 i24

in decisionmaking of " presently unquantified environmental amenttles and values," along with economic and technical issues.

The lack of environmentally-caused stress is certainly an environmental amenity and is therefore subject to these provisions.

NEPA does not list the specific impacts to be constdered in the environmental impact review process but rather aims to include a wide variety of concerns which are not limited to simply economic and technical concerns. Indeed, Paragraph A

. of Section 102 of NEPA mandates the use of a " systematic, Enterdisciplinary approach which will insure the in".egrated use of the natural and social sciences" in decisionmaking which may have an impact on the environment. The reference to social sciences could reasonably include the evaluation of psychological stress caused by the proposed restart. Petitioner contends therefore that psychological stress associated with Unit I restart is subject to NEPA review, even Lf only in qualitattve form, and that such stress must be balanced agains" any pctential benefits of the proposed action.

Contention # 13 It is contended that the Unit 1 computer system does not meet t a requirements for instrumentation and control specified in GDC 13, and ts inadequate to insure proper operation of the Unit I reactor under all conditions of normal operatton, including anticipated operational occurrences and postulated accident conditions. It is further contended that the lack of real-time printcut capability during accident conditions and the lack of sufficient redundancy in the computer 1296 125

system place the public health and safety at significant risk during accident conditions, especially if computer function is lost and no backup unit is available. It is contended that untti the Unit I computer system is upgraded to meet the standards of GDC 13 and until suitable redundancy is provided within the computer system to assure real-time printout capability at all times , permission for restart must be denied on the basis of risk to public health and safety due to inadequate availability of operational information to Unit 1 operators.

Basts for Contention # 13 The requirements of GDC 13 for adequate instrumentation to provide for monitoring variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure safety, are certainly applicable to the Unit I computer system. The computer system is a crucial tool used by the plant operators to maintain control over the reactor and is used also to provide rapid availability of information for the operators. The records of the computer are also used in the analysis of tran!_ents and accidents by investigating teams within and outside the Licensee's organization and the NRC.

Computer generated records are typed on the Alarm Printer, the Utility Typer, and the Log Typer, and are taked on Analog Trend Recorders. For the period covered by the 3/28/79 Unit 2 1296 126

accident, several information gaps exist, partially because of equipment failures and partially because the records in question were either lost, discarded, or purposely destroyed.

Alarm Typer and Utility Typer our put were lost due to equipment problems and due to probable operator action to actuate the alarm suppress function at 0648 hours0.0075 days 0.18 hours 0.00107 weeks 2.46564e-4 months on the 28th to print out current data and wipe out historical information. Further problems with the Alarm Typer and Uttlity Typer were recorded during the period 1848 to 1910 hours0.0221 days 0.531 hours 0.00316 weeks 7.26755e-4 months when the typers jammed with paper. Investigators from the Office Of Inspection and Enforcement concluded that the information lost was not critical to understanding the accident, but the Petittoner feels that this concluston is at best speculative due to the unprecedented r.ature of the accident. Utility Typer output from0000to}324andfrom 2008 to 2012 was lost (3/23/79) and were not found by investigators. In addition, the strip chart from Analog Trend Recorder # 2 was never found for the day of the 28th of March. Pettrioner finds this lack of records appalling and believes it conclusively demonstrates the inadequacy of the Unit 2 computer system, and, due to similarities ,

also the Unit I computer system. Had redundancy been available, such computer generated records would not have been lost.

Printers were running behind real time early in the accident. According to NUREG-0600 at page I-4-79, by 0646 hours0.00748 days 0.179 hours 0.00107 weeks 2.45803e-4 months on the 25th of March, the Uttlity Typer was running one hour and 33 minutes behind real timer by 1313 hours0.0152 days 0.365 hours 0.00217 weeks 4.995965e-4 months , the i296 127

Alarm Typer was running two hours and 39 minutes behind (had the printer memory not been wiped out at 0648 hours0.0075 days 0.18 hours 0.00107 weeks 2.46564e-4 months , the lag would have been over four hours because the lag is cumulative). Lack of real-time data presented problems to operators in analyzing what was happending during the course of the Unit 2 accident. Such conditions would atso cause problems with the Unit 1 operation. Lack of available information could easily have caused operator error, and thus risked the public health and safety.

