ML18107A319

From kanterella
Jump to navigation Jump to search
Review of Submittal in Response to NRC GL-88-20,Suppl 4: Ipeees, Fire Submittal Screening Review Technical Evaluation Rept:Salem Generating Station,Rev 1
ML18107A319
Person / Time
Site: Salem  
Issue date: 05/21/1999
From: Lachance J, Pepping R, Ross S
Battelle Memorial Institute, COLUMBUS LABORATORIES, SANDIA NATIONAL LABORATORIES, SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY
To:
NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
Shared Package
ML18107A309 List:
References
CON-FIN-W-6733 GL-88-20, NUDOCS 9905270222
Download: ML18107A319 (23)
v • d • e


Text

.,

"Individual Plant Examination-External Events" Fire Submittal Screening Review Technical Evaluation Report:

Salem Generating Station Revision 1: October 20, 1997 Prepared by:

R.. E. Pepping Risk Analysis and Systems :Modeling Department Sandia National Laboratories Albuquerque, New Mexico 87185-0747 Steven B. Ross Battelle Albuquerque, New Mexico 87106 Jeffrey L. LaChance Science Applications International Corporation Albuquerque, New Mexico 87106 Prepared for:

Probabilistic Risk Assessment Branch Division of Systems Technology Office of Nuclear Regulatory Research U. S. Nuclear Regulatory Commission Washington, D.C. 20555 USNRC JCN W6733 1

\\"

1.0 INTRODUCTION

This Technical Evaluation Report (TER) presents the results of the Step 0 review of the fire assessment reported in "Salem Generating Station Individual Plant Examination of External Events" [l].

1.1 Plant Description Salem Generating Station (SGS) is a two-unit plant, both units being Westinghouse 4-loop pressurized water reactors rated at 3411 MWt. Each unit's nuclear steam supply is enclosed in a large, dry, reinforced, steel-lined containment. SGS Unit 1 began commercial operation in June 1977. Unit 2 began commercial operation in October 198L The tWo units are essentially identical. Both units share a 700-acre site in Salem County, New Jersey, with the Hope Creek Generating Station.

1.2 Review Objectives The performance of an IPEEE was requested of all commercial U.S. nuclear power plants by the U.S. Nuclear Regulatory Commission (USNRC) in Supplement 4 of Generic Letter 88-20 [2].

Additional guidance on the intent and scope of the IPEEE process was provided in NUREG 1407 [3]. The objective of this Step 0 screening review is to help the USNRC determine ifthe Salem Generating Station submittal has met the intent of the generic letter and to also determine the extent to which the fire assessment addresses certain other speCific issues and ongoing programs.

1.3 Scope and Limitations The Step 0 review was limited to the material presented in the Salem Generating Station IPEEE submittal. Furthermore, the review was limited to verifying that the critical elements of an acceptable fire analysis have been presented. An in-depth evaluation of the various inputs, assumptions, and calculations was not performed. The review was performed according to the guidance presented in Reference 4. The results of the review against the guidance in this document are presented in Section 2.0. Conclusions and recommendations as to the adequacy of the SGS IPEEE submittal with regard to the fire assessment and its use in supporting the resolution of other issues are presented in Section 3.0.

2

l "

2.0 FIRE ASSESSMENT EVALUATION The following subsections provide the results of the *review of the Salem Generating Station fire assessment. The review compares the frre assessment against the requirements for performing the IPEEE and its use in addressing other issues. Both areas of weakness and strengths of the fire assessment are highlighted.

2.1 Compliance with USNRC IPEEE Guidelines The USNRC guidelines for performance of the IPEEE fire analysis derive from two major documents. The first is NUREG-1407 [3], and the second is Supplement 4 to USNRC Generic Letter 88-20 [2]. In the current screening assessments, the adequacy of the utility treatment in comparison to these guidelines has been made as outlined in "Guidance for the Performance of Screening Review of Submittals in Response to U. S. NRC Generic Letter 88-20', Supplement 4:

Individual Plant Examinations - External Events," Draft Revision 3, March 21, 1997 [4]. The following sections discuss the utility document in the context of the specific review objectives set forth in this Screening Review Guidance Document and assess the extent to* which the utility submittal has achieved the stated objectives.

2.1.1 *Documentation A single submittal was reviewed which documents the fire assessment of SGS Unit 1. A section of the submittal addresses the differences between the units and finds them insignificant in their effects on the fire analysis. Comparing the basic events data indicates the Unit 2 core damage frequency (CDF) to be about 4% above that for Unit 1.

The analysis begins with the definition of the fire areas and screening according to the FIVE methodology._[5] SGS uses the fire areas defined by its Safe Shutdown Analysis (SSA) and Fire Hazards Analysis (FHA). These are bounded by fire barriers with a minimum two-ho:ur rating.

Within these fire areas, compartments are defined as bounded by non-combustible barriers, not necessarily rated.

The analysis steps outlined in the submittal are as follows:

1) Qualitative screening of fire areas based on equipment, cables, and reactor trip initiators, either automatic and manual, that might result from the loss of all equipment assigned to the fire area.
2) Quantitative screening of fire areas (numerical screening criterion of lE-6 per year) based on fire frequencies and the probability of core damage, given the loss of all equipment in the area due to the frre. Both fixed and transient combustibles are considered. Both fire frequencies and the plant IPE model are needed for quantitative screening, although the submittal does not note their introduction at this point. Sequences of concern involve transients with and without power conversion system (PCS) available, loss of off-site power (LOSP), and a small loss of coolant 3

- --*-- -,..,_ ------~-

.-.-------~*-**------*-.-***--::-~~-,,.....n~*-*--------*-..,.--.-*--~-----------

~-----

  • accident (SLOCA).
3) Fire areas remaining after screening are treated in more detail, beginning with subdividing the areas into compartments and applying the FIVE Fire Compartment Interaction Analysis (FCIA).

