B12288, Forwards Summary of Safety Impact Model Project Analysis for Isap Topic 1.21, Reg Guide 1.97 Instrumentation

Forwards Summary of Safety Impact Model Project Analysis for Isap Topic 1.21, Reg Guide 1.97 Instrumentation

ML20211C986
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	10/10/1986
From:	Opeka J, Sears C CONNECTICUT YANKEE ATOMIC POWER CO.
To:	Charemagne Grimes Office of Nuclear Reactor Regulation
References
RTR-REGGD-01.097, RTR-REGGD-1.097 B12288, NUDOCS 8610220069
Download: ML20211C986 (15)

Text

I CONNECTICUT YANKEE ATOMIC POWER COMPANY B E R L I N, C O N N E CTIC U T P o. BOX 270 HARTFORD. CONNECTICUT 06141-0270 T ELE PHONE October 10,1986 Docket No. 50-213 B12288 Office of Nuclear Reactor Regulation Attn: Mr. Christopher I. Grimes, Director Integrated Safety Assessment Project Directorate Division of PWR Licensing - B U.S. Nuclear Regulatory Commission ,

Washington, D.C. 20535 Gentlemen:

Haddam Neck Plant Integrated Safety Assessment Program Summaries of Public Safety impact Model Project Analyses In a letter dated July 31,1985,(I) the NRC outlined the scope of issues to be evaluated Neck Plant. inSubsequently,in the Integrated Safety Assessment a letter Program dated February (ISAP))for 14, 1986,(2 wethe Haddam identified a selected number 'of topics for which we would provide the Staff with public safety risk oriented analyses.

In order to facilitate the Staff review of our project analyses, we are providing the Staff, in Attachment 1, with a summary of the following project we have evaluated for public safety impact:

ISAP Topic No.1.21 " Regulatory Guide 1.97 Instrumentation" It is noted that since we have not completed our analyses of the entire set of ISAP projects, the public safety impact scores are to be considered preliminary at this time. Upon completion of our analyses of the entire ISAP project set, including all five attributes, we will review our analyses and revise our public safety impact results, if necessary, to assure consistency in the ranking of the ISAP projects.

(1) H. L. Thompson letter to J. F. Opeka, " Integrated Safety Assessment Program," July 31,1985.

(2) 3. F. Opeka letter to C. I. Grimes, " Integrated Safety Assessment Program Schedule for Implementation," dated February 14,1986.

00 8610220069 861010 i

hp PDR ADOCK 05000213 P PDR

. If you have any questions on this material, please feel free to contact my staff.

Very truly yours, CONNECTICUT YANKEE ATOMIC POWER COMPANY -

1iS.D

3. F. Opeka '

Senior Vice President 4 duob By: C. F. Sears Vice President

ISAP #1.21 Regulntory Guida 197 Instnauentition Safety Issue

-During an accident, it may be difficult for the operator to determine immediately what accident is occurring and therefore to determine the appropriate response. For this reason, reactor trip and.certain other safety actions (e.g., emergency core cooling actuation, containment isolation) have been designed to perform automatically during the initial stages of an accident. Instruraentation is also provided to indicate information about plant variables required to enable the operation of manually initiated safety systems and other appropriate operator actions. Adequate instrumentation to monitor plant variables and systems during and following an accident is necessary to minimize the risk to the public from operation of nuclear power plants.