Contention # 14 It is contended that the Licensee has negligently violated NRC regulations and technical specifications and that such violations place the safety of the pubile and the protection of the public health in question. It is contended further that the performance of the Licensee during the Unit 2 accident in terms of violations of regulations and technical specifications, and in terms of timely execution of safety-related functions , is directly applicable to Unit 1 since such violations call into question the management and administrative capabilities of the Licensee. It is further contended that until the Licensee can conclusively demonstrate that it possesses the necessary managerial and administrative capabilities required to cperate Unit 1 in compliance with all applicable rules and regulations while, at the same time, properly and safely decontaminate and restore Unit 2, permission for restart 1206 123

of Unit 1 must be denied.

Basis for Contention # 14 The past record of violations of NRC rules and regulations and technical specifications by the Licensee demonstrates a lack of managerial and administrative control. The proposed items of noncompliance due to the Unit 2 accident, contained in Appendix IB and Appendix IIF of NUREG-0600, also call such capabillties into question.

In addition, the NUREG-0600 report also point s out numerous examples of actions on the part of the Licensee which, while not constituting violations, collectively demonstrate a ack of managerial and administrative contro'_,

especially with respect to plant operating procedures.

NUREG-0600 ts replete with examples of Licensee staff actions which deconstrate inadequate knowledge of opcrating procedures and basic radiation safety concepts.

Loss of crucial records and lack of maintenance of other records which are required by NRC regulations to be maintatned is further evidence for lack of effective managerial and admintutrative controls.

Specific examples (these are examples, not a definitive itsting) of Ltems which demonstrate lack of sufficient managerial and administrative capabilities on the part of the Licensee include the following, all taken from NUREG-0600 at the indicated page(s):

1296 129

a. Failure of staff to comprehend methods used to evaluate radiation doses and the concepts and techniques used to minimize radiation dose (II-3-68 and 69);

b. Inadequate supply of lapel air samplers ,

extremity monitors, wide-range beta-gamma survey meters, high-range pocket dostmeters, and alarming personnel dostmeters (II-3-68);

c.

Loss of computer recorder records p(rinter I-4-45 and analog through 47 trend and I-4-78 through 81);

d. Potential item of noncompliance, in which eight environmental air samplers had not been calibrated in nearly fLve year s (II-F-4);

e. Potential item of noncompliance, improper change to Health Physics Procedure 1670.7 made without required review and approval (II-F-3);

f. Potential item of noncompliance, personnel assigned to emergency teams did not receive proper training during 1978 (II-F-2 and 3);

g. Failure to declare site emergency u of condition for such declaration (pon detectionII-F-5 and 6);

h. Failure to maintain records of 500 surveys of off-site areas for radiation levels (II-F-6 and 7);

i. Failure to conduct radiatton survey in well-established plume for over fLve hours (II-3-84);

j. Failure to make every reasonable effort to maintain occupational exposures as low as was reasonably achievable (II-3-70 and 71);

k. Failure to keep three independent steam generator emergency feedwater pumps and flowpaths operable during Power Operations due to improper surveillance tests procedure (IB-2).

Contention # 15 It ts contended that the design of the Unit 1 Control Room, instrumentation, and controls is such that operators cannot maintain system variables and systems within prescribed operating ranges during feedwater transients and LCCA's. It is further contended that this violates the provisions of GDC 13 regarding instrumentation and controls. It is contended that in view of the numerous operating difficulties encountered with Unit 2, and the similarttles in design and construction between i296 ,30

Units 1 and 2, a thorough human factors enginering review of Unit l's Control Room is called for in order to provide assurance that the operator-instrumentation interface is such that the operators can exercise adequate control over the reactor and prevent off-site consequences from anticipated operational occurrences and postulated accidents. It is further contended that in order to assure maximum protection for the public health and safety, the human factors engineering reivew and any necessary changes recommended as a result of this review must be completed prior to restart.