The FIVE methodology generally performs the FCIA before the quantitative screening steps.

Since all screening was done at the fire area level, this is. of no consequence.

4) The analysis proceeds at the compartment scale and introduces the EPRI Fire Events Database fire frequencies [6], the COMPBRN Ille model for fire and damage modeling [7], explicit fire detection and suppression, operator actions, and the internal events plant model.

The submittal notes that the ASEP methodology was used to estimate probabilities of successful operator actions. [8] Some fire-specific human error probabilities (HEP) are discussed below.

Also noted in the submittal, some safe shutdown equipment assigned to Unit 1 is physically located in Unit 2 fire areas. Shared areas are addressed in the analysis by assigning areas to the unit under study regardless of the location, according to the equipment and cabling contained.

The IPEEE fire submittal from Salem is generally well written and easily read. There are several references to NSAC/181L [9] for methodology guidance.

2.1.2 Plant Walkdown Numerous walkdowns were performed as a part of this study. The walkdown team consisted of four staff members of the Public Service Electric and Gas Company and four supporting contractors. The expertise represented by the walkdown team included knowledge of fire protection systems, safe shutdown systems, PRA and Appendix R analysis methods, and the plant layout and operation. Walkdowns were performed after screening to confirm screening results and the analyses performed, that is, whether the as-analyzed plant and as-built plant were the same. A final walkdown was also performed in accordance with the FIVE methodology and to address the fire risk scoping study (FRSS) issues.

In addition to these walkdowns, numerous informal walkdowns were perfonrted to verify cable routing, COMPBRN modeling input, and the modeling details used in the control room.

The walkdowns performed appear to be thorough, well-focused.

2.1.3 Fire Area Screening Initially, there are fifty fire areas for Unit 1 tabulated in the submittal. Thirteen fire areas, including the containment, are eliminated by qualitative screening. A summary table presented in the submittal indicates whether there is safe shutdown equipment, or if a trip demand would result from a fire in the area. The containment is scre*ened without regard to the qualitative screening criteria, which the submittal notes as in accordance with FIVE guidance. Several Unit 4

.l*

  • 2 areas appear in the tabulation, which indicates some equipment or cabling associated with Unit
1.

Quantitative screening results for Unit 1 are also tabulated in the submittal. The area fire frequencies are tabulated also. For each area screened, a short calculation shows the numerical work and a short explanation of the quantities and/or assumptions included. Sixteen fire areas are quantitatively screened and include several areas associated with the diesel generators, a battery room, and some auxiliary building areas.

Following screening, twenty-one fire areas remained and received the most detailed analysis.

The screening analysis appears to have been performed properly.

2.1.4 Fire Occurrence Frequency The EPRI fire events database (FEDB) [6] is the cited source of fire frequencies. Both fixed and transient combustibles are considered. The submittal notes that the plant's fire history is consistent with the generic data and, therefore, no updating or plant-specific modifications have been performed. Summary tables list the fire area frequencies and scenario descriptions for the fire areas and compartments surviving screening. The listed values appear reasonable.

2.1.5 Fire Propagation and Suppression Analysis Both COMPBRN Ille and the FIVE methodology are employed in performing damage estimates from fires in unscreened fire compartments. COMPBRN appears to have been used for fixed combustibles and targets ari.d the FIVE formula for transient combustibles.

  • A.II cabling appears to be IEEE qualified. (The submittal only specifically notes cables in cable trays, and cabling in cabinets as IEEE qualified.) That feature is offered as the basis for discounting cable fires that are either self-ignited or initiated by welding activities. The damage parameters for qualified cable are taken to be a relatively conservative 622 °F and 1 Btu/ft2-s.

Cabling in a conduit is assumed to become damaged at the same temperature and heat flux, but not ignite or offer a propagation path. Non-cable electrical components are assumed to damage at 322 °F.

One weakness of the submittal is the limited discussion of fire detection and suppression. The discussion addresses the time delays calculated by COMPBRN for detection and actuation of these systems. The characteristics for these systems were taken from the FEDB. The implicit assumption is that they are designed, installed and maintained in accordance with industry standards. The submittal does not specifically state that this is the case. The control room fire scenario contains a discussion of detection and suppression specific to the control room. Beyond that, there is little discussion other than the appearance of some non-s_uppres*sion probabilities, typically 0.04 and 0.05, in other scenarios.

Manual suppression is credited only in scenarios involving fires in completely enclosed electrical cabinets and some areas in the auxiliary building. The fire brigade response time has been 5

~-*.****-

-*--,-...,..,....._..,-~-------... -.---..--.

..:.............. _--~~*~---.~

determined by fire drills and determined to be five minutes for the first member to arrive. The time required to extinguish a fire is not discussed. However, it is not clear that either of these time intervals were used in the analysis of any particular scenario.

Other assumptions of note include, Electrical cabinets are assumed to have a heat release rate of 1200 kW, unless otherwise stated.

Cabinets, including control room cabinets, are credited with fire containment capabilities, if they are closed and sealed. Propagation is possible only through open doors, ventilation louvers, and unsealed penetrations. Thermal damage is still possible outside of the cabinet.

