ML19275A333
| ML19275A333 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 08/20/1979 |
| From: | Grueneich D CALIFORNIA, STATE OF |
| To: | |
| References | |
| NUDOCS 7910040121 | |
| Download: ML19275A333 (19) | |
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NUCLEAR REGULATORY COMMISSICN
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Before the Atomic Safety and licensing Scard
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In the Matter of:
)
)
Sacramento Municipal
)
Utilities District
)
Docket No. 50-312
)
(Rancho Seco Nuclear
)
Gbnerating Station)
)
)
Revised Statement of Issueu of Concern to the California Energy Commission In accordance with the prehearing conference held on August 1, 1979, the State Energy Resources Conservation and Development Cc= mission (" California Energy Commission"), participating pursuant to 10 CFR section 2.715(c), submits this revised statement of issues which it intends to address during this prcceeding. /
The Cali-fornia Energy Commission emphasizes that it does not take a position on any of the issues at this time, but is participating to ensure that this proceeding examines carefully all relevant and meaningful issues relating to the operation of the Rancho Seco 1100 166
- T".e California Energy Commissicn will submit a brief cn Au gust 27, 1979, regarding its position on the secpe of the 3 card's jurisdiction in this proceeding.
7 91004o t
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Page la Nuclear Generating Station (" facility").
The California Energy Commission reserves the right to participate on other issues which are approved as proper Ocntentions in this proceeding.
It shall identify any of those issues at a later date after the Board has ruled on the contentions of other parties.- /
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- We have not joined other parties in stipulations of specific 00ntentions because the California Energy Ccamission is participat-ing as an interested state under section 2.715(0).
The prehearing conference was clear that the California Energy Ccm=1ssion should j oin in the stipulaticns only if the state expands its role to a party under se: tion 2.714 Nevertheless, we have distributed these revised issues Oc the NRC staff and Counsel for SMUD in advance of theil submission to the Scard, so that there will be early notice of the maj or areas which we plan to address.
.? age 2 The maj or issues which the California Energy Ccamission has ids..cified as requiring examination in this proceeding are as follows :
1.
'4hether P
short-term modifications and actions described in subparagraphs (a) through (e) o_'Section I'l of the Nuclear Regulatory Commission's ("NRC") May 7, 1979 order are sufficient to provide reasonable assurance that the facility will respond safely Oc feed-water transients, pending completion of the long-term modifications set forth in Section II of the May 7 Order.
This issue will require consideration of whether additional actions should be required prior Oc further facility operation.
rindings contained in NUREG-0560 and NUREG-0578 suggest substantially greater areas of concern that have not been addressed by those actions listed in the 20nfirma:Ory order of May 7, 1979, or addressed by the NRC staff reports of April 25, 1979, or the Compliance Evaluation Report of June 19, 1979, but whose consideration appears essential for the Board to decide whether reasonable assurance of plant safety exists. /
The following are those findings contained in NUREG-0560, MUREG-0573, and the April 25, 1979 Office of Nuclear Reactor Regulation Status Reper applicable to 9/
Rancho Seco for which inquiry should be required: /
" A general lessen learned from our review of the TMI-2 sccident is that the frequency "ith which some safety systems, such as the high-pressure safety injection system (part of the Emergency Core Cooling System provided pursuant to General Design Criterion 35 of 13 CFR part 50, Appendix A), are called upon to fune:10n for reac cr ccolan system pressure or volume control may exceed their generally understood and previously accepted design basis."
(NUREG-0573, p. 6)
"The heat removal path by natural circulation is no well understcod."
(NRR Status Report, April 25, 1979, p. 1-5) 1100 u3 tr
? age 3 "A generic question raised by TMI-2 is the need to expand the applicability of existing reliability criteria to equipment not previously included in the licensing interpretation of equipment designated as 'important to safety'."
(NUREG-0573, p. c)
"The TMI-2 accident sequence included a failure of a power-operated relief valve to close.
