ML20052F554
| ML20052F554 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 05/07/1982 |
| From: | Shum D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20052F540 | List: |
| References | |
| NUDOCS 8205130182 | |
| Download: ML20052F554 (14) | |
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UNITED STATES OF AfiERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of TEXAS UTILITIES GENERATING COMPANY,)
Docket Nos. 50-445 ET A1 50-446 (Comanche Peak Steam Electric
)
Station, Units 1 and 2)
)
AFFIDAVIT OF DAVID SHUM I, David H. Shum, being duly sworn, do depose and state the following:
Q.1. By whom are you employed, and what is the nature of the work you perform?
A.1. I am employed as a Senior Containment Systems Engineer in the Containment Systems Branch of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission. A statement of my professional qualifications, including the nature of the work that I perform, is attached to this affidavit.
1 Q.2. What is the nature of the responsibilities you have regarding the Comanche Peak Steam Electric Station ("CPSES")?
A.2. I was responsible for the review of Sections 6.2.1 through 6.2.6 of the CPSES Final Safety Analysis Report ("FSAR"), with regard to the adequacy of the containment systems.
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Q.3. Would you describe the subject matter of your affidavit?
A.3. My affidavit addresses Board Question 1, which states:
Describe in detail the planned method of handling any hydrogen gas in the CPSES containment structure.
In particular, I will discuss why the current CPSES design for handling hydrogen gas conforms to the NRC rules and regulations for a combustible gas control system in light water cooled power reactors.
Q.4. Have you read Applicant's Response to Board Question 1, concerning the conformance of the CPSES design to current NRC regulations? Do you believe that Board Question I raises a serious safety issue which would preclude the issuance of an operating license for CPSES?
A.4. Yes, I have read the Applicants' Response to Board Question 1.
Board Question 1 does not raise a serious safety issues which would prohibit the issuance of an operating license for CPSES.
Q.5. What are the current NRC design requirements for the control of post-accident hydrogen?
A.S. The current NRC design requirements for the control of post-accident hydrogen are specified in 10 C.F.R. 6 50.44. According to 9 50.44, boiling water reactor ("BWR") and pressurized water reactor ("PWR") plant designs must: (a) include means for control of hydrogen that may be generated by core metal-water and corrosion
of metals following a loss of coolant accident ("LOCA"); (b) be pro-vided with the capability for measuring the hydrogen concentration in the containment, insuring a mixed containment atmosphere and controlling combustible gas concentrations in the containment; and (c) be provided with an inerted atmosphere or an oxygen deficient condition if certain conditions cannot be met prior to effective operation of the combustible gas control system.
Postaccident conditions should be such that an uncontrolled hydrogen / oxygen recombination would not take place in the containment.
If this condition cannot be met the plant must be designed to withstand the consequences of uncontrolled hydrogen / oxygen recombination without loss of safety functions.
Q.6. Did you review and evaluate the CPSES design for control of post-accident hydrogen to see if it conformed with 10 C.F.R. $ 50.44 requirements?
A.6. Yes, I reviewed CPSES for compliance with 10 C.F.R. 9 50.44. My evaluation is included in the Safety Evaluation Report ("SER") for CPSES in Section 6.2.4, which is attached and incorporated into this affidavit.
Q.7. Was your review of the CPSES design for control of post-accident hydrogen control performed in accordance with Standard Review Plan
("SRP") Section 6.2.5?
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A.7. Yes.
Section 6.2.5 of the SRP establishes the criteria and areas of review concerning the control of combustible gases in the con-
concentration in conformance with the provisions of the Regulatory Guide 1.7, " Control of Combustible Gas Concentra-tions in Containment Following a loss-of-Coolant Accident,"
Revision 1.
The Applicants have used the same assumptions as Regulatory Guide 1.7 to calculate the rate of hydrogen released by radiolysis and corrosion of metals, and a 5%
zirconium-water reaction in the reactor core. The analysis indicates that the hydrogen concentration in the containment would not reach the lower flammability limit of 4 v/o (volume
%) until about 26 days after a postulated LOCA. Hydrogen recombiner operation, however, will be initiated within 24 hr 1
after a postulated LOCA.
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2.
The proposed recombiner system incorporates several design features that are intended to assure the operability of the system in the event of an accident. Among these are:
(1) seismic Category 1 design; (2) protection from missile and jet impingement; and (3) redundancy to the extent that no single component failure disables both recombiners. A post-accident vent system is also provided for post-accident cleanup of the containment atmosphere.
3.
The Applicants calculate that the hydrogen concentration will be limited to 1.9 v/o with operation of a single recombiner started 1 day after an accident. The hydrogen concentration of 1.9 v/o calculated by the Applicants is significantly lower
tainment following a loss-of-coolant accident. The review includes the following general areas:
1.