Proposed Project Regulatory Guide 1.97, Revision 2, " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," was issued in December, 1980 (Reference 1) to provide guidance concerning implementation of the instrumentation mentioned above. The proposed project at the Haddam Neck plant (Connecticut Yankee (CY)) involves the evaluation of all instrumentation designated for evaluation in ISAP as listed in Attachment No.1 of Reference 2, as well as the additional concerns specified in Reference 3. Each type of instrumentation is to be evaluated for one or more of the following: installation, instrument range determination, environmental qualification, seismic qualification, and redundancy. A summary of this instrumentation is given below:

1. Flow in High Pressure Safety Injection System

2. Main Steam Flow from Safety Relief Valves (Steam Generator Safety Relief Valve Status)

3. Radiation monitor for Vent from Steam Generator Safety Relief Valves or Atmospheric Dump Valve

4. Steam Generator Level

5. RCS T eold CONNECTICUT YAfEEE INTEGRATED SAFETY ASSESSMEf6 PROGRAM

6. Subcooled' Margin Monitor

7. Core Exit Temperature B. Containment Hydrogen Monitor

9. Reactor Vessel Level Monitoring System

10. Steam Generator Pressure

11. RCS Makeup Flow

12. Containment Pressure 13, Neutron Flux

14. Containment Isolation Valve Position

15. Low Pressure Safety. Injection Flow Indication

16. Containment Atmosphere Temperature

17. Emergency Ventilation Damper Position

18. Containment Fan-Cooler Heat Removal Analysis of Public Safety Impact All of the instrumentation from the above list are analyzed individually in terms of safety in the following sections.

1. Flow in High Pressure Safety Injection System The safety significance of High Pressure Safety Injection (HPSI) System flow indication is that it might assist the operator in the event that HPSI flow needs to be regulated. Typically, HPSI System operation is at maximum flow. There is no capability to " throttle" the flow except for opening and closing MOV's not designed for such purposes. The CY Probabilistic Safety Study (PSS) (Reference 4) identified two potential uses of regulated HPSI flow. The first is following bleed and feed operation, where the charging system is unavailable thereby necessitating HPSI pump operation for the feed portion. The second is concerned with RCS make-up following RCP seal failure, again with charging unavailable. In both of these cases, the existence of HPSI flow can be verified through HPSI pump current indication CONNECTICUT YANKEE IEEGRATED SAFETY ASSESSMEE PROGRAM

and HPSI discharge valve position indication. There is only one dominant sequence from the CY PSS in which charging is unavailable and HPSI is required to_ perform either of the above functions. This sequence is initiated by the loss of motor control center 5 (MCC-5). MCC-5 is subsequently recovered, but primary integrity is lost due to a loss of RCP seal cooling (i.e., the loss of both charging and component cooling water). HPSI flow would then be required for RCS make-up. Even in this case, since HPSI flow can be verified using pump current and valve position indicators and primary system volume can be determined using pressurizer level indication and the Reactor Vessel Level Monitoring System, actual HPSI flow indication is of no real additional benefit. However, for the purpose of quantifying a maximum potential benefit from the implementation of HPSI' flow indication, it is assumed that the human error probability associated with the start and operation of the HPSI pumps can be reduced from 3E-03 to 1E-03. Requantifying this' sequence with the reduced human error contribution results in a maximum core melt frequency reduction of 1.3E-07/yr. This frequency reduction occurs in release category "SLC".

From the Public Safety Impact Model the corresponding decrease in public risk is 0.007 man-rem over the remaining lifetime of the plant.

2. Main Steam Flow from Safety Relief Valves (Safety Relier Valve Status)

The proposed modification involves the implementation of indication to monitor main steam flow from the steam generator (SG) safety relief valves (SRV's) in order to determine SG SRV status. The safety significance of this instrumentation would be to assist the operator in identifying a stuck open SRV following a Steam Generator Tube Rupture (SGTR). A benefit to safety from this indication can potentially be realized in two ways:

reduction in core melt frequency due to a higher reliability of identifying a stuck open SRV following a SGTR, and reduction in direct radiological release to the public due to a decrease in expected time to isolate a stuck open SRV following a SGTR. Both of these effects are evaluated below:

a. Core melt frequency (CMF)