This and other operating experience raise a significant question about the performance qualificaticn of two types of valves in the primary coolant boundary; safety and relief valves."
(NUREG-0573, p. 7)
"A widely accepted lesson learned from the TMI-2 accident is that the can-machine interfac; in some reactor control rocas needs significant improvement."
(NUREG-0578, p. 7)
"The need for an emergency feedwater system of high reliability is a clear lesson ~1 earned from the TMI-2 accident.
The IE Sulletins and the Commission's Confirmatcry Shutdown Order for the B&W designed plants deal with this aspect of the accident in scme respects."
(NUFEG-0573, p.
10)
"In the Three Mile Island accident, the performance o f im-portant safety systems was degraded due to human errors.
Scme cf the human errors during the TMI accident were caused, in part, by inadequate coordination of transient and accident analysis, emergency procedure preparation, and operator training.
In its study of the accident, the Task Force has found that, in the pas-shd full analytical capabilities of the licensees and re ec' andors have not been used in the development of emergency p.3cedursa or in the training of reactor operators.
Similarly, the NRC review of emergency i
i100 169
? age 4 procedures and operator tr '-ning has placed little or no emphasis en the appropriateness of the analytical bases of the procedures or training.
A substantial improvement in safety can be obtained by improving operator performance during transients and accidents. "
(NUREG-057 8, p. 11)
"The Task Force has concluded that the need for improved operations reliability is the most important lessen learned from the accicent at TMI-2.
Cne part of this overall lesson that is amenable to early implementation inciudes more definitive and clearly articulated operations command responsibilities and improved adminis-trative procedures and controls (to support the ccamand function) for both normal and emergency conditions.
Improvements in operator qualifications, training and licensing; technical qualifications of overall reactor operations organizations; and display and system diagnostic equipment will be recommended by 'IRR and others in the coming months."
(NUREG-0578, p. 12)
" Prior to the TMI-2 accident, the general approach usec for accident analyses was to ensure conservatism in the analysis models and results.
Consideration has been given to the development of best-estimate codes, but licensing calculations were done on a con-servative bases.
It is recognized that shortcomings resulted from this apprcach."
(NUREG-0560, p.
2)
"There are locations in the primary system where steam or cther gases can accumulate if the primary system is permitted to de-pressurize to saturation conditions.
These loca icns are in the 1100 170
Page 3 upper reactor vessel, in the region of each reactor coolant pump, and in the upper level of the hot legs and steam generators.
There appears to be no specific reason for voids to accumulate only in the pressurizer under these conditions, although during normal operation only the pressurizer is operated at saturation conditions."
(NUREG-0560, p. 8-2)
"The B&W once-through steam generators have much smaller water inventories than those associated with Combustion Engineering and Westinghouse plants.
As a result, the B&W steam generators boil off cn loss of feedwater much more quickly.
This leads to a more rapid increase in primary pressure on loss of main feedwater in 34W plants and therefore requires greater performance and reliability of the AFW delivery."
(NUREG-0560, p. 3-3)
"Feedwater transients have been initiated from a variety of human and equipment failures.
Although some improvements can and should be made to feedwater system reliability and to identify and correct design deficiencies, the occurance of feedwater transients cannot be eliminated."
(NUREG-0560, p. 3-3)
"The design requirements and criteria for plant process controls are not well defined in NRC regulations.
Furthermore, the interaction of these features, especially in the B&W integrated control system and the auxiliary feedwater system, have not been thoroughly explored in previous NRC licensing reviews.
The plant control systems play an essential part in plant operations and the control of transient situations that would otherwise introduce challenges to the plant safety sys tem....
Failure.cf cent rols could initiate a transient or could inhibit the ccntrol of a transient other-wise mitigated."
(NUREG-0560, p.3 4 )
1100 i71
Page 6 "As related to the TMI-2 accident, the failure of the PCRV to close changed a loss-of-feedwater t"ansient into a small loss-of-coolant accident.