Basis for Contention # 15 It is clear from a close reading of NUREG-0600 that there are numerous examples of control room design inadequacies which impacted on the ability of the operators to control the sequence of events during the Unit 2 3/28/79 accident. A few examples of this ares

a. At 18 minutes into :he event, the fuel handling exhaust monitors showed a ramp increase in todine readings , but the location of the instrument on the lower part of the vertical back panel prevented the operator standing at the front panel from viewing the trend (NL' REG-0600, page IA-24) ;

b. Lack of positive indication of valve closure led to permitting PORV on Unit 2 pressurizer to remain open, thus causing loss of reactor coolant (NURBi-0578, pages A-9 and 10);

c. Lack of instrumentation for detecting inadequate core cooling which provided positive indication of such lack led indirectly to core damage during LCCA (NURB3-0578, page A-11 and 12).

In fact , NUREG-0578 at page 7 states, "A widely accepted lesson learned from the TMI-2 accident is that the man-machine 1296 131

interface in some reactor control rooms needs significant improvement."

A study performed by the Electric Power Research Institute (EPRI Report # NP309) detailed human factors engineering problems with nuclear power plant control rooms. From pubitely available pictures of Unit l's control room and Unit 2s control room, it is obvious that many of the concerns discussed in the EPRI Report are present at Three Mi'le Island Unit 1 and that these concerns should be addressed in a human factors engineering review of the Unit I control room. Inasmuch as the ability of the operators to control the reactor and safety systems has a great deal to do with the protection of pu.)1ic health and safety, with specific reference to the requirements of GDC 13 and the general philosophy of the General Design Criteria, these reviews and necessary modifications as a result of the reviews should be accomplished prior to restart.

Respectfully submitted, Yp., b j/

Dated: 22 October 1979 Steven C. Sho11y U/

304 South Market Street Mechanicsburg, PA 17055 (717) 766-1857 (717) 566-3237 w rk numbers (717) 566-3238 i296 132

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION

)

In the Matter of )

) Dock MEIROPOLITAN EDISON COMPANY No 50-289 (Three Mile Island Nuclear )

Station, Unit No. 1) )

)

CERTIFICATE OF SERVICE I hereby certify that single copies of " SUPPLEMENT TO PETITION TO INTERVENE CONTAINING FINAL CONTENTIONS AND BASES SET FORTH WITH SPECIFICITY, STEVEN C. SHOLLY, PEIITIONER" have been served upon the following, either by deposit in the United States mail, postage prepaid, on the 22nd of Ocqober 1979, or by hand at the meeting of intervenor groups in Lancaster, Pennsylvania, on the 21st of October 1979:

Ivan W. Smith, Esquire George F. Trowbridge, Esquire Chairman, Atomic Safety and Shaw, Pittman, Potts, and Licensing Board Panel Trowbridge U.S. Nuclear Regulatory 1800 M Street, N.W.

Commission Washington, D.C. 20006 Washington, D.C. 20555 Secretary of the Commission Counsel for NRC Staff U.S. Nuclear Regulatory Office of the Executive Legal Commission Director Washington, D.C. 20555 U.S. Nuclear Regulatory ATTN: Docketing and Service Commission Section Washington, D.C. 20535 Marjorie A. Aamodt Theodore A. Adler, Esquire R.D. # 5 Widoff, Reager, Selkowitz, and Adler Coates v i.11e , PA 19320 P.O. Box 1547 Holly S. Keck Harrisburg, PA 17105 ANGRY Frieda Berryhill 245 W. Philadelphia Street 2610 Grendon Drive York, PA 17404 Wilmi ngton, Delaware 19808 Robert Q. Pollard Dr. Chauncey Kepford Chesapeake Energy Alliance ECNP 609 Montpelier Street 433 Orlando Avenue Baltimore, MD 21218 State College, PA 16801 Marvin I. Lewis Jordan D. Cunningham, Esquire 6504 3radford Terrace Fox, Farr, and Cunningham Philadelphia, PA 19149 2320 North Second Street Harisburg, PA 17110

)?96 \55

ML19260A226

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