A reflectivity of 0.7 is noted for electrical cabinets.

Most fires are modeled with COMPBRN Ille as spills of combustible liquids.

Fire wrap was credited as effective. Cables with fire wrap are excluded from combustible loads. (The word "Thermo-Lag" does not appear in the submittal.)

The general discussion of the fire growth and propagation is adequate, although less detailed than desired, since it relies on approved or familiar methods.

2.1.6 Fire-induced Initiating Events and Fire Scenarios Core damage frequency (CDF) calculations.were performed with internal events IPE models.

Fault trees containing equipment that could be affected by fire were modified by adding a "house event" to allow insertion of failures due to fire. Sequences of concern include transients with and without power conversion system (PCS) available, loss of off..:site power (LOSP), and a small loss of coolant accident (SLOCA). Four event trees are included in the submittal as.

examples.

Operator actions were treated by considering the action required and the location of the fire. Any action required to take place in an area affected by the fire was not allowed to be effective until the fire was extinguished, typically requiring 30 minutes. A particular instance where the 30 minutes are required occurs for switchgear room fires that also result in a need to execute remote shutdown. In this instance, the operator is required by procedure to enter the switchgear room itself, thereby delaying the shutdown. Similarly, for fires in the control room or relay room that disable safety equipment and lead to a requirement for remote shutdown, an arbitrary 15 minute delay is assumed to account for a period of confusion before smoke has forced the evacuation or before the conclusion has been made that instrument readings and control functions are being affected by the fire.

6

  • HEPs were raised in cases where the fire affected instruments needed to make emergency decisions. If the system was failed, the HEP was.set to unity. If the system was merely degraded, the HEP was raised by a factor of ten. If long time periods were available to execute a recovery, the IPE value was used.

For the 21 remaining fire areas, summary tables describe the resulting CDF contributions from each scenario and I or fire compartment and are included in the submittal. For three fire areas, the control room, the relay room, and a switchgear room, a detailed discussion in the text provides examples of how the CDF was determined.

The control room The major fire hazard in the area is the large amount of paper stored in filing cabinets. Other hazards are presented by other paper combustibles and low voltage electrical cabling in the control consoles. Fire area 12FA-AB-122A contains the two control rooms for the two units.

The two rooms are separated by a walled corridor containing glass panels. Since the area is continuously occupied, any fire is expected to be rapidly extinguished. No credible fires are expected to challenge the modest barriers. A smoke-induced evacuation is a concern.

Smoke and fire detectors are located throughout the area, most notably in the control room consoles and the return air ventilation ductwork. Portable C02, Halon, and water extinguishers are available and a hose station is located in the corridor.

A "critical cabinet" analysis, as described in NSAC/181L was performed. The fire frequency is partitioned equally among the 70 cabinets. Those not affecting safety equipment are of concern only from the possibility that a fire* in one of them would produce enough smoke to force abandonment. The smoke buildup is assumed to require 15 minutes, during which time manual suppression is expected to fail with a probability-of 3.4E-3. A remote shutdown is then necessary. This scenario is an insignificant contribution to the CDF, according to the submittal, and is not discussed further.

The significant contributors to the CDF are those that lead to loss of control. Three scenarios are developed and quantified: 1) a recorder panel which is damaged by fire, but the fire does not propagate, 2) a control console which is damaged by fire, but the fire does not propagate, 3) a multi-cabinet fire that both damages and propagates, and leads to evacuation. (A "severity factor," derived from FEDB data indic8;ting that no evacuations have been forced by 12 control room fires, was introduced in this scenario.)

The discussion of the scenarios concludes with the event trees assumed and a discussion of the considerations and assumptions at each branch point. There are 18 scenarios deemed credible, differing only in the specific cabinet(s) is involved. Their combined contribution to the CDF is 7E-6 per year. This result appears to have been obtained in a straightforward manner.

7

.e

  • The Relay Room
  • The submittal notes that both transient and fixed ignition sources were considered in this room.

Cable insulation and plastic components in electrical cabinets dominate the combustible load.

The majority of the equipment represented by these cabinets is safety related. A significant fire in the area could lead to a need to execute remote shutdown.

The two scenarios for this room are distinguished by the number of emergency power trains affected by the fire. While single train involvement allows alternate trains to be used, a hot short in the affected train can lead to a stuck PORV and SLOCA. A multi-train involvement can lead to both a stuck PORV and a need for remote shutdown.

Automatic suppression is credited in this scenario, although its characteristics are not discussed.

A manual non-suppression is also included and given 0.1 probability.

The relay room contribution to the plant CDF is 7.2E-6 per year.

The 460V/230V Switchgear Room There are eighteen scenarios for this fire area (1FA-AB-84A). Each corresponds to a particular cabinet (or transient combustibles) and more than fifty targets are potentially damaged. The targets include cabling and busses to pumps and coolers, or their supplies (batteries and generators).

The fire modeling for these cabinets used walkdown observations that the cabinets were divided into front and rear sections, each isolated from the other by a steel wall without penetrations.

The rear section contained ventilation louvers that could allow fire propagation, but the top penetrations were sealed. Transformers, if present in a particular cabinet, are dry-type with minimum combustible loads. Only the ventilation louvers allow fire to propagate, that is, the cabinet is otherwise assigned a fire containment capability. Low combustible loads and available oxygen are considered in making this determination. Battery chargers are discredited as sources of propagating fires following NSAC/181L guidance.