This was not immediately apparent to the operators."
(NUREG- 0560, p. 8-5)
"For reasons not yet understood, the low-pressure heat removal system was not placed into operation during the early (first 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) atages of the accident."
(NUREG-0560, p. 8-7)
"The number and complexity of possible event sequences for nuclear pcwer reactors make it impossible to assure that operators are specifically trained to respond correctly to each and every off-normal or accident condition."
(NUREG-0560, p.3-7)
" Training programs have underemphasized the possibility of failures in various systems, nonstandard passive conditions (misaligned systems), possible failure of engineered safeguard equip-ment when called upon, and even the effects o f multiple failures.
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as a design basis, it is not clear that it sho uld be considered as a limiting basis for training purposes.
Training aspects include the technical staff."
(NUREG-0560, p. 3-9)
"The operator has been trained to rely on his instrumentation.
He will continue to do so until he suspects an erroneous reading; howaver, he must be trained not to rely solely on a single indicatcr since it may be erroneous or misleading ander certain circumstances....
If the operator has too many additicnal manual functions to perform, he may reduce hic Obse"vations on other system parameters, which may lead him to have tunnel vision."
(NUREG-0560, p. 3-11) i100 172
.Page t
" Human factcrs engineering has not been sufficiently emphas&2ed
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in the design and layout of the control rooms.
The location of instruments and controls in many power plants often increases the likelihood of operator error or, at the least, impedes the operator in efficiently carrying out the normal, abnormal, and emergency actions required 'of him."
(NUREG-056C, p. 5-11)
"The analysis of feedwater and other anticipated operational transients has been found to be sc=ewhat idealized in terms of the TMI-2 accident.
The models are simplistic and do not always include provisions to consider single failures and progressively degraded conditions based upon human error and/or equipment =alfunction."
(NUREG-0560, p. 8-12)
"Small break LOCA events have been extended dcwn to the 2
range of approximately 0.05ft It was believed that smaller breaks were well within the capability of the available coolant makeup systems and were not limit in g.
Recent preliminary calculations of the TMI-2 accident performed a Idaho Nuclear Engineering Laboratory (INEL) show evidence that suggests voiding in the coolant system can occur in conj unction with a rising water level in the pressuriner.
This is also predicted from new studies performed by the ?WR vendors Ihe TMI-2 accident indicates that the possible effects cn core coolability for smaller breaks are not conpletely understood."
(NUREG-0560, p. 3-13)
"The computer codes general 2 y used for transient and small break LOCA analyses are complex and do not always include provisicns 1100 173
'Page 8 for extending the calculations to cover the event duration through the time period until stable cooling (e.g. cold shutdown) is achieved.
In some cases conservative bounding types of assumptions and models are used that may mask out realistic system and equipment behavior.
In addition, many of the vendor codes have not been reviewed in detail by the NRC."
(NUREG-0560, p. 8-14)
"The applicable GDC for anticipated transients (e.g., GDC 10, 13, lu, and 15) appear to reasonably encompass the necessary require-ments for plant design features.
Although the GDC may be adequate, their general nature leads to broad interpretation of specific requirements.
The matter of defining a passive failure, as noted in Appendix A to 10 CFR 50, and its application to such failures as the FORV or other valves leads to misunderstanding as to their treatment in transient and accident analyses."
(NUREG-0560, p. 3-15)
"There are Technical Specifications requirements that appear to place excessive reliance on single parameters, such as pressurizer level centrol, and do not include the significance of other parameters that the operator should be considering while making plant adj ust-ments and action decisions.
Reporting requirements appear to be too narrowly constrained to violations of Technical Specifications."
(NUREG-0560, p. 8-16)
An additional issue is the facility's ability to safely respcnd to Repcrtable Occurence No 79-6 for Rancho Seco reported in an [[letter::05000312/LER-1979-006-02, /02T-0:on 790720,B&W Informed District That Addl Small Break LOCA Analyses Indicated That Certain Spectrum of Small Leaks in RCS Coupled W/Loss of Reactor Coolant Pumps Later in Transient Could Cause 10CFR50,App K to Be Exce|August 2, 1979 letter]] from J. Mattince, Assistant General Manager and Chief Engineer for SMUD to Mr.
