ML19260D792

Response to Licensee 800124 Motions for Summary Disposition. Opposes Blanket Motion & Disposition of Castro & Hursh Remaining Contentions

ML19260D792
Person / Time
Site:	Rancho Seco
Issue date:	02/04/1980
From:	Black R, Lewis S NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD)
To:	Atomic Safety and Licensing Board Panel
References
NUDOCS 8002120373
Download: ML19260D792 (9)

Text

{{#Wiki_filter:. I UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION 2/4/80 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of ) ) SACRAMENTO MuhICIPAL UTILITY ) Docket No. 50-312 (SP) DISTRICT ) ) (Rancho Seco Nuclear Generating ) Station) ) NRC STAFF'S RESPONSE TO LICENSEE'S MOTIONS FOR

SUMMARY

DISP 0SITI0t! I. INTRODUCTION On January 24, 1980 Licensee filed two motions for sumary disposition, one with respect to all of the contentions asserted by Intervenors Castro-Hursh and the other with respect to the California Energy Comission's Issue 5-2. As to Castro-Hursh, Licensee's motion is divided into parts. The first part asks for sumary disposition of all of the Castro-Hursh contentions (hereinaf ter " blanket motion") on the basis that the Castro-Hursh responses to interrogatories demon-strate that they have nc factual basis for their contentions. The second part asks for sumary disposition of eight specific contentions of Castro-Hursh on the ground that the affidavits attached to the Licensee's motion demonstrate that there is no genuine issue of material fact with respect to the matters raised in those contentions. In this response the NRC Staff takes the following position with respect to Licensee's motions: soosuo3.2 1952 128

1. We oppose the blanket motion; 2. We oppose summary disposition of Castro-Hursh Contentions 9, 21, and 22; and 3. We do not oppose summary disposition of Castro-Hursh Conte,tions 3, 5, 8, 20, and 25 and of CEC Issue 5-2. Attached to our response are: 1. NRC Staff's Statement of Material Facts As to Which There Exists An Issue to Be Heard; 2. The Affidavit of Dale F. Thatcher in support of our position on Castro-Hursh Contentien 9; 3. The Affidavit of Philip R. Matthews in support of our position on Castro-Hursh Contention 21; and 4. The Affidavit of Paul E. Norian in support of our position on Castro-Hursh Contention 22. II. ARGUMENT A. The Blanket Motion The NRC Staff opposes summary disposition of any of the Castro-Hursh contentions on the basis of the blanket motion. 1952 129

I The Commission's summary disposition rule (10 C.F.R. 52.749), which is modeled after Rule 56 of the Federal Rules of Civil Procedure, establishes a very exact-ing standard for the granting of a motion for surrmary disposition. Such a motion will only be granted if the filings in the proceeding, depositions, answers to interro-gatories, and admissions on file, together with the statements of the parties and the affidavits, if any, show that there is no genuine issue as to any material fact and that the moving party is entitled to a decision as a matter of law. 10 C.F.R. 52.749(d). The moving party has the " burden of showing the absence of a genuine issue as to any material fact." Cleveland Electric Illuminating Co. (Perry Nuclear Power Plant, Units 1 and 2), ALAB-443, 6 NRC 741, 753 (1977), citing Adickes v. Kess Co., 398 U.S. 144, 157 (1970). To meet this burden, there must be " annexed to the motion a separate, short and concise statement of the material facts as to which the moving party contends that there is no genuine issue to be heard." 10 C.F.R. 52.749(a). This statement of material facts to support the motion is "not merely a procedural technicality, but it is of sub-stantive significance." Pacific Gas & Electric Co. (Stanislaus Nuclear Project, Unit No. 1), LBP-77-45, 6 NRC 159, 163 (1977). The statement of material facts serves to 1) put the other parties on notice as to exactly what it is they must controvert in their statement of material facts as to which they contend a genuine 1/ issue exists and 2) enables the Board to go beyond the pleadings and determine, ~ based on depositions, answers to interrogatories, admissions, statements and affidavits of parties, whether there is a genuine issue as to any material fact bearing upon the contentions as to which summary disposition is sought. Stanislaus, supra, 6 NRC at 163. 1952 130 1] Pursuant to 10 C.F.R. 52.749(a), such a statement must be filed by a party opposing a motion for summary disposition. As its statement of inaterial facts supporting its blanket motion, Licensee merely states: Intervenors have admitted that there is no factual basis, or refusetodivulgethebasisifoneexists,foranygfthe contentions they have advanced in this proceeding. _/ Licensee cites to the Intervenors' serveral responses to interrogatories. The Licensee's statement identifies no specific facts as to which it asks the Licens-ing Board to make summary findings in its favor and as to which the other parties might respond. Rather, the Licensee is in the posture of asking the Board to grant summary disposition of all of the Castro-Hursh contentions on the basis of a legal argument alone. That argument relates to whether Intervenors have demon-strated bases for their contentons.-3/ The NRC Staff would agree with the Licensee that at this juncture in the proceed-ing the Intervenors have not ad'enced the proper support and foundation for these 4/ 5/ contentions. Indeed, the Staff sought and the Licensing Board issued an order compelling Castro-Hursh to respond to certain interrogatories in order to determine if they have the requisite support and basis for these contentions in the context of this proceeding. However, even though the Intervenors have failed to come for-ward with the requisite support and basis for certain of their contentions, the 2/ Licensee's Statement of Material Facts As to Which There Is No Genuine Issue to Be Heard (Castro-Hursh Contentions), at 1. -3/ Licensee's Brief in Support of its Motions for Summary Disposition of Contentions by Intervenors Gary Hursh and Richard Castro, at 7-15. 4/ NRC Staff Motion for an Order Compelling Discovery Against Intervenors Hursh- _ E- -q _~ Castro (December 17,197$; 5/ Order Relative to the NRC Staff's Motion to Compel Intervenors Gary Hursh and Richard Castro (January 21,1980). Pursuant to the Order, responses were to be filed on January 31, 1980. 1952 131 Licensee, as the moving party, is not relieved of its burden under 10 C.F.R. 52.749 to show initially the absence of a genuine issue concerning any material fact concerning those contentions. See, eg., Adickes v. Kess & Co., supra, 398 U.S. at 159. That is, we do not believe that the Licensee's assertions of lack of foundation with respect to these contentions meet its requirement under 10 C.F.R. 52.749(a) that a statement of material facts must be annexed to the motion to show that there is no genuine issue to be heard. Consistent with its assertion of legal grouno.; for the blanket motion, Licensee has not filed affidavits in support of its position. While not required to file affidavits (10 C.F.R. 52.749(a)), the moving party has the burden of establish-ing a record on which findings can be made. Stanislaus, supra, 6 NRC at 163; Perry, supra, 6 NRC at 752-754. This the Licensee has clearly failed to do. Accordingly, the Licensee's blanket motion must be denied for failure to meet the procedural and substantive requirements of 62.749. B. Motion As to Specific Castro-Hursh Contentions 1. Contention 9 This contention states: Rancho Seco, being a Babcock and Wilcox designed reactor, has not installed adequate hard-wire con-trol grade reactor trip on loss of main feedwater and/or on turbine trip, and therefore is unsafe ~ and endangers the health and safety of Petitioners, constituents of Petitioners and the public. .L _~ 1952 132