COMPBRN Ille was used to model propagation to nearby cable trays from those cabinets that are not completely enclosed. The cabinet was modeled as a 97-kW pool fire with 30 minute duration. Sandia tests on vertical cabinets with in-situ loads consisting of qualified cabling are cited as the basis for this heating rate. Suppression is not credited and ventilation is assumed to be disabled in the calculation. The COMPBRN calculation showed little lateral propagation of these fires and a maximum resulting room temperature of 158 °F. The consequence of this finding is that no multiple safety train scenarios are significant contributors to the CDF.

The various fires considered for this fire area resulted in transients both with and without recoverable PCS. The event trees used to quantify these scenarios are presented and discussed, one branch and a time. The final CDF contribution from this area is 1. 7E-6 per year.

8

- -~.==:--=-. --*---------........-.--.-,,,,.~------*-* -,.-.-.-~~-*~-----------,.-.. ----------*--*..--.------------ -------* ---~-*-..-----*-..,... -

2.1.7 Quantification and Uncertainty Analysis Quantification of the CDP from the fire scenarios considered are tabulated. In addition to the detailed discussion of the three fire compartments, CDP contributions from the remaining areas of Unit 1 are tabulated below. The total fire CDP for Unit 1is2.3E-5 per year. The submittal states that there are no significant differences between Units 1 and 2 insofar as the fire risk is concerned. An estimate of the Unit 2 fire risk, based on basic events data, is 2.4E-5 per year.

Unit 1 CDF Contributions Fire Area Description CDF x 1E6 yr 1 F A-AB-1 OOA Relay Room 7.2 12FA-AB-122A Control Rooms, Peripheral 7

Room, Ventilation Room 1FA-AB-84A 460 V Switchgear Room 1.7 1FA-AB-64A 4160 V Switchgear Room 1.7 1 F A-EP-78C Lower Electrical Penetration 1.4 Area 1 FA-EP-1OOG/1 FA-PP-Upper Electrical and Piping 1.3 1 OOH Penetration Area 1 FA-AB-848 Reactor Plant Aux Equip Area 1.1 12FA-SB-100/ 1FA-TGA::ia8'bine and Service Buildings 0.64 1 FA-SW-90A/90B Service Water Intakes (2 areas) 0.42 1 FA-AB-10C Reactor Plant Aux Equip Area 0.29 1 FA-MH-94 Service Water Duct, Manhole 0.21 1 FA-DG-84F C02 Equip Room 0.06 9

~**-*-- ------*-

-*-*-*-*--*-.*---*~*... -~.

_..,.~.... ---------*--*--- -- -. -- -----*- ~---~--****--*---~--

No uncertainty analysis results were presented. The submittal notes that several conservatisms were included in the analysis, such as heat release rates, fire durations, and notes ongoing work to improve modeling and data assumptions.

2.1.8 Sensitivity and Importance Ranking Studies No sensitivify or importance ranking studies were performed.

2.2 Special Issues As a part of the IPEEE fire submittal, the utilities were asked to address a number of fire-related issues identified in the Fire Risk Scoping Study (FRSS) and USNRC Generic Safety Issues (GSI). Specific review guidance on these issues is found in Reference 4.

The submittal response on the FRSS issues was minimal in the sense that FIVE guidance was*

used. The responses did not exceed the information specifically requested in.that guidance.

There was no discussion of those aspects of generic issues that differed from the areas addressed

'by the FIVE methodology.

2.2.1 Decay Heat Removal (GSI A-45)

The SGS response on this topic is reproduced here in its entirety: "The SGS IPE evaluated the relative importance of the core decay heat removal function, and it was concluded that there are no vulnerabilities in the systems used to perform decay heat removal. On the basis of a similar contribution for loss of decay heat removal sequences obtained from the fire PRA as was obtained in the IPE, it is concluded that no decay heat removal function vulnerabilities exist with respect to internal fire events. Therefore, USI A-45 is judged to be resolved."

The estimated CDF of2.3E-5 and 2.4E-5 per year or Units 1 and 2, respectively, are below the limits where NUREG-1289 would require corrective action. [11]

2.2.2 Effects of Fire Protection System Actuation on Safety-related Equipment (FRSS, GSI 57, MSRP)

This issue is associated with the concern that traditional fire PRA methods have generally

.considered only direct thermal damage effects. -Other potential damage mechanisms such as smoke and fire suppression damage (either from fixed systems or manual actions) have not been considered. In general, this is an area where the data base on equipment vulnerability is rather sparse.

The submittal notes FIVE guidance as indicating no short-term damage effects attributable to smoke. It notes, however, that smoke effects were considered in the evaluation of operator recovery actions in the PRA portion of the fire study and in determining the procedures and systems required to complete post-fire shutdown.

10

*--------~-------------

  • A seismic/fire walkdown was performed, during which the effects of fire suppressant sprays were addressed. Protection from suppressant sprays has been installed as needed. Typically these involve spray deflectors and baffles, and floor drains. No credible scenario-leads to the loss of more than one shutdown train at SGS. The GSI 57 concern addressing seismic actuation of the fire protection system was not discussed.

2.2.3 Fire-induced Alternate Shutdown/Control Room Panel Interactions (FRSS, GSI 147)

The issue of control systems interactions is associated primarily with the potential that a fire in the Main Control Room might lead to failures in the remote shutdown capability.