R.H. Engelken, Director, Region 1100 174
PaSe 9 7 Office of Inspection and Enforcement, NRC.
As stated in the letter:
"These analyses identified a potentially unsafe condition.
Basically, the analyses have shown that loss of reactor coolant pumps sometime after two minutes into an event involving a certain spectrum of small breaks (about 0.025 ft2 to 0.20 ft2) could e.cceed 10 CFR 50, Appendix K criterion....
This is con-sidered a very low probability event.
However, the consequences of the worst case event could result in major fuel damage."
2.
Whether the long-term modifications described in the NRC's May 7 Order are sufficient to provide continued reasonable assurance that the facility will respond safely to feedwater transients.
This issue will require consideration of whether additional actions,
including actions suggested by the NRC staff in NUREG-0500 and NUREG-0578, are required to provide reasonable assurance that the facility will respond safely to feedwater transients-and the time-table for implementation.
Actions and areas requiring consideration include those set forth under Issue 1, above.
3 Whether facility operators and associated personnel have adequate training and experience to respond safely and responsibly to feedwater transients and other unexpected events.
This issue will require inquiry, inter alia, into whether the recent training and testing of facility personnel to respond to the kinds of problems encountered at Three Mile Island represents an acceptable level of operation competence.
This issue also will require inquiry into:
Whether persennel adequately understand the nechanics of the facility, basic reactor physics, and other fundamental aspects o f its operation?
1100 17f:r
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Whether personnel are properly apprised of new information pertinent to the facility's safe operation and ability to respond to transients, particularly information on operating experience of other reactors?
Whether NRC and SMUD adequately ensure that emergency instruc-tions are understood by and are available to plan' cersonnel in a nanner that allows quick and effective implementat on during an emergency?
Whether adequate procedures are used to assure that plant personnel and SMUD management report unsafe or improper practices or conditions at the facility to SMUD and/or the NRC7 Whether adequate design and operation changes have been made to the Rancho Seco control room to allow ope stors to bette" comprehend and respond to upset conditions in the facility, developing act ient situations, and abrupt, unscheduled shutdowns? Factors involved include the plant-status information available to the Operators from
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to such situations, human responses under stress conditions, and control "com layout and data presentation in design of annunciators, warning lights, and display panels.
Whether the procedures and training adopted as a result of
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Page 11 Whether the problems identified by the NRC staff in NUREG-0560 and NUREG-0578 relating to operator training and experience (listed under Issue 1, above) have 'oeen adequately resolved?
4.
Whether, notwithstanding measures taken and contemplated to deal with feedwater transient problems, the facility should be required to revise emergency planning procedures so that, in the event of future problems, persons in the immediate reactor area and in the 9/
facility's reasonable impact area will not be exposed to dange &.
As stated in NUREG-0560:
"Although some improvement; can and should be made to feedwater system reliability and to identify and correct de-sign deficiencies, the occurance of feedwater transients cannot be eliminated....
The emphasis should be on coping and mitigating the consequences of feedwater transients."
This issue will encompass certain of the same concerns raised in NUREG-0396 and a U.S. GAO Report ESO 73-110, March 30, 1979 This issue will require analysis of whether the facility 's current energenc,y plans and the state and local plans associated therewith are adequate, or whether changes should be required within a definite timeframe or before the fac;2_..
is permitted to operate further.
This issue will also require in-quiry into:
Whether the scope of accidents covered by the facility's emergency planning procedures should be expanded to cover planning for protection of Class 9 accidents, TM:-level incidents, and other more serious events not currently covered?
Whether accident notification procedures such as the criteria for requiring NRC notification used by SMUD should be revised?