As noted in the attached Affidavit of Dale F. Thatcher, the Staff is presently unable to verify the two loss of feedwater transients relied upon, in part, in the supporting Affidavit of Robert A. Dieterich as demonstrating the reliability of the control grade anticipatory reactor trips (ARTS) installed at Rancho Seco. Information conveyed to the Staff at a August 23, 1979 meeting with B&W owners also raises a question as to the reliability of the ARTS. See Sumary of Meeting held on August 23, 1979, with Babcock & Wilcox (B&W) Operating Plant Licensees to Discuss Recent (Post TMI-2) Feedwater Transients (dated September 13,1979),a copy of which was sent to the service list in this proceeding. Thatcher Affidavit, para. 2. The determination of whether the ARTS installed in B&W operating plants are adequate is one requiring the application of judgment both as to the importance of the function performed and the reliability of the equipment to perform it. Thatcher Affidavit, para. 3. Since we are presently unable to verify an important basis of the Licensee's conclusion that the ARTS will function reliably and since we believe it to be a matter of judgment whether the ARTS are adequate, Intervenors Castro-Hursh are entitled to probe the judgment of the witnesses who will be offered on Contention 9. Furthermore, since the Board has independently raised the question of reliability of the ARTS (Additional Board Question 1), testimony will be received on this matter whether or not sumary disposition is granted with respect to Contention 9. For the above reasons, the motion for sumary disposition of Contention 9 should be denied. 2. Contention 21 1952 133 This contention states: Rancho Seco, being a Babcock and Wil5eEtiesigned reactor, i has a pressurizer tank and quench tank which are of in-adequate size to accomodate the volume of gas or liquid that may be required to be stored in the event of a loss of feedwater transient, and therefore is unsafe and endangers the health and safety of Petitioners, constituents of Petitioners and the public. As noted in the attached Affidavit of Philip R. Matthews, the Staff cannot fully support the statement in the Affidavit of Bruce A. Karrasch filed in support of Licensee's motion for summary disposition of this contention that the volume of the pressurizer results in it not emptying during a feedwater transient. Matthews Affidavit, para. 3; Karrasch Affidavit, para. 5. In the Staff's view, there is a potential for emptying the pressurizer or lowering the pressurizer level below the instrument range as a result of a feedwater transient that over-cools the reactor coolant system. Matthews Affidavit, para. 3. Literally interpreted, Contention 21 only raises the question of whether the pressurizer and pressurizer relief tank are large enough to accommodate the volume of gas or liquid that they may be required to store during a loss of feedwater transient. However, the B&W nuclear steam supply system design has been shown to be sensitive to both undercooling and overcooling events. The issues designated by the Conmission for consideration in this proceeding are broad enought to encompass overcooling, as well as undercooling, events. In both issues 1 and 3 set forth in its June 21, 1979 Order, the Commission referred to the ability of the facility to respond safely to "feedwater transients." The Licensing Board later stated that the scope of the proceeding" includes all matters and issues which hinge upon response to feedwater transients." Order Ruling on Scope and Contentions, at 3 (October 5, 1979). Although Castro-Hursh have only 1952 134