The submittal notes that independent remote control and monitoring functions exist from the auxiliary shutdowt1 panels at SGS, and that the control room circuits can be isolated from the rest of the plant from outside the control room. The discussion of the main control room fires also notes that an abnormal operating procedure dealing with remote shutdown exists and describes operator actions that are required.

Not discussed in the submittal are the GSI 147 concerns that address the loss of power and control of systems prior to transfer of control to the remote shutdown panels, verification of the remote shutdown capability, and the location and capability of the remote shutdown panels.

2.2.4 Smoke Control and Manual Fire Fighting Effectiveness (FRSS, GSI 148)

Smoke control and manual fire fighting effectiveness is associated with the concern that nuclear power plant ventilation systems are known to be poorly configured for smoke removal in the event of a fire, and hence, a significant potential exists for the buildup of smoke to hamper the efforts of the manual fire brigade to suppress fires promptly and effectively.

The Nuclear Fire Protection Organization (NFP) is a fire department dedicated to SGS. In that function, NFP provides fire department personnel and equipment, operates fire protection systems, performs surveillance and maintenance activities, and monitors the effectiveness of the transient combustibles program.

The submittal notes that "the 16 checklist questions identified in FIVE" were all answered in the affirmative, which indicates that the SGS manual fire fighting program is consistent with FIVE guidelines. (The checklist questions themselves were not identified.)

Not discussed in the submittal are the GSI 148 concerns that address firefighting practices as they potentially affect redundant safety trains, opening of fire barriers to access fires, and the various delay times relevant to fire damage modeling, such as time to extinguish a fire, time to assemble the fire brigade, time to put on protective clothing and breathing gear, and delays of fire brigade effectiveness caused by working in a smoke-obscured environment.

11

2.2.5 Seismic/Fire Interactions (FRSS, MSRP)

The issue of Seismic Fire Interactions involves primarily two concerns. First is*the potential that seismic events might also cause fires internal to the plant, and second is the potential that seismic events might render inoperable or spuriously actuate fixed fire detection and suppression systems. Following this postulated event, fire suppressant inventories may be depleted and/or damage to safety-related equipment may be an issue. It had been anticipated that a typical fire IPEEE submittal would provide for some treatment of these issues through a focused seismic/fire interaction walkdown.

The seismically-induced fire was addressed in the seismic walkdown. It was concluded that there are no sources of flammables that could cause a seismically induced fire impact on safe shutdown equipment. Mechanical failure of suppression systems was found not to be of concern at SGS. Not discussed explicitly in the submittal are electrical cabinet anchorage, depletion of suppressant inventories, or seismic dust-actuated suppression systems.

2.2.6 Adequacy of Fire Barriers (FRSS)

Barrier reliability and inter-compartment fire effects are related to the potential that fires in one area might impact other adjacent or connected areas through the spread of heat and smoke. In general, it is expected that a utility analysis would provide for some treatment of such potential by considering that (1) manual fire fighting activities might a_llow for the spread of smoke and heat through the opening of access doors, and (2) that the failure of active fire barrier elements such as normally open doors, water curtains, and ventilation dampers might compromise barrier integrity.

The submittal notes that "the six checklist questions identified in FIVE" were all answered affirmatively, indicating that SGS has verified that the fire barriers, doors, penetration seals, and dampers are being maintained and inspected properly. (The checklist questions themselves were not identified.) While there is no discussion in the submittal of responses to the various fire seal and damper-related Information Notices (IN), these INs are typically included in the FIVE guidance.

2.2. 7 Effects of Hydrogen Line Ruptures (MSRP)

The use of flammable gases in the plai;:it, including hydrogen, introduces the potential that a rupture of the gas flow lines might lead to the introduction of a serious fire hazard into plant safety areas. It had been anticipated that a typical fire IPEEE analysis would include the consideration of such sources in the analysis.

The submittal states that the FIVE seismic/fire interactions guidance was used, which explicitly states that hydrogen in the plant be examined as potential fire hazard. According to the submittal, all elements of the FIVE seismic/fire interactions guidance were verified at SGS.

12

2.2.8 CommoJi Cause Failures Related to Human Errors (MSRP)

Common cause failures resulting from human errors include operator acts of omission or commission that could be initiating events or could affect redundant safety-related trains needed to mitigate other initiating events. It had been anticipated that a typical fire IPEEE analysis would include the consideration of such failures in the submittal.

SGS discussed several modifications to models to account the effects of the fire environment on damage estimates. Although appearing somewhat ad hoc numerically, the study correctly identifies several areas where such modifications are needed. In two cases, delays of operator actions were introduced (15 minutes in the control room, 30 minutes in the switchgear room) to account for expected effects of the fire on the recovery actions. In other cases, the human error probability was adjusted to account for the fire-damaged state of the plant.

2.2.9 Non-safety Related Control System/Safety Related Protection System Dependencies (MSRP)

Multiple failures in non-safety-related control systems may have an adverse impact on safety-related protection systems as a result of potential unrecognized dependencies between control and protection systems. The licensee's IPE process should provide a framework for systematic evaluation of interdependence between safety-related and non-safety related systems and identify, potential sources of vulnerabilities. It]1ad been anticipated that the fire IPEEE analysis would include the consideration of such dependencies in the submittal.

Non-safety systems were not discussed in the SGS submittal.

2.2.10 Effects of Flooding and/or Moisture Intrusion on Non-Safety-and Safety-Related Equipment (MSRP)

Flooding and water intrusion events can affect safety-related equipment either directly or indirectly through flooding or moisture intrusion of multiple trains of non-safety-related equipment. This type of event can result from external flooding events, tank and pipe ruptures, actuations of the fire suppression system, or backflow through part of the plant drainage system.