The production and accumulation of combustible gases within the containment following a postulated loss-of-coolant accident.
2.
The capability to mix the combustible gases with the con-tainment atmosphere and prevent high concentrations of combustible gases in local areas.
3.
The capability to monitor combustible gas concentrations within containment.
4.
The capability to reduce combustible gas concentrations within containment by suitable means, such as recombination, dilution, or purging.
Q.8. What were your conclusions concerning the acceptability of the CPSES design for control of post-accident hydrogen?
A.8. The design of the combustible gas control system is acceptable because it conforms to the current regulation, 10 C.F.R. s 50.44.
This conclusion is based on the following:
1.
The Applicants have analyzed the production and accumulation of hydrogen within containment using the guidelines of Branch Technical Position CSB 6-2, " Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident." The Applicants will provide redundant Westinghouse electrical thermal hydrogen recombiners to limit the hydrogen
than the lower flammability limit of 4 v/o as expressed in the Regulatory Guide 1.7.
4.
The Staff has performed a similar analysis of hydrogen generation and hydrogen accumulation in the containment following a LOCA, and the Staff's results of about 2.0 v/o l
is in agreement with the Applicants'.
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5.
Mixing of combustible gases within containment, to prevent l
j excessive stratification and to ensure uniformity of the i
l hydrogen concentration throughout containment, is provided by the containment spray systems. The turbulence accompanying the initial blowdown and natural circulation within contain-I ment will also enhance the mixing process. The Staff finds that these systems and mechanisms are adequate to ensure an essentially uniform hydrogen concentration within containment, and limit the potential for local hydrogen pocketing.
6.
The Applicants have committed to provide redundant contain-ment hydrogen concentration indication (over a range of 0%
to 10%) at the main control board.
In addition, the para-meters will be provided as input to the safety parameter display system (SPDS) for recording. The instruments will be designed to meet the provisions of Regulatory Guide 1.97, Rev. 2.
The Staff concludes that the containment hydrogtn monitoring instruments provided for Comanche Peak are acceptable.
The above statements and opinions are true and correct to the best i
of my personal knowledge and belief.
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v David H. Shum l
Subscribed and sworn to before me A ay of May, 1982.
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DAVID H. SHUM PROFESSIONAL QUALIFICATIONS i
i I am a Senior Containment Systems Engineer in the Containment Systems' Branch of the Nuclear Regulatory Commission.
In this position I am responsible for the technical analysis and evaluation of the public health and safety aspects of containment systems.
From September 1980 to June 1981. I was assigned to review the Final Safety Analysis Report for Comanche Peak Steam Electric Station, Units 1 and 2.
I participated in the preparation of NUREG-0797, " Safety Evaluation Report Re-lated to the Operation of Comanche Peak Steam Electric Station, Units 1 and 2."
Prior to September 1980, I have had about 14 years of experience in the nu-clear engineering field.
F I received a Master of Science (M.S.) degree in Mechanical Engineering from North Carolina State University in 1966 and a Bachelor of Science (B.S.) de-gree in Mechanical Engineering from the National Taiwan University in 1963.
I am a registered Professional Engineer in Ohio and Pennsylvania.
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NUREG-0797 Safety EvaDuation Report
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re atec :o t7e 03eration o" Comancle 3ea < S::eam ' Ele ~ tric Sta: ion,c ::
c Units 1 anc 2 Docket Nos. 50-445 and 50-446 Texas Utilities Generating Company, et al.
t U'.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation July 1981 r "c a
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Tbe staff will' require that the leakage integrity tests be performed on the 48-in. containment purge system isolation valves and on the 18-in. containment pressure relief system isolation valves in accordance with the following recommended testing frequency:
"The leakage integrity tests' of the isolation valves in the containment a.
purge lines shall be conducted at least once every six months.
b.
"The leakage integrity tests of the isolation valves in the containment pressure relief line shall be conducted at least.once every three months."
The purpose of the leakage integrity tests of the isolatio'n valves in the containmentpurgeandpressurerelieflinesistoidentifyexcessivedeg(ducted adation of the resilient seats for these valves. Therefore, they need not be con with the precision required for the Type C isolation ~ valve tests in 10 CFR'
.These tests would be performed in addition to the quanti-Part 50, Appendix J.
tative Type C test.s required by Appendix J and would not relieve the applicant of the responsibility to conform to the requirements of Appendix J.
The staff has reviewed the containment isolation system; finds it'is in
'conformance with GDC 54, 55, 56, and 57, and Regulatory Guide 1.11; and concludes
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the design of the containment isolation system is acceptable.