focused on undercooling events, since this was the type which occurred at TMI-2, the Staff believes that a genuine issue of fact remains as to whether pressurizer level will be drawn very low during overcooling events. In view of the Commission's and Licensing Board's statements as to the scope of this proceeding, we believe this issue to also be material. Because the motion of the Licensee with respect to Contention 21 does not dispose of the issue of pressurizer level drawdown during overcooling events, we believe that summary disposition of this contention should be denied. 3. Contention 22 This contention states: Rancho Seco, being a Babcock and Wilcox designed reactor, does not provide control room operators with sufficient data on the water level in the pressurizer and vessel because the operators must interpret information on temperature and pressure in the primary loop and extrapo-late water level, and therefore is unsafe and endangers the health and safety of Petitioners, constituents of Petitioners and the public. The Affidavit of R. J. Rodriguez filed in support of Licensee's motion states that:

1) "as long as reactor coolant temperature and pressure are. maintained in a subcooled condition, pressurizer level (which is available directly to operators in the. control room) is a direct and adequate indicator of system water level" and

2) "during the current outage a subcooling. indicator will be installed which, together with the pressurizer level reading, will eliminate any need for interpretation of temperature and pressure data."

Rodriguez Affidavit, para. 3. we :- d52 135. As noted in the attached Affidavit of Paul E. Norian, the subcooling indicator will not, however, provide a direct indication of vessel water level. Norian Affidavit, para. 5. The NRC Staff has, however, required the Licensee (and all other PWR licensees) to analyze the need for instrumentation to measure vessel water level. If such instrumentation is required it should be installed by January 1, 1981. NUREG-0578, Table B-1, Section 2.1.3.b. Because the Affidavit of Mr. Rodriguez does not address the question of indication cf vessel water level in other than subcooled situations, it 13 insufficient to support summary disposition of Castro-Hursh Contention 22. IV. CONCLUSION For the reasons set forth in this brief the Licensing Board should deny Licensee's blanket motion for summary disposition of the Castro-Hursh contentions. The Licensing Board should also deny. Licensee's motion with respect to Castro-Hursh Contentions 9, 21, and 22. We do not oppose summary disposition of the remaining contentions which are the subject of Licensee's motions. Respectfully submitted, Stephen 1. Lewis Counsel for NRC Staff n citCf c Richard L. Black Counsel for NRC Staff Dated at Bethesda, Maryland this 4th day of February, 1980. 1952 136

NRC STAFF'S STATEMENT OF MATERIAL FACTS AS TO WHICH THERE EXISTS AN ISSUE TO BE HEARD 1. (Castro-Hursh Contention 9) Performance of the control grade anticipatory reactor trips (ARTS) installed at operating B&W plants (including Rancho Seco) raises a question as to whether the ARTS are sufficiently reliable. Thatcher Affidavit. 2. (Castro-Hursh Contention 21) The Rancho Seco pressurizer has the potential for being shown extremely low or emptying as a result of a feedwater transient that overcools the reactor coolant sy' stem. Matthews Affidavit. 3. (Castro-Hursh Contention 22) During transient conditions, such as following loss of feedwater, void forma- . tion may occur and pressurizer level may not provide an adequate indication of system inventory. Norian Affidavit. 1952 137 . a y12 by 20020307"3

UNITED STATES OF AMERICA NUCLEAR REGULATORY C0ffilSSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of SACRAMENTO MUNICIPAL UTILITY ) DocketNo.50-312(SP) DISTRICT (RanchoSecoNuclearGanerating ) Station) ) AFFIDAVIT OF DALE F. THATCHER I. Dale F. Thatcher, depose and state under oath as follows: 1. I am a reactor engineer in the the Nuclear Regulatory Comission Staff's Instrumentation aiid Control Systems Branch. I was responsible for the review and evaluation of the instrumentation and control systems as part of the Bulletins and Orders Task Force. My professional qualifications are attached hereto. 2. I have reviewed the Affidavit of Robert A. Dieterich filed by the Licensee in support of its motion for sumary disposition of Castro-Hursh Contention 9. Among other things, Mr. Dieterich states that the control grade anticipatory reactor trips (ARTS) circuitry "has operated successfully in two loss-of-feedwater transients... at Rancho Seco." Dieterich Affidavit, para. 5. I have been unable to verify the transients referred to by Mr. Dieterich. I also note that the Licensing Board has focused upon the question of whether the control grade trips which have been installed at Rancho Seco are adequately reliable. Additional Board Ouestion 1. The Board's question was based upon the reported fact that through August 23, 1979, the ARTS installed 1952 138 oug 20020703M at all B&W operating plants had been challenged four times and had failed once. See Summary of Meeting Held on August 23, 1979, with Babcock & Wilecx (B&W) Operating Plant Licensees to Discuss Recent (Post TMI-2) Feedwater Transients (dated September 13,1979). 3. The question of the adequacy of the ARTS is one requiring the application of judgment both as to the relative importance of the function to be performed and the reliability of the design to perform this function. In view of the information raised at the August 23, 1973 meeting and my inability to verify the two loss-of-feedwater transients at Rancho Seco relied upon, in part, by Mr. Dieterich as demonstrating the t-eliability of the ARTS, I believe some question remains as to the adequacy of the ARTS. I hereby certify that the answers given by me are true and accurate to the best of my knowledge. -{ f,, M-< h-/ - Dale F. Thatcher Subscribed and sworn to before me this 4th day of February,1980. ~. f .r t,u.,j7! f... n'/ W L-ti i n /) Notary Public Hy Coci6ilssion Expires: July 1, 1982 1952 139