CONfECTICUT YAfEEE ItEEGRATED SAFETY ASSESSMEfff PROGRAM

Th2 pot;ntial reduction in CMF from the proposed instrumentation would occur through an increase in the reliability of an operator to identify and mitigate the consequences of a stuck open SG SRV following an SGTR prior to draining the refueling water storage tank (RWST). Identification of a stuck open SRV is comprised of two parts. First is knowledge of an SRV being open, and second is a check of SG pressure to ascertain that SG pressure is below the SRV reseat point. The proposed instrumentation serves only to assist operator recognition of the fact that an SRV is open. It does not, in itself, indicate that an SRV is stuck open. Currently, knowledge of an SRV being open would come from audible indication in the control room, since an SRV makes a tremendous noise when it is open. Such a noise, especially in the time frame considered (i.e. the operator has a

, minimum of 100 minutes to take action assuming five failed tubes),

would be quite recognizable to the operators and plant personnel.

Also, even if control room personnel were to somehow be unaware of an SRV opening, shortly after recognition of the SGTR, an auxiliary operator would be sent to the Terry Turbine room to isolate steam flow from the faulted SG. Due to the resulting noise inside the Terry Turbine room, it would be virtually impossible for the auxiliary operator not to be cognizant of the fact that an SG SRV is open.

Therefore direct indication of SRV status, or steam flow, would not noticeably increase operator reliability in identifying that an SG SRV is stuck open. As such, this modification would produce no significant reduction in core melt frequency.

b. Direct radiological release Following a SGTR, a stuck open SG SRV will provide a direct radiological release path to the environment. Implementation of the proposed instrumentation might reduce this release by decreasing the expected time for an operator to identify the stuck open SRV and close the associated loop isolation valves (LIV's). From the previous discussion, an operator would determine that an SRV is stuck open based on awareness that an SRV is open combined with the recognition that SG pressures are all below the SRV reseat point. Even though the 4

CONNECTICUT YANKEE IfEEGRATED SAFETY ASSESSMEfE PROGRAM

operator has et least 100 minutes to close the LIV's before uncovering the core, he would typically recognize the situation and take action in no more than 30 to 45 minutes. Given the nature of the incident, and the fact that isolation of the faulted RCS loop is a secondary action, the quickest that the operator can be expected to take action, regardless of instrumentation, is on the order of 15 to 30 minutes.

Therefore, the maximum realistic reduction in direct radiological release time due to implementation of this instrumentation is one-half hour.

During the R.E. Ginna SGTR incident, the total whole body exposure rate for all individuals within 40 miles of the site was approximately 1 man-rem per hour (Section 8.6 of Reference 5), assuming the SG SRV was open for approximately ten minutes. Adjusting for CY's population density (Table D.1-2 of Reference 6) results in an exposure rate of approximately 2 man-rem per hour. The expression for the expected benefit from this modification is:

Expected benefit = (frequency of SGTR)(exposure rate)(decrease in recovery time)(remaining lifetime of the plant).

From the CY PSS (Section 2.0 of Reference 4), the frequency of an SGTR is 1.7E-02/yr. Therefore, the maximum expected benefit from implementation of this instrumentation is:

(1.7E-02/yr)(2 man-rem /hr)(1/2 hr)(20 yrs) = 0.34 man-rem.

3. Vent from Steam Generator Safety Relief Valves or Atmospheric Dump Valve The safety significance of this instrumentation would be to assist the operator in identifying a stuck open SRV following an SGTR. As such, the safety benefit of this proposed modification would be nearly identical to that of the proposed modification involving main steam flow from SRV's or SRV status.