It had been anticipated that the fire IPEEE analysis would include the consideration of such events in the submittal.

Non-safety systems were not discussed in the SGS submittal. Actions to mitigate local flooding and suppressant spray effects on safety systems were noted above in the discussion of GSI 57.

13

*------*-~------------*---**--**--------~

.. -.-~----:--..--..,,...,,-.---....-~----------------------***---------***--*-.-.-~------------*-----------*--*-***-*-*

.L I

I\\ rl 2.2.11 Shutdown Systems and Electrical Instrumentation and Control Features (MSRP)

The issue of shutdown -systems addresses the capacity of plants to ensure reliable shutdown using safety-grade equipment. The issue of electrical instrumentation and control addresses the functional capabilities of electrical instrumentation and control features of systems required for safe shutdown, including support systems. These systems should be designed, fabricated, installed, and tested to quality standards and remain functional following external events. It had been anticipated that the fire IPEEE analysis would include the consideration of this issue in the submittal.

Shutdown heat removal is discussed under the section addressing decay heat removal (GSI A-45). There was no explicit discussion of instrumentation and control in this regard.

2.3 Containment Performance Issues Unique to Fire Scenarios The submittal addresses containment bypass, containment isolation, and containment heat removal. In the case of containment bypass, for example, interfacing systems LOCA and steam generator ruptures, the fire study identified no new vulnerabilities and none at frequencies significantly higher than that in theJPE.

One containment isolation valve was'found to have a hot short concern. This valve could, however, be manually closed.

The loss of containment heat removal was found to be similar to and comparable in frequency to the IPE result for this issue. Hence, there are no new or unusual impacts on containment heat removal performance due to fire.

2.4 Plant Vulnerabilities and Improvements The supplemental information accompanying the submittal notes that two improvements will be implemented by July 1996 and will address the transient combustibles control program, and the auxiliary building ventilation. The transient combustibles control improvement addresses concerns that a transient fire could initiate a LOSP from a location were redundant cable-trains

  • are in close proximity. The ventilation issue follows from a concern that a fire could fail normal HV AC functions in some switchgear and control areas. Procedures have been implemented to ensure operator actions can be credited for re-establishing cooling to these areas.

14

_:_:..:.::..._:._-*-:._-_--:._-:._*--_-_-_-_--*_--_*-_---_--_--_--_--_-_--_--_:::-_-_-_--*-_-.::::-.:::-...=-_-__c~_-_----'-*--*--=--=----_-_~_-_-_-_-_--==-----

.J_ I'

3.0 CONCLUSION

S AND RECOMMENDATIONS l

The Salem Generating Station IPEEE fire assessments was performed by application of the FIVE

  • methodology and existing PRA models. In general, the submittal documents the methods and results of the fire.assessment, and the FRSS issues. Some information is presented relevant to resolution of the Generic Safety Issues. In particular, the submittal regards its GSI A-45 issue resolved, and the FRSS concerns adequately addressed with the submission of the IPEEE.

Strengths and weakness of the submittal are noted here along with general observations that, in this submittal-only review, can not be assessed as to whether they enhance or detract from the quality of the submittal, Strengths:

  • The fire study was well documented and is easily read. Examples and summary tables illustrate the work performed and results obtained.
  • Numerous walkdowns were performed by a competent team of analysts.and plant personnel.
  • The plant IPE model was used to supplement the FIVE methodology.
  • A detailed and conservative (in its assumptions, such as worst-case spurious control failures) analysis of the control room.

Weaknesses:

    • Containment was screened without the benefit of a confirmatory walkdown.
  • The description of fire detection and suppression systems was limited, and not discussed ii:i the context of specific scenarios.
  • The treatment of the various time intervals for detection, actuation, suppression, and in particular, for manual suppression was not described.
  • Fire suppression systems' reliabilities were taken from Reference 6, without noting the design, installation, and maintenance requirements necessary to support those reliabilities.

General Observations

  • Both units have been assessed as being virtually identical, to the extent that Unit 2 is expected to have a fire CDF that is 4% higher than Unit 1, based only on consideration of basic events data.
  • Numerous references are made to the use of NSAC/181, without mentioning of any of the specific, known weakness in that study. [10]
  • One-hour fire-wrap was credited in some Appendix R areas. The failure of the barrier may then lead to the loss of redundant safety trains. An important sensitivity to the fire... wrap would be quantified by repeating the analysis with credit not taken for the fire wrap for those areas.

The reviewer recommends that a sufficient level of documentation and appropriate bases for analysis have been established to conclude that the Salem Generating Station IPEEE fire 15

---*------*---~ ------* -- - -

  • submittal has substantially met the intent of the IPEEE process. No further review is recommended at this time.

16

I i

I I..