DALE F. THATCHER PROFESSIONAL QUALIFICATIONS. INSTRUMENTATIJN & CONTROL SYSTEMS BRANCH DIVISION OF SYSTEMS SAFETY I am a Senior Reactor Engineer in the Instrumentation and Control Systems Branch, Division of Systems Safety, Nuclear Regulatory Comission. From May to December 1979, I was assigned to the Bulletins and Orders Task Force as a technical reviewer in the area of instrumentation and control. Just prior to this assignment I was a member of the NRR team which aided in the Three Mile Island Recovery Operation. In the ICSB, my primary responsibility is to perform technical reviews of the design, fabr' cation, and operation of instrumentation and control systems for nue.9ar power plants. This review encompasses evaluation of applicant's safety ar.alysis reports, generic reports and other related information on the instrumentation and control designs. I graduated from Lehigh University with a Bachelor of Science Degree in Electrical Engineering in June 1971. From my graduation in June 1971 until my employment at the Comission, I was an Instrumentation Engineer with Gilbert Associates Inc., an Architect-Engineering company located in Reading, Pennsylvania. My responsibilities included the design and evaluation of various instrumentation and control systems including primarily the areas of reactor protection systems and other safety systems for various domestic nuclear power plants. I joined the Regulatory staff of the Atomic Energy Commission in March 1974 as a Reactor Engineer. Since then, I have participated in the review of instrumentation control and electrical systems of numerous nuclear power stations and standard plant designs. In addition, I have participated in the formulation of related standards and regulatory guides. I am a member of the Instituto of Electrical and Electronics Engineers (IEEE) and have participated in the development of IEEE Standard 379-1977, "IEEE Standard Application of the Single Failure Criterion to Nuclear Power Generating Station Class IE Systems" and other proposed standards. 1952 140

UNITED STATES OF AMERICA NUCLEAR REGULATORY COR11SSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of SACRAMENTO M' NICIPAL UTILITY ) Docket No. 50-312 (SP) J DISTRICT ) ) (Rancho Seco Nuclear Generating ) Station) ) AFFIDAVIT OF PHILIP R. MATTHEWS I, Philip R. Matthews, depose and state under oath as follows: 1. I am a Section Leader in the Nuclear Regulatory Comission Staff's Auxi-liary Systems Branch. I am currently responsible for the review and evaluation of nuclear plant auxiliary systems. My professional qualifi-caticns are attached hereto. 2. I have reviewed the affidavit of Bruce A. Karrasch, submitted by the licensee in support of its motion for sumary disposition of Castro-Hursh Contention 21. Mr. Karrasch states that the pressurizer " volume results in the pressurizer neither filling nor emptying during an anticipated loss of feedwater transient." Para. 5. 3. In my opinion, there is a potential for emptying the pressurizer or lowering the pressurizer level below the instrument range as a result of a feed-water transient that overcools the reactor coolant system. Mr. Karrasch's affidavit does not adequately address overcooling events, such as a reactor 1952 141 baF Sc;c 2 o 903Gt trip coindicent with a system failure or operator error which results in high feedwater flow. The staff concern with overcooling events is discussed in NRC Memorandum dated November 16, 1979 from D. Eisenhut to S. Scott,

Subject:

" Rancho Seco Board Notification 10 C.F.R. 50.54 Re-quest Regarding Design Adequacy of Babcock e d Wilcox NSSS." 4. Some transients in B&W reactors occurring sin:e the TMI-2 accident, as reported in Attachments A and B hereto, have resulted in overcooling the reactor coolant system and lowering the pressurizer level below the instrument range. I hereby certify that the answers given by me are true and accurate to the best of my knowledge. A d/1 JW @ilip R. Nattfiew's' ' ~ Subscribed and sworn to before me this 4th day ci February,1980. j/ s'7 - (MW n h J/t. F " ' ' L't A' ij

" Notary Public My Comission expires

July 1, 1932.