6.2.4, Combustible Gas Control System Following a LOCA, hydrogen may accumulate within the containment as a result of:
(1) metal-water reaction between the fuel cladding and the reactor coolant; (2) radiolytic decomposition of the postaccident emergency cooling water; and (3) corrosion of metals by ECC and containment spray solutions. The applicant has analyzed the production and accumulation of hydrogen within containment from the above sources using the guidelines of Branch Technical Position CSB 6-2,
" Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident." The applicant will provide redundant Westinghouse electrical thermal hydrogen recombiners to limit the hydrogen concentration in conformance with the provisions of the Regulatory Guide 1.7, " Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident," Revision 1.
The applicant has used the same assumptions as Regulatory Guide 1.7 to calculate the rate of hydrogen released by radiolysis and corrosion of metals, and a 5%
zirconium-water reaction in the reactor core. The analysis indicates that the hydrogen concentration in the containment would not reach the lower flamm&bility limit of 4 v/o (volume %) until about 26 days after a postulated LOCA. Hydrogen recombiner operation, however, will be initiated withih 24 hr after a postulated LOCA.
i The proposed recombiner tystem incorporatas several oesign features that are intended to assure the operability of the system in the event of an accident.
(1) seismic Category 1 design; (2) protection from missile Among these are:
and jet impingement; and (3) redundancy to the extent that no single component failure disables both recembiners. A pestaccident vent system is also provided for postaccident cleanup of the containment atmosphere.
The applicant calculates that the hydrogen concentration will be limited to 1.9 v/o with operation of a single recombiner started 1 day after an accident.
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t The staff has performed a similar analysis of hydrogen generation and hydrogen accumulation in the containment following a LOCA, and the staff's results are t
in agreement with the appifcant's.
Based on its review, the staff concludes that the combustible gas control system design meets the requirements of GDC 41, 42, and 43 and the prov'isions of Regulatory Guide 1.7, and is, therefore, acceptable.
f 6.2.5 Containment Leakage Testing Program The containment design includes the provisions and features necessary to satisfy i
the testing requirements of Appendix J to 10 CFR Part 50. The design of the containment penetrations and isolation valves will permit periodic leakage rate j
testing at the pressure specifjed in Appendix'J.
Included are those penetrations l
that have gasketed seals and electrical penetrations.
u.
The containment leakage testing program complies with' the requirements of Such compliance provides adequate assurance that containment leak-Appendix J.
tight integrity can be verified throughout service lifetime-and that the leakage rates will be periodically checked during service on a timely basis to maintain The plant'ges within the specified limits of the Technical Spe such leaka requirements for containment leak testing, including test frequencies.
Maintain,ing containment leakage rates within such limits provides reasonable assurance that, in the event of any radioactivity releases within the containment, the loss of the containment atmosphere through leak paths will not be in excess I
of acceptable limits specified for the site; that is, the resultant doses will be well within 10 CFR Part 100 guidelines in the event of a des.ign-basis LOCA.
The staff concludes that the applicant's program complies with the requirements of Appendix J, and provides an acceptable basis for satisfying the requirements of GDC 52, 53, and 54.
m 6.3 Emergency Core Cooling System 6.3.1 Design Basis The emergency core cooling system (ECCS) is designed to cool the reactor core and to provide shutdown capability.for accident conditions that result in These accident significant depressurization of the reactor coolant system (RCS).
conditions include mechanical failure of the RCS piping up to and including
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f instantaneous circumferential rupture of the largest pip'e in the RCS, ejection
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i of a rod-cluster control assembly, pipe breaks in the steam system, and steam I
generator tube rupture.
The design bases for selecting ECCS functional requirements are derived from I
Appendix X limits for fuel cladding temperature as delineated in 10 CFR 50.46.
Subsystem functional parameters have been selected so that Appendix K require-l ments are met over the range of anticipated accidents and single-failure i
l assumptions.
The applicant states that the requirements will be met with niinimum engineered safeguards available, such as loss of one emergency power
. bus with loss of offsite power.
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pctential for an accident that could affect a particular control system, and effects of the control system failures, may differ frcm plant to plant.
Tnerefore, it is not possible to develop generic answers to these concerns, but rather plant-specific evaluations are required.
The purpose of this Unresolved Safety Issue is to define generic criteria that will be used for plant-specific evaluations.
The Comanche Peak control and safety systems have been designed with the goal of ensuring that control system failures will not prevent automatic or manual initiation and operation of any safety system equipment required to trip the plant or to maintain the plant in a safe-shutdown condition following any
" anticipated operational occurrence" or " accident.." This has been accomplished by either providing independence between safety and nonsafety systems or pro-viding isolating devices between safety and nonsafety systems. These devices preclude the propagation of nonsafety system equipment faul.ts to the protection system.