I J

4.0 References

1) "Salem Generating Station Individual Plant Examination of External Events", Public Service Electric and Gas Company, January 1996.
2) USNRC, "Individual Plant Examination of External Events for Severe Accident Vulnerabilities - 10 CPR §50.54(f)," Generic Letter 88-20, Supplement 4, April 1991.
3) USNRC, "Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities," NUREG-1407, May 1991.
4) S. Nowlen, M. Bohn, J. Chen, Guidance for the Performance of Screening Reviews of Submittals in Response to U.S. NRC Generic Letter 88-20, Supplement 4: "Individual Plant Examination - External Events," Rev. 3, 21 Mar 1997.
5) EPRI TR-100370, Fire-Induced Vulnerability Evaluation (FIVE), Professional Loss Control, Inc., April 1992.
6) Electric Power Research Institute, "Fire Events Database for US Nuclear Power Plants, NSAC/178L.
7) Electric Power Research Institute, "COMPBRN IIIE: An Interactive Computer Code for Fire Risk Assessment", EPRI NP-7282, May 1991.
8) USNRC, Accident Sequence Evaluation Program Human Reliability Analysis Procedure, NUREG/CR-4772, February 1987.
9) Electric Power Research Institute, "Fire PRA Requantification Studies", NSAC/181L, March 1993.
10) J. Lambright, et al., "A Review of Fire PRA Requantification Studies Reported in NSAC/181," Draft, Sandia National Laboratories, April 1994.
11) USNRC, "Regulatory and Backfit Analysis: Unresolved Safety Issue A-45, Shutdown.

Decay Heat Removal Requirements," NUREG-1289, November 1988.

17

t:...._' (

Addendum: Specific reference to NSAC/181 in the Salem submittal and comments

1. Page 4..:5: Reference to damage criteria for use in COMPBRN Ille; The damage criteria presented were somewhat more conservative than FIVE, e. g. 622 °F for qualified cable, rather than 700 °F.
2. Page 4-17: Electrical components in closed cabinets, such as adjacent to a fire-containing cabinet, fail at 322 °F. Relay data are offered as a basis. While some relays failed at lower temperature in actual tests, this value conservatively bounds the majority of observed failures.
3. Page 4-17: Exclude IEEE-qualified cable as an ignition source, from both self-ignition and transient welding sparks. The welding case is without experimental basis, but data indicate the need for piloted ignition with flame impingement.
4. Page 4-21: The main control room "critical cabinet" scenario methodology. There have been no objections raised to this method. The assumption of knowledge of a given cabinet's contents was questioned in the analysis of another plant.
5. Page 4-26: Fire in a cabinet fails all cabinet contents. Barriers, including single-walled dividers are respected by fire propagation, but not respected by the potential for component damage.. In general, this is not a good assumptio~. In this instance, use of the "fails all cabinet contents" attribute may compensate. Then propagation to adjacent cabinets becomes an issue and the effectiveness of suppression would be introduced into the analysis. This assumption was made only in the control room, which is occupied.
6. Page 4-28: The probability of manual non-suppression of the control room cabinet fires of 3.4E-3 was used. The submittal states that detectors are located both in-cabinet, and in the ventilation exhaust stream.

'7. Page 4-38: Fire propagation from battery chargers is not credible. In this specific instance, the absence of nearby and overhead combustibles was also noted.

18

a:.

I t

~ J I

SALEM GENERATING STATION INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS (IPEEE)

TECHNICAL EVALUATION REPORT HIGH WINDS, FLOODS, AND OTHER EXTERNAL EVENTS

TECHNICAL EVALUATION REPORT ON THE REVIEW OF THE SALEM NUCLEAR GENERATING STATION INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS (IPEEE) SUBMITTAL ON HIGH WINDS,. FLOODS, AND OTHER EXTERNAL EVENTS (HFO)

Brad Hardin, USNRC April 1999

  • ~

I

\\,L.,f

(

I J' ~J 2

Salem Nuclear Generating Station IPEEE Review Results High Winds, Floods, and Other External Events (HFO) 1.0 INTRODUqTION

  • Salem Generating Station (SGS) consists of two similar units with each unit having a Westinghouse 4-loop PWR rated at a thermal power of 3411 MW. SGS is owned and operated by the Public Service Electric Company (PSE&G), and is located on the east bank of the Delaware River in Salem County, New Jersey. Unit 1 received its operating license (OL) in 1977, whereas Unit 2 received its OL in 1981. Due to its construction schedule relative to the development of the 1975 SRP criteria, Salem is not considered to be an SRP plant; however, certain aspects of its licensing basis do conform to the 1975 criteria. In addition, SGS is located on the same site as the Hope Creek Nuclear Generating Station which is an SRP plant, and site-related information from the Hope Creek updated FSAR (UFSAR) that is not design dependent was utilized in the SGS HFO screening along with information from the SGS (UFSAR). Plant walkdowns confirmed that there were no plant changes since the issuance of the OL that impacted on the HFO related risk.

2.0 SCREENING OF EXTERNAL HAZARDS Beginning with the list of external events found in NUREG/CR-2300, the HFO-related external events have been screened out either by compliance with the 1975 Standard Review Plan (SRP) criteria or by bounding probabilistic analyses that demonstrate a core damage frequency (CDF) contribution of less than the IPEEE screening criterion (i.e., 1 E-06/reactor year (RY)).

The licensee used the progressive screening approach described in NUREG-1407 to screen external hazards and found that there were no other plant-unique external events that pose a significant threat of severe accidents within the context of the NUREG-1407 screening approach.

3.0 HIGH WINDS The design basis of SGS is consistent with the design criteria for winds given in NUREG-1407.

The UFSAR for SGS indicated that the safety-related refueling water storage tank (RWST) and auxiliary feedwater storage tank (AFST) were well anchored but not diked, and that they could therefore be susceptible to tornado missiles. An analysis was performed with the result that the contribution to core damage frequency from this susceptability was less than the IPEEE screening criterion. Non-Category 1 structures adjacent to Category 1 structures are designed not to collapse on the Category 1 structures. A walkdown was performed to evaluate hazards due to high winds. As a result, a design change is being made to Unit 2 to better secure the hydrogen tanks. Otherwise, it was concluded that no significant unreviewed plant changes exist that would impact the plant hazard or licensing bases regarding winds and tornadoes.