1952 142

F\\ Attachment A UNIT TRIP REPORT Or.- a DATE OF TRIP: Septenber 18, 1979 FACILITY: Davis-Besse Unit 1 TDENTIFICATION OF OCCURRENCE: Turbine / Reactor Trip due to electro-hydraulic control (EHC) pressure transient The unit was in Mode 1, with Power = 2765 WT, and Load = 915 INITIAL CONDITIONS: We DESCRIPTION OF OCCURRENCE: At 12:43:01 hours on September 18, 1979,' Number 2 EBC 5193.15, Pump automatic backup function was tested during the performance of PT EHC Hydrhulic Power Unit Periodic Test". A dip in EEC pressure resulted which caused a turbine trip and subsequent reactor; trip from ARTS (Anticipatory Reactor Reactor Coolant System (RCS) temperature and pres-01 hours. Trip System) at 12:43: sure dropped and the Integrated Control System (ICS) reduced the feedvater flow. 1, 2, 3, 4, 5, 8, and 9 on the east header and 1, 2, 3, 4, Main steam line safeties 5, 7, 8, and 9 on the vest header lif ted. The operator started the. second makeup pump at 12:43:27 hours in an attempt to stop the decreasing pressurizer level. 12:43:52 hours The pressurizer level indication dropped off scale from approximately to 12:44:43 hoars mainly due to an extended safety Walve' blevdown on the Steam Generator (SG) 1-2 main steam line causing a reduction in RCS average temperature and a density increase of the RCS vater. An analysis of the data indicates the pressurizer had at least 50 inches of water remaining, but the lowest level sensed by the level indicator is approximately 75 inches above the bottom of the pressurizer. The reduction of the pressurizer level expanded the steam' space and reduced RCS pres-sure to a lov of approximately 1710 psig. When RCS pressure dropped to the Reactor - Protection System (RPS) low pressure temperature setpoint (1985 psig) all four RPS channels actuated as designed but the reactor had already been tripped by ARTS. the turbine bypass valves were maintaining the Within three minutes of the event, SG outlet pressure at approximately 995 psig, feedvater flovrate had stabilized, Within six minutes of the event, and pressurizer level indication was returning. RCS pressure was returned to within normal operating range. When the second EBC pump was started, an instantantous pertur-CAUSE OF OCCURRENCE: bation was induced in the EBC system pressure, and PSL 2320 tripped the turbine on If the EHC system was operational properly, the starting of a lov EHC pressure. At the time this report second EEC pu=p would not have caused a pressure transient. was prepared, it is believed the instantaneous perturbation in pressure was caused by a sticking pump pressure controller on No. 2 EHC pump which is not'. properly con-trolling pressure during a pressure transient. The cause of this occurrehee is id'entical to the cause of the Octobey 3, 1978 trip. ~ Y MI.Tr-T&., bu.g 195_2 -l 43 em 2 o3%

• 8*

=e e ee e

3 REV. 1 - 9/25/79 UNIT TRIP R/ PORT September 18, 1979 Page 2 The pressure controller was repaired and successfully; tested af ter the October 1978 occurrence, and the problem was believed to have been resolved. The cause of the extended blowdown on main steam line 1-2 is believed to have been 1 caused by a recently installed (March, 1979) safety valve SPl7A2. The valve setpoint was factory set and tested but the blevdown ring was not adjusted to where past experi-ence with these valves at Davis-Besse has shown is required to limit blowdown. CORRECTIVE ACTION: Periodic Test PT 5193.15 vill not be performed until the problem with the pressure controller is corrected. The Number 1 ESC punp was placed in standby since testing verified the Number 1 automatic start functions correctly. The Nu=ber 2 EHC pump and the pump pressure controller vill be disassembled and cleaned at the earliest possible date. The cleaning vill also be performed on 1l Number 1 EEC string if it cannot be verified the problem vill not develop. General Electric personnel have suggested that the setpoint of the pressure switch ~ which trips the turbine (PSL 2320) be decreased from 1100 psig to 1000 psig. They have also suggested an orifice be installed in the source line to the pressure switch to prevent the actuation of the switch from an instantaneous pressure surge. An FCR is being prepared to request the incorporation of these General Electric suggestions. On September 23, 1979, the blowdown on SP17A2 was adjusted under Haintenance Work Order 79-3029. The maintenance procedure for installation of safety valves yill be 1 modified to include a verification of blevdown rings to the setpoint established by station experience. TRANSIENT CIASSITICATION: This RCS transient was c1assified as an BD transient by station personnel. DATA PACKAGE: 1. Alarm Printout 2. Post Trip Review and Sequence of Events 3. Reactineter Printouts and Graphs 4. Control Room Charts (later) ~ 5. ICS Response Strip Charts (master only) d 1952 144