This ensures that operation of the safety system equipment is not impaired.
A systematic evaluation of the control system design, as contemplat.ed for this Unresolved Safety Issue, has not been performed to determine whether postulated accidents could cause significant control system failures which would make the accident consequences more severe than presently analyzed.
However, a wide range of bounding transients and accidents is presently analyzed to ensure that-the postulated events such as steam generator overfill and overcooling events would be adequately mitigated by the safety' systems.
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The staff has requested additional information on this matter, as described in Section 7.7.2 of this report.
Subject to an acceptable resolution of this out-standing issue, the staff has concluded that there is reasonable assurance that Comanche Peak can be operated before the ultimate resolution of this generic issue without endangering the health and safety of the public.
A-48 Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Eouioment:
Following a loss-of-coolant accident in a light water reactor plant, combustible gases, principally hydrogen, may accumulate inside the primary reactor contain-ment as a result of:
(1) metal-water reaction involving the fuel element cladding; (2) the radiolytic decomposition of the water in the reactor core and the containment sump; -(3) the corrosion of certain construction materials by the spray solution; and (4) any synergistic chemical, thermal, and radiolytic effects of postaccident environmental conditions on containment protective ~
coating systems and electric cable insulation.
Because of the potential for significant hydrogen generation as the result of an accident, 10 CFR 50.44, " Standards for Combustibie Gas Control System in Light Water Cooled Power Reactors" and GDC 41, " Containment Atmosphere Cleanup,"
require that systems be provided to control hydrogen concentrations in the containment at.Tosphere following a postulated accident to ensure that containment integrity is maintained.
10 CFR 50.44 requires that the combustible gas control system provided be capable
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of handling the hydrogen generated as a result of degradation of the emergency C-18
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I core cooling system.such that the hydrogen release is five times the amcunt calc'ulated in demonstrating compliance with 10 CFR 50'.46 or the amount corres-ponding to reaction of the cladding to a depth of 0.00023 in., whichever is greater.
The accident at TMI-2 on March 28, 1979 resulted in hydrogen generation well in excess of the amounts specified in 10 CFR 50.44.
As a result, it became apparer.t to NRC that specific design measures are needed for handling larger j
hydrogen releases, particularly for smaller low-pressure containments. Thus, the Commission determined that a rulemaking proceeding should be undertaken to define the manner and extent to which hydrogen evolution and other effects of a degraded core need to be taken into account in plant design. An advance notice of this rulemaking proceeding on de~ graded core issues was published in
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the Federal Register on October 2, 1980.
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' Recognizing that a number of years may be required to. complete this rulemaking
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proceeding, a set of short-term or interim actions relative to hydrogen control requirements were developed.and implemented.
These interim measures were described in a second October 2, 1980 Federal Register notice.
For plants with large ~ dry containments, such as Comanche Peak, no near-term mitigation measures are required by the interim rule.
4 Comanche Peak has about 3 million ft3 of net free volume.
Assuming 30 to 50%.
O metal-water reaction in the core, the resulting uniformly mixed concentration of hydrogen in the containment will range from 6 to 10%. This is well below the concentrations for detonation and even below the limits for combustion with r
expected. steam concentrations in the containment atmosphere following a LOCA.
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The design pressure of Comanche Peak is 50 psig.
Analyses performed on the I
Zion and Indian Point plants show that the failure pressures are greater than twice the design pres.sures.
The staff believes, therefore, that the failure pressure of the Comanche Peak containment would be considerably in excess of the design pressure.
If the substantial amount of metal-water reaction were to occur shortly after the onset of a large LOCA and while the containment is still near its peak pressure, the pressure increase caused by the noncondensible hydrogen gas and its associated exothermic formation energy will be substantially less than the failure pressure.
If the metal-water reaction were to occur well after the onset of the large'LOCA, the containment heat removal system would have con-densed much of the steam in the containment and reduced the containment pressure.
This-would provide a substanial margin for accommodating hydrogen generated by the metal-water reaction.
A substanti'al margin would exist for accommodating the hydrogen generated by the metal-water reaction.
At this later time, the containment heat removal system would be able to condense much of the steam in the containmnet and reduce the containment pressure.
l In addition, tee "Short-Term Lessons Learned" from the TMI-2 accident have been implemented at Lomanche Peak.
This action will reduce the likelihood of accidents that could lead to substantial amounts of metal-water reaction.
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O Accordingly, pending resolution of this Unresolved Safety Issue and the rule-7 making p'roceeding on hydrogen generation, the staff concludes that Comanch Peak can be operated without undue risk to the health and safety of the public.
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