7
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.Page 12-5 Whether, notwithstanding measures taken and contempl ated to deal with feedwater transient problems, the facility should be required to revise its accident responses and mitigation measures so that, in the event of future problems, the risk of hazardous consequences will be reduced.
This issue will require analysis of the following:
Whether those systems identified as contributing to releases of radioactivity during the TMI accident, which are outside contain-T.e n t, should be changed to vent into the containment building?
Whether the containment building should be mcdified 50 provide overpressurization protection with a controlled filtered venting system to mitigate unavoidable releases of radionuclides?
Whether the following hazards associated with transients by the TMI Task Force in NUREG-0578 have been adequately resclved:
a)
"The NRC staff and the ACRS have for some years emphasized the need for special features and instruments tc aid in accident diagnosis and control.
Although some degree of capability of this type was available at TMI-2, and exists on other plants, the TMI-2 experience shows that more is needed."
(p. 11) b)
"Our current understanding of the response of the licensee to the accident at Three Mile Island shows a need to improve cperati:ns procedures and preparations for accident conditions."
(p. 13) 1100 17B-
, P a ge 13 c)
"At TMI-2, the systems external to the containment building that contained radioactive material had several deficiencies.
For example, the licensee had little knowledge of their operational leakage characteristics, and shielding provisions for personnel access were inadequate.
The difficulties arose not only in safety systems, but also in systems outside the scope of previous ' safety grade' requirements (such as the maekup and letdown system)."
(p. 9) d)
" Evaluation of the containment isolatfon experience 9.t TMI-2 shows that design features at some other plants may be inadequate in three respects.
The sequence of events at TMI-2 illustrated the need for careful reconsideration of the isolation provisions of non-essential systems inside containment.
Reconsiderat_on should include the identification of those systems that can be isolated indefinitely and those systems that should be selectively isolated only after it is established that they are not essential to continued core cooling or performance of engineered safety features.
- Third, the experience gained at TMI-2 indicates that the resetting of the containment isolation signal in some designs may result in automatic reopening of some containment isolation valves."
(p. 3) e)
"The TMI-2 accident resulted in the production of quantities
.of hydrogen gas in excess of the amounts required by NRC regulations to be considered in the design and accident analysis of nuclear power plants."
(p. 3) i100 179
Page 14 o.
Whether, notwithstanding the short-term and long-term modi-fications described in the May 7 Order, the facility should be required to operate at less than full' rated capacity in order to produce an additional margin of safety to respond to feedwater and other transients,
pending a complete analysis and understanding of the ramifications of the Three Mile Island accident.
7.
Whether the facility should be required to retrofit as promptly as practicable in order to have the same or better safety devices to respondato feedwater transients as are required on new plants which are currently being licensed.
"This issue will requira, inter alia, analysis into devices and procedures which are being required for new planto prior to receiving NRC licenses as a result of TMI found on the facility.4/
but which are not 8.
Whether the Three Mile Island events and subsequent inquiries and analyses have identified areas in addition to transients originated by. failures in the feedwater systems where current design margins are inadequate to provide reasonable assurance that the facility will respond safely if a problem should arise.
If so, should this Board require further modification to correct these deficiencies, either prior to further facility cperation or according to a definite timetable?
Areas where inquiry should be required are as follows:
Whether proper considerations have been given tc other transients with potentially equally hazardous consequences as thos.e initiated in the feedwater system and which may not occur at higher frequences as a result of those actions taken at Rancho Seco since Three Mile Island?
The principal areas of concern 'are those plan: transients occurring 1100 160
Page 15 with a loss of offsite power and the increased frequency of full power reactor trips as a result of compliance with item (c) in the May 7 Order.
Whether the following has 'ceen resolved:
(a)
"In some designs, loss of pressurizer heaters due to a loss of offsite power requires the use of high-pressure enargency core cooling system to maintain reactor pressure and volume control for natural circulation cooling.