N Attachment B UNIT TRIP REPORT DATE OF TRIP: September 26, 1979 FACILITY: Davis-Jesse Unit 1 IDENTIFICATION OF OCCURRENCE: Turbine / reactor trip due to electro-hydraulic control (EHC) throttle pressure limiter failure INITIAL CONDITIONS: The unit was in Mode 1, with Power = 2772 WT, and Load = 914 We. DESCRIPTION OF OCCURRENCE: At 20:56:33 hours on September 26, 1979, a high pressure turbine throttle pressure limit alarm was received. The throttle pressure limiter is used to protect against an excessive decrease of steam pressure when steam generation of the NSSS falls below the steam demand of the turbine. It acts directly to close the high pressure turbine control valves in an effort to maintain steam header pres-The control valves going shut rapidly-caused a mismatch between heat generation sure. and heat removal with a resulting increase in Reactor Coolant System (RCS) temperature and pressure. Seven seconds later at 20:56:40, the reactor tripped on RCS high pres-sure followed by a turbine trip. RCS temperature and pressure then dropped and the Integrated Control System (ICS) redu ed the feedwater' flow abruptly. The second makeup pump was started at 20:57:03 to counter the decrease in pressurizer level. a s e.. ;.* Reactor Coolanc System Tave decreased to' 548.50F at approximately 57 seconds af ter the ~ trip. The pressurizer level indication dropped off scale at 12:57:21 and remained below the indicating range for approximately 21 seconds. An analysis of the data indicates the pressurizer level decreased to -10 inches below the lower level sensing tap, however, there remained approximately 65 inches of water above the bottom of the pressurizer. The reduction in pressurizer level can be attributed to the decrease in RCS average ~ temperature and the resultant contraction of the reactor coolant. The reduction in pressurizer level subsequently reduced RCS pressure to a low of approximately 1686 psig. A channel 2 Safety Features Actuat! ion System (SFAS) low pressure trip was re-ceived at 20:57:24. - rm.. '. u s: N - .wa. ...~ Within two minutes of the trip, the turbine bypass valves were maintaining steam gen-erator outlet pressure at approximately 995 psig and RCS Tave at a nominal 550 F. By this time feedwater flows had stabilized, allowing OTSGs to slowly steam down, reaching their low level limits approximately 15 minutes af ter the trip. Also, by this time, pressurizer level indication had returned with the level being restored to above the low level pressurizer heater cutoff (40 inches) within 3.5 minutes of the event. Within seven minutes of the event, RCS pressure was returned to the normal operating. value of 2155 psig. xJ5

s it CAUSE OF OCCURRENCE: The cause of the trip was attributed to the failure of the high pressure turbine throttle pressure transmitter power supply.

The failure had the effect of sensing a low steam gupply pressure causing the turbine control valves to 1952 145 W g w w (N M

GNIT TRIP REPORT SEPTEMBER 26, 1979 PAGE 2 shut in an effort to maintain throttle pressure. An investigation of tue transmitter revealed that one capacitor on the power supply circuit board had fallen completely off and another was loose. A similar problem was identified on August 22, 1979 when the unit experienced a 200 MWe load reduction due to an erroneous throttle pressure limiter signal. In this event, loose capacitors were determined to be the cause and the affected circuit board was repaired. The turbine vendor, General Electric, is aware of reliability problems with the Rosemount pressure transmitter, in particular the problem of defective circuit boards. It will supply an Electro-Sen pressure ~ transmitter as a replacement. Until the present Rosemount pressure transmitter is replaced, the throttle pressure limiter signal vill be jumpered out of the turbine load control circuitry. ANALYSIS OF OCCURRENCE: The loss of pressurizer level is attributed to low RCS temp-erature caused by excessive steam safety valve blowdown and by some overfeeding of the OTSG. In a similar event on September 18, 1979, the unit experienced a turbine-reactor anticipatory trip also from a nominal 100% FP. During this occurrence an OTSG 1-2 safety valve blew back to 939 psig causing RCS Tave to decrease to 546.30F. This lowered pressurzer level to approximately -30 inches and RCS pressure to 1709 psig. The blowdown on this valve was adjusted per the vendor, Dresser Valve, and B&W recommendations on September 23, 1979. The September 26 event resulted in a OTSG 0 1-2 safety valve blowback to 954 psig causing RCS Tave to decrease to 548.5 F. Pres-surizer level was lowered to -10 inches however RCS pressure decreased to 1689 psig. The lower RCS pressure can be attributed to three factors: 1. Prior to the September 18 event, all the heaters were on and the pres-surizer was being sprayed down to prevent a buildup of boron in the pressurizer. Having all the heaters on served to mitigate the pressure decrease during the transient. t 2. There was no power mismatch, hence no RCS temperature vi~ and subsequent pressurizer insurge, associatdd with the September 18 event. However, there was an insurge during the September 26 event causing cooler RCS water to disturb the equilibrium conditions in the pressurizer and leading to a lower RCS pressure. 3. Another factor is that there was little or no feedwater kicker present during he September 18 event, where there was one present during the Septec ser 26 event. The feedwater kicker circuit provides a proportional boost to feedwater pump demand on a high steam header pressure. It is an anticipatory feature to miqlpize the underfeeding and resultant high RCS temperature and pressures initially encountered during a runback. On a reactor trip the steam header setpoint is adjusted from 870 psig to 995 psig. Therefore, on a reactor trip the signal from the kicker will be pro-portionally less than during a load rejection. During the September 26 event, the turbine shed load for seven seconds prior to the trip, permitting a much more pronounced feedwater kicker to be observed. 1952 146