4. Steam Generator Level CONNECTICUT YANKEE IEEGRATED SAFETY ASSESSMEE PROGRAM

Tmnlamanted

5. RCS Teold Tmnlamanted

6. Subcooled Margin Monitor Tmniamanted (see ISAP #1.13)

7. Core Exit Temperature Tmniamanted (see ISAP #1.13)

8. Containment H 2 M nitor See ISAP # 1.21

9. Reactor Vessel Level Monitoring System Tmniemented (see ISAP #1.13)

10. Steam Generator Pressure This proposed modification incorporates expanding the range of the steam generator (SG) pressure indicators, as well as environmentally and seismically qualifying them, and making them redundant. Currently the pressure indication range is from 50 psig to 1050 psig. The purpose of SG pressure indication in the low range is to make the operator cognizant of the need to isolate the appropriate SG. Since it will be obvious to the operator to isolate an SG long before the pressure decreases to 50 psig, having the indicator register pressures down to atmospheric pressure (as currently recommended) would provide no additional benefit to safety. The current upper set point is 1050 psig, which is above the highest SG SRV set point. Given the capacity of the SG SRV's at CY, there is no need to CONfECTICUT YANKEE ItITEGRATED SAFETY ASSESSMEffr PROGRAM

increase the upper pressure set point above the current level. Therefore, no significant safety benefit would be realized from expanding the range of the SG pressure indicators. However, if the SG pressure indicators are to be environmentally qualified, it would probably be prudent to expand their range at the same time.

Installing redundant channels for the pressure indicators would be of little benefit. The unavailability of these indicators is insignificant compared to the human error probability that would accompany their usage.

Also, their most likely cause of failure, given the conditions in the scenarios of concern, would be due to an adverse environment or seismic event, which would probaoly disable all sensors.

In evaluating the need for environmental qualification, the safety significance of the SG pressure indicators is to allow the operator to identify a faulted SG following a steam line break (SLB) upstream of its main steam nonreturn valve (NRV). To determine the expected benefit from this modification, the assumption was made that the pressure indicators have a 1.0 failure probability given exposure to an adverse environment (i.e., an SLB inside containment). The potential benefit of this modification is also limited by the fact that the operators can use auxiliary feedwater flow indication as an additional aid in identifying the faulted SG. Since the auxiliary feedwater flow transmitters are in the turbine building they would not be subjected to an adverse environment following an SLB in containment. Requantification of the affected accident sequences from the CY PSS (Reference 4) reveals a reduction in core melt frequency of 1.2E-5/yr. Incorporating the appropriate release category values from the Public Safety Impact Model results in a reduction of 0.64 Man-Rem over the remaining life of the plant.

In evaluating the need for seismic qualification, the safety significance of the SG pressure indicators is again to allow the operator to identify a faulted SG following an SLB upstream of its NRV. The identification and isolation of the faulted SG is necessary to prevent diversion of auxiliary feedwater flow away from the three intact SG's. The situation of concern involves a seismic event which causes an SLB upstream of an NRV. However,

_7-CONtJECTICUT YAtlKEE ItEEGRATED SAFETY ASSESSMEffr PROGRAM

cy:n given a saismic cytnt of sufficient magnitude to cause such an SLB, auxiliary feedwater flow indication exists, which is already seismically qualified, that could instead be used to identify the faulted steam generator (since auxiliary feedwater flow will be significantly higher to the faulted SG). Therefore, there appears to be no need to seismically qualify the steam generator pressure indicators.

11. RCS Makeup Flow This proposed modification involves both extending the range and environmentally qualifying the Chemical and Volume Control System (CVCS) makeup flow transmitters. The current range of these transmitters extends from 0 to 200 gpm. The primary safety benefit from makeup flow indication occurs during bleed and feed operation and high pressure recirculation.

However, the emergency procedures for bleed and feed operation specify maximum flow, so extending the transmitter range would have no safety benefit in this case. Also, while high pressure recirculation flow following a small or medium LOCA would exceed 200 gpm, current procedures just instruct the operator to use maximum flow, making ' knowledge of the exact flow immaterial. As such, no safety benefit would be realized by extending the range of these transmitters.

The CVCS flow transmitters are located in the Primary Auxiliary Building (PAB). Therefore, the primary concern from an environmental qualifying point of view is damage to the tramsmitters from high levels of radiation in the charging line. However, sequences involving charging do not result in high fuel damage unless there is a core melt, in which case charging flow regulation would be of no relevance. Another environmental qualification concern is a moderate energy line break that could lead to significantly elevated temperatures in the PAB. However, there is very little risk significance associated with these breaks since they are of very low frequency and none are postulated in which charging flow regulation would be either possible or of concern following their occurrence. As such, there is little need to environmentally qualify these transmitters.