Similarly, in some designs the inability to close the power-operated relief valve upon loss of off-site power.could result in additional challenges to the high-pressure emergency core cooling system. "
9 Whether the procedures and criteria used by the NRC and SMUD for determining when to require corrective action or to s.lut down the facility are sufficeint to provide reasonable assurances that the facility will be operated safely.
For instance, a memo entitled
" Decay Heat Removal During a Very Small 3reak LOCA for a 3&W 205 -
Fuel Assembly PWR" was written by C. Michelson, a member of TVA's staff and now a consultant to the Advisory Committee on Reactor Safe-guards in January, 1973 which warned that small pipe breaks in 3&W reactors could lead to the sequence of events that actually occurred at Th:.'ee Mile Island.
Though Michelson pointed out that "these un-certainties may re flect on the adequacy of proposed emergency operat-ing procedures and operators training for a very small break 1CCA",
his report failed to result in adequate modifications to prevent the Three Mile Island accident.
In addition, while the [[letter::05000312/LER-1979-006-02, /02T-0:on 790720,B&W Informed District That Addl Small Break LOCA Analyses Indicated That Certain Spectrum of Small Leaks in RCS Coupled W/Loss of Reactor Coolant Pumps Later in Transient Could Cause 10CFR50,App K to Be Exce|August 2, 1979 letter]] from Mr. Mattimoe cf SMUD (see Issue.1, above) notified the i100 1El
Page lo NRC that recent snall break analyses indicate a " potential unsafe condition", the problem has yet to be resolved.
10.
Whether the procecures and criteria used by SMUD and the NRC for determining the actions necessary prior to restart of the facility after either a forced outage or shutdown required by the NRC, are sufficient to provide reasonable assurances that the facility will be operated safely.
Respectfully submitted,
,s
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DIAN M.
GRUENEICH CHRISTOPHER ELLISON Attorneys for the California Energy Commission
~1'100 182
Page 17 FOOTNOTES 1.
The necessity of examining these additional matters has been recognized by the NRC itself in the "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations" (July, 1979), in NRC 's statement that it has identified measures "whose implementation is judged to provide substantial additional protection which is required for the public health and safety" (p. o).
2.
It is anticipated that further potent 4a' actions or areas of concern not listed here which will also require review during this proceeding will co'me to light after participants in the proceeding are afforded discovery.
3 Recent NRC Licensing Board notifications concerning Three Mile Island releases have contained the following, showing risks may be greater than originally anticipated:
"10 CFR Part 100 required that the assumed fission product release used for site suitability calculations should be one
'that would result in potential hazards not exceeded by those from accident considered credible.'
The TMI release of 13 million curies of Xe-133 is substantially greater than that which was estimated as the maximum credible release by the staff in its review of the OL for TMI-2 and is orobably larger than that which would be predicted to occur in any.cf the. site suitability analyses for plants reviewed by the staff in the last decade."
4 Because of lack of discovery prior to preparation of this statement, the California Energy Commission is unable to list at this time all specific devices and procedures required of new plants but not Rancho Seco.
Instead, we reference the July 2, 1979 NRC Memorandum frca H.
Rood to Robert L. Saer, " Summary of Meeting to Discuss Case-Work Schedules" which describes a meeting of Harold Denton, Director 1100 1857
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'Page 13 Cffice of Nuclear Reactor Regulation, and other members of the NRC staff with representatives of utility companies having CP ani 01 applications under review.
The minutes of the meeting indicr.te t:.at additional requirements will be imposed for new pcwer plants as the result of Three Mile Island but does not specify what those require-ments will be.
An interview with Dominic Vassalo of the NRC licensing staff reported in the July 13, 1979 Energy Daily also indicates that additional requirements will be imposed for new plants.
- Finally, at the June 21, 1979 NRC staff briefing before the California Energy Commission, Mr. Centon Ross stated that a new plant would have to comply fully with NUREG-0560 before receiving a CP whereas Rancho Seco is not se required (RT 94).
1100 184'
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