Philip R. Matthews Professional Qualifications I am employed by the U.S. Nuclear Regulatory Commission as a Section Leader in the Auxiliary Systems Branch, Division of Systems Safety, Office of Nuclear Reactor Regulation. I am responsible for supervision of techni-cal personnel engaged in analysis and safety evaluation of nuclear power plant auxiliary systems including the main steam and feedwater, auxiliary feedwater, component cooling water, service water, new and spent fuel storage and handling, plant ventilating and air conditioning, and fire protection systems. I attended the University of California, Berkely California and received a. Bachelor of Science degree in Chemistry in 1947. Subsequently, I have completed several graduate courses in mechanical and nuclear engineering. In 1947, I commenced work at the Knolls Atomic Power Laboratory, General Electric Co., Schenectady, N.Y. I worked there until 1968 on various naval nuclear submarine and surface ship propulsion power plant projects. I had technical and management responsibility for nuclear plant nchani-cal and fluid systems design, testing, performance evaluation, proto-type and shipboard reactor plant start-up and sea trials. In 1968, I transferred to the General Electric Co., Nuclear Energy Division in San Jose, California. I was Quality Assurance tranager for the Atomic Power Equipment Department responsible for quality assurance of APED purchased engineered equipment and installation of APED equipment at BWR nuclear plant sites. I joined the Nuclear Regulatory Commission in 1973 as a nuclear engineer en the Office of Standards. In 1975, I assumed my present duties as Section Leader in the Auxiliary Systems Branch. In this position, I have had two major special assignments; namely,1) to direct the technical preparation, issuance and plant specific implementation review of nuclear pitnt fire protection guidelines following tne 1975 fire at Browns Ferry Nuclear Plant and 2) in 1979, to direct a Task Force in reviewing the design and operation of Auxiliary Feedwater Systems of operating nuclear plants with Westinghouse and Combustion Engineering designed reactors and provide specific recommendations for ingroving Auxiliary Feedwater System reliability. 1952 147

UNITED STATES OF AMERICA ~ NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of ) ) SACRAMENTO MUNICIPAL UTILITY ) Docket No. 50-312 (SP) DISTRICT (Rancho Seco Nuclear Generating ) Station) ) AFFIDAVIT OF PAUL E. NORIAN I, Paul E. Norian, depose and state under oath as follows: 1. I am a Sectioa Leader in the Analysis Branch, Division of Systems Safety, NRC. I served as the Alternate Group Leader of the Analysis Group, Bulletins and Orders Task Force. I coordinated the reviews of small break loss-of-coolant accidents (LOCA) and transient analyses submitted by vendor owners' groups since the Three Mile Island Accident. My professional quali-fications are attached hereto. 2. I have reviewed the affidavit of R. J. Rodriguez, submitted by the licensee in support of its motion for summary disposition of Castro-Hursh Contention 22. Mr. Rodriguez states that "as long as reactor coolant temperature and pressure are maintained in a subcooled condition, pressurizer level (which is avail-able directly to the operators in the control room) is a direct and ^uate indicator of system water level." Para. 3. I essentially agree with this statement. 3. During transient conditions, however, such as following loss of feedwater or turbine trip with subsequent reactor trip, void formation.may occur in 19'52 148 b a>A 9c0207D3))

the reactor pressure vessel. See NRC Staff's Response to California Energy Commission's Interrogatory 5, Second Set. If the primary system contains significant voids, pressurizer level may not provide an adequate indication of system inventory. 4. In view of this limitation on the pressurizer level as an indicator of whether adequate core cooling is taking place the Staff has required pressurized water reactor (PWR) licensees to undertake a two-phase program: a. install coolant saturation meters (which can essentially be done with existing plant instrumentation); and b. study and develop system modifications which can provide a direct indication of vessel water level. NUREG-0578, Section 2.1.3.b, elaborated upon at pp. A-9 to A-10. 5. Mr. Rodriguez states that the subcooling indicator will be installed at Rancho Seco during the current outage. Rodriguez Affidavit, para. 3. While this instrumentation will provide a direct indication of subcooling in the core, it will not provide a direct indication of vessel water level. We expect that the Licensee's analysis in response to item 4b, above, will indicate the need to install additional instrumentation to indicate vessel water level. If so, such instrumentation should be installed by January 1, 1981. NUREG-0578, Table B-1, Section 2.1.3.b. 1952 149

I hereby certify that the answers given by me are true and accurate to the best of my knowledge, v2 / Paul E. Norian Subscribed and sworn to before me this 4th day of February, 1980. M f tiotary Public / My Commission Expires: July 1, 1982. Ih52150