CONNECTICUT YANKEE IEEGRATED SAFETY ASSESSMEE PROGRAM

120 Containment Prossuro There is no need to environmentally qualify the high range containment pressure transmitters since they are located outside of the containment and would be subjected to very limited radiation exposure.

13. Neutron Flux The safety significance of the neutron flux monitors is to alert the operators to initiate emergency boration in the event of an anticipated transient without scram ( AWS) . However, CY's emergency cperating procedures (References 7 and 8) clearly instruct the operator to emergency borate if the Rod Bottom Lights do not come on following a reactor trip.

Since these lights, if they fail, fail safe (i.e., unlit), they alone are sufficient to alert the operator as to the need for emergency boration.

(The probability of all rod bottom lights failing by spuriously transferring to the lit position,-coincident with failure to scram and failure of the neutron flux monitors is extremely low). The additional indication of neutron flux would only be of benefit should the operator fail to take the proper action in response to the unlit Rod Bottom Lights.

In this case, however, it is reasonable to assume that if the operator fails to respond to the Rod Bottom Lights, he will also fail to respond to the neutron flux monitors. As such, the neutron flux monitors appear to have no significant impact on public safety.

14. Containment Isolation Valve Position The safety significance of containment isolation valve position indication is to alert the operator to take action to secure a containment penetration in the event of failure of the isolation valve (s) to close automatically.

The modification considered here involves implementation of a redundant channel of containment isolation valve position indication. For this modification to have any impact on safety requires the following segaence of events: a severe accident which would lead to a radioactive release, in most cases rupture of one of the containment penetration lines, failure of 1 or 2 containment isolation valves, failure of the first channel of valve 9

CONNECTICUT YANKEE IIITEGRATED SAFETY ASSESSMEllr PROGRAM

position indication, and the operator has to know, and be able, to take corrective action. The frequency of this sequence of events is so insignificant that no real safety benefit would be realized. In. addition, as an insight from the CY PSS (Reference 4), the only containment

~

penetrations that could contribute significantly to the release following a severe accident are the containment purge inlet and exhaust valves, which are locked closed during normal operation. Therefore, this modification would have negligible impact on public safety.

15. Low Pressure Safety Injection Flow Indication The NRC recomends the implementation of flow indication (0-110 %) for the Low Pressure Safety Injection (LPSI) System. However, since indication

, currently exists for LPSI pump current and for LPS1 m. charge (i.e., core deluge) valve position, indirect flow verification already exists. Also, since the LPSI system is never used for recirculation (where flow control would be necessary), but rather for injection (where maximum flow is always used), direct flow indication would be of 'no benefit, especially since there are no operator actions that could be taken even if no, or low, flow was detected. It has therefore been determined that this modification would have negligible impact on public safety.

16. Containment Atmosphere Temperature This project proposes to environmentally qualify the' containment temperature indicators and to increase the upper limit of'their range from 150 F to 400 F. However, none of CY's operating procedures refer to containment temperature for decision-making. All operator decisions that are concerned with containment environment are based on containment pressure. As such, containment atmosphere temperature indication has no effect on human reliability. Therefore, this modification would have negligible impact on public safety.

17. Emergency Ventilation Damper Position CONNECTICUT YANKEE I E EGRATED SAFETY ASSESSMEE PROGRAM

This modification involves upgrading the quality of the emergency ventilation damper position indication. For this indication to be of benefit to public safety requires a severe accident and breach of containment (most likely a LOCA outside containment leading to core melt),

a failure of the damper to transfer to the filter position, and impairment of the operator to take proper actions. The frequency of this serics of events is very low, and it nust also be noted that once a situation exists where the control room would be exposed to high radiation, there is very little the operator could do that would benefit public safety. It has therefore been determined that this modification would have negligible impact on public safety.