3 The licensee concluded that the contribution to CDF from high winds, including tornadoes, is less than 1 E-6/RY and is thus considered to be acceptable.

4.0 EXTERNAL FLOODS A plant walkdown was performed to evaluate possible paths of significant water ingress into safety-related structures. The walkdown revealed four possible in-leakage pathways, and a further evaluation was done to determine the need for any plant changes. As a result, improved penetration seals were installed between the Service and Auxiliary buildings that reduced the calculated core damage frequency contribution from approximately 1 E-4/reactor year (RY) to 1 E-7/RY. The licensee responded to a staff request for additional information regarding the surveillance of the improved seals saying that the integrity of the seals would be monitored by the use of periodic visual inspections.

The licens.ee also evaluated the effects of dam failures on the flood level. The licensee determined that dam failures contribute only nominally to the flood level because of the long.

distance between the site and the.dams.

A plant walkdown was performed and documented to identify any plant changes that could affect the plant's hazard data or the licensing base~ reg~rding flooding. The results of the walkdown indicated that there were no significant plant changes since receipt of the OL.

5.0 TRANSPORTATION AND NEARBY FACILITY ACCIDENTS Transportation and nearby facility accidents were considered, and no significant risk was found.

The only major transportation route within a 5-mile radius of the plant is the lntercoastal Waterway which is 1.5 miles west of the plant site, an acceptable distance for potential explosions. No major highways or railroads are located within 5 miles.

The licensee evaluated aircraft impact accidents, road and rail accidents, and fixed facility accidents, including industrial facilities, military facilities, and pipeline accidents. The effect of aircraft impact on the plant CDF was reviewed and was negligible. The contribution to core damage frequency of a ship or barge on the Delaware River colliding with the service water intake structure was evaluated to be negligible because of the low likelihood of a ship impacting the intake structure. Because of the distance from the shipping channel to the intake structure (approximately 1.1 miles), the hazard from an explosion in the channel was also judged to be negligible.

The licensee evaluated accidents involving a fire or an explosion from nearby facilities. The only significant facility within 5 miles of the site was a natural gas line, located about 4.5 miles from the site. The licensee determined that there was sufficient distance between the gas line and the site such that any fire or explosion from the gas line would pose no threat to the site.

The licensee also evaluated a release of toxic chemicals as a potential hazard. No facilities within 5 miles of the plant were identified as using or storing toxic chemicals. On-site use of

~---~~ *--- -*---------*--*----* ------*------~-----------**.---;°'"".. --:0--"-.......------------,.... --~-----* ----- ****- --** -*-

~

    • "-.'i' ;¥*1...,..-.

4 sulfuric acid, hydrazine and ammonium hydroxide for water treatment at the SGS and co-located Hope Creek Generating Station sites was evaluated against Regulatory Guide 1. 78 criteria, and it was determined that control room hability would not be impacted significantly by postulated releases of these chemicals.

The licensee has performed a confirmatory walkdown and determined that no significant changes were found to impact the plant's original design conditions regarding the hazard due to transportion and nearby facility accidents. No potential vulnerabilities were identified.

6.0 OTHER EXTERNAL EVENTS The effects of blockage from debris (detritus) on the circulating water intake at SGS have caused several plant shutdowns or power reductions in the past. As a result, actions have been taken to protect the intake through the use of blowdown fittings on screen wash headers, improved screen wash pumps and other improvements. It is now estimated that the blockage induced loss of service water is below the IPEEE CDF screening criterion.

7.0 GENERIC SAFETY ISSUES GSl-103, "Design for Probable Maximum Precipitation"

  • The licensee addressed GSl-103 in Section 5.5.3 of its submittal and determined that while the additional rainfall associated with this issue gave rise to ponding and additional roof loading, both Units 1 and 2 are adequately robust when assessed against the new PMP. The staff finds that the licensee's GSl-103 evaluation is consistent with.the guidance provided in Section 6.2.2.3 of NUREG-1407.

GSl-172, Multiple System Response Program" (MSRP)

The licensee has addressed the MSRP issues in its submittal and concluded that there were no vulnerabilities associated with these issues. Specifically, the licensee addressed the HFO-related issue "Effects of flooding and/or moisture intrusion on non-safety and safety-related equipment" in Section 4.8.1.2 of its submittal and concluded that there were no vulnerabilities at SGS due to this issue. Based on the staff review, the staff considers that the licensee's process is capable of identifying potential vulnerabilities associ~ted with GSl-172. Therefore, on the basis that no particular vulnerability associated with this issue was identified in the IPEEE submittal, the staff considers the IPEEE-related aspects of this issue to be resolved for SGS.

8.0 CONCLUSION

S Based on its review of the licensee's submittal and the response to the staff's RAI regarding monitoring of the improved penetration seals for the Ser\\tice and Auxiliary buildings, the staff has concluded that the SGS IPEEE submittal for the H.FO area is acceptable and meets the intent of Supplement 4 to Generic Letter (GL) 88-20.. *

~--,*----*-~-----

*----~**-.-..~~--~--

'

Retrieved from "https://kanterella.com/w/index.php?title=ML18107A319&oldid=4999079"