PAUL E. NORIAN PROFESSIONAL QUALIFICATIONS I =.m Section Leader of the Systems Analysis Section, Analysis Branch, Division of Systems Safety. I have held this position since 1975 and am responsible for supervising the review of reactor vendor transient and LOCA analysis methods, the improvement of NRC analysis methods used in related accident analyses, and the performance of staff audit calculations for transients and LOCAs. From June through December 1979, I was assigned to the Bulletins and Orders Task Force as a member of the Analysis Group. I served as Alternate Group Leader and coordinated the reviews of small break loss-of-coolant accidents (LOCA) and transient analyses submitted by the vendor owner's groups since the Three Mile Island accident. I graduated from Lehigh University in June 1955 with a Bachelor of Science Degree in Engineering Physics. I also attended Drexel Institute of Technoiogy, Catholic University of America, and the University of Maryland where I have taken various graduate courses in mathematics, physics, and electrical engineering. In July 1955, I began work as a physicist with the duPont Company at the Savannah River Plant in Aiken, South Carolina. From that time until March 1962, I worked in the Works Technical Department on operational physics problems associated with the heavy water production reactors at Savannah River. This work included such assignments as the development of monitoring systems, performance of physics calculations required in reactor operation and in the development of new fuel elements, the review of operating procedures, and the analysis of various operating problems. In March 1962, I was transferred to the duPont Company's Chestnut Run Laboratories in Wilmington, Delaware, and worked for its Film Department on the development of industrial applications for plastic films. In December 1963, I accepted a position with the Division of Reactor Licensing of the U.S. Atomic Energy Comission, and was project leader in the construction permit review of Consolidated Edison's Indian Point No. 2 reactor and Wisconsin. Michigan's Point Beach No. 1 reactor. I was assigned as a nuclear engineer in the Systems Performance Branch of the Division of Reactor Standards in March 1967. My responsibilities included analyzing and evaluating the performance of engineered safety systems and performing computer calculations fer the evaluation of contain-ment response and loss-of-coolant accidents. In Martn 1971, I participated in the Regulatory Task Force reappraisal of emergency core cooling systems for light water reactors. My main responsibility for the task force was the review of computer codes and input assumptions for LOCA analyses. In May 1973, I was assigned to the Core Performance Branch in the Directorate of Licensing. I served as Section Leader in the Thermal Hydraulics Section and supervised the review of portions of reactor vendor model changes to confonn with the new requirements for LOCA models specified in Appendix K to 10 CFR Part.50. 1952 \\5\\

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of SACRAMENTO MUNICIPAL UTILITY Docket No. 50-312 (SP) DISTRICT Rancho Seco NJclear Generating Station CERTIFICATE OF SERVICE I hereby certify that copies of "NRC STAFF'S RESPONSE TO LICENSEE'S MOTIONS FOR

SUMMARY

DISPOSITION" in the above-captioned proceeding have been served on the following by deposit in the United States mail, first class, or, as indicated by an asterisk, through deposit.in the Nuclear Regulatory Comission's internal mail system, this 4th day of February, 1980:

• Elizabeth S. Bowers, Esq., Chairman Atomic Safety and Licensing Board Panel Gary Hursh, Esq.

U.S. Nuclear Regulatory Comission 520 Capitol Mall Washington, D.C. 20555 Suite 700 Sacramento, California 95814

• Dr. Richard F. Cole Atomic Safety and Licensing Board Panel Mr. Richard D. Castro U.S. Nuclear Regulatory Comission 2231 K Street Washington, D.C.

20555 Sacramento, California 95816

• Mr. Frederick J. Shon James S. Reed, Esq.

Atomic Safety and Licensing Board Panel Michael H. Remy, Esq. U.S. Nuclear Regulatory Comission Reed, Samuel & Remy Washington, D.C. 20555 717 K Street, Suite 405 Sacramento, Califorriia 95814 David S. Kaplan, Esq. General Counsel Christopher Ellison, Esq. Sacramento Municipal Utility District Dian Grueneich, Esq. P. O. Box 15830 California Energy Comission Sacramento, California 95813 1111 Howe Avenue Sacramento, California 95825 1952 152

L._. .m '

• Atomic Safety and Licensing Mr. Michael R. Eaton Board Panel Energy Issues Coordinator U.S. Nuclear Regulatory Comission Sierra Club Legislative Office Washington, D.C.

20555 1107 9 Street, Room 1020 Sacramento, California 95814

• Atomic Safety and Licensing Appeal Board Panel Thomas A. Baxter, Esq.

U.S. Nuclear Regulatory Comission Shaw, Pittman, Potts & Trowbridge Washington, D.C. 20555 1800 M Street, N.W.

• Docketing and Service Section Office of the Secretary U.S. Nuclear Reguhtory Comission Washington, D.C. 20565 Herbert H. Brown, Esq.

Lawrence Coe Lanpher, Esq. Hill, Christopher and Phillips, P.C. 1900 M Street, N.W. Washington, D.C. 20036 Stennen s. Lewis Counsel for NRC Staff \\]Y}}