18. Containment Fan-Cooler Heat Removal This project proposes to incorporate two modifications: environmental qualification of the instrumentation used to monitor heat removal by the Containment Fan-Cooler Heat Removal System and installation of similar equipment in the control room. The control room at CY currently includes indication of Service' Water System (SWS) flow as well as indication of which containment air recirculation fans are running. This indicatiion can be used indirectly to monitor containment fan-cooler beat removal. Also, the safety significance of monitoring containment fan-cooler heat removal is to alert the operator of the need to initiate containment spray.

However, the initiation of containment spray is, per procedure, based entirely on containment pressure. Therefore, these modifications would have negligible impact on public safety.

Results The benefit to public 37Pety for each instrumentation modification is

summarized in Table 1. iLe cumulative benefit of all the modifications is about 1 man-rem, recognizing that the modifications concerned with SG SRV Status and SG SRV rad monitors accomplish the same purpose. The reduction in core melt frequency and public risk correspond to a score of 1.2 on a scale of

-10 to 10.

i l

l i

CONNECTICUT YANKEE INTEGRATED SAFETY ASSESSMENT PROGRAM

Referene n

1. Regulatory Guide 1.97, Revision 2, " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," December, 1980.

2. Letter from W. G. Counsil to D. M. Crutchfield/J. R. Miller of NRC, dated May 31, 1984.

3 Letter from J. A. Zwolinski, NRC to J. F. Opeka, dated July 3, 1985.

4. J.F. Opeka letter to C.I. Grimes, "Haddam Neck Plant Probabilistic Safety Study - Summary Report and Results," Docket No. 50-213, dated March 31, 1986.

5. NUREG-0916, " Safety Evaluation Report Related to the Restart of R. G. Ginna Nuclear Power Plant," U. S. Nuclear Regulatory Commission, May 1982.

6. NUREG/CR-2239, " Technical Guidance for Siting Criteria Development," Sandia National Laboratories, November 1982.

7. Connecticut Yankee Emergency Operating Procedure No. E0P 3.1-0, Rev. O, Emergency Response Procedures.

8. CY Functional Restoration Guideline, FR-S.1, " Response to Nuclear Power Generation /A1WS," HP-Rev. 1, September 1, 1983.

CONNECTICUT YANKEE IEEGRATED SAFETY ASSESSMEE PROGRAM

. . 'N)

Table 1. Safety Benefit of Reg. Guide 1.97 Instrumentation Modifications Risk Reduction CMF Instrumantation Modif'ications (Man-Rem) Reduction (v1 1 Se,gg HPSI System Flow I 0.007 1.3E-7 0 SG SRV Status I 0.34 Negligible O SG SRV Rad Monitor I 0.34 Negligible O SG Level E Implemented -- --

RCS T E,S,R Implemented -- --

cold Subcooled Margin Monitor RD,E,R See ISAP #1.13 -- --

Core Exit Temperature RD See ISAP #1.13 -- --

Containment Hydrogen Monitor I See ISAP #1.23 -- --

Reactor Vessel Level Monitoring System I See ISAP #1.13 -- --

SG Pressure RD,E,S,R 0.64 1.2E-5 1.2 Makeup Flow RD,E Negligible Negligible O Containment Isolation Valve Positions R Negligible Negligible O LPSI System Flow I- Negligible Negligible O Containment Atmosphere Temperature RD,E Negligible Negligible O Emergency Ventilation Damper Position E,S Negligible Negligible O Containment Fan Heat Removal I,E Negligible Negligible O Neutron Flux E,S Negligible Negligible O Containment Pressure E Negligible Negligible O Notes: 1) Modifications are as follows:

I - installation RD - range determination E - environmental qualification S - seismic qualification R - redundancy CONNECTICUT YANKEE INTEGRATED SAFETY ASSESSMEffr PROGRAM