ML20135E245
| ML20135E245 | |
| Person / Time | |
|---|---|
| Issue date: | 03/01/1990 |
| From: | Taylor J NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | Carr, Curtiss, Remick, Roberts, Rogers K NRC COMMISSION (OCM) |
| Shared Package | |
| ML20127B259 | List: |
| References | |
| M891214, NUDOCS 9703060380 | |
| Download: ML20135E245 (50) | |
Text
r
).:
O (3 hu~4un g gf y
MAR 01 1990 I
gF M Fw l_ 'm(cmp MEMORANDUM FOR:
Chairman Carr g
o Commissioner Roberts R
Commissioner Rogers 4
Commissioner Curtiss Commissioner Remick FROM:
James M. Taylor hk 2
Executive Director for Operations 4
SUBJECT:
RESPONSE TO ITEM 3 0F THE STAFF REQUIREMENTS MEMORANDUM ON SEVERE ACCIDENT INTEGRATION PLAN (M891214)
Item 3 of the January 2,1990, Staff Requirements Memorandum on the Severe Accident Integration Plan requested the staff to submit an information paper describing the candidate accident management strategies, prior to issuing them to industry for information. The purpose of this memorandum is to provide the Commission with the requested information. A detailed description of the accident management strategies is provided in NUREG/CR-5474 (Enclosure 1).
1 ie a proposed generic letter supplement which will transmit these strategies to the industry for information. As described in the Accident Management Program Plan (Enclosure 3) and subsequent Commission guidance (Enclosure 4), transmittal of the strategies to licensees will be via a supplement to the IPE generic letter (Gena 'c Letter 88-20), on a schedule such that licensees have these insights to e.onsider in conjunction with development of their IPEs.
As part of conducting the Individual Plant Examination, licensees are encouraged to consider accident management strategies, such as those identified in the Generic Letter Supplement and NUREG/CR-5474, for applicability and effectiveness at their plants.
NUREG/CR-5474 includes a technical assessment of each of the accident management strategies, as well as evaluation guidance and cautions for each strategy to provide added assurance that use of the strategy will not detract from safety.
We have discussed the proposed generic letter supplement and a draft of NUREG/CR-5474 with the ACRS.
The ACRS provided suggestions and a favorable evaluation of the information (Enclosure 5).
We have incorporated their comments in both the Generic Letter Supplement and the NUREG/CR.
/ /
i D
i b
Contact:
G.N. Lauben, RES 492-3530 l
l C W.
D A eb Co#
060056 g
9703060300 900301 l
)
r (O
/^N
."The Commissioners b 2
h
,.The contents of the draft version of NUREG/CR-5474 were also reviewed on several occasions by the special interoffice PRA experts task force especially established for that purpose. The document was also reviewed by instructors at the NRC Technical Training Center and by contractors involved in related accident management programs. The draft version of NUREG/CR-5474 was placed in the Public Document Room.
Industry comments were requested to assure technical accuracy of any statements describing current industry practices and procedures.
This review was coordinated by NUMARC.
In general, those comments indicated the strategy report would be useful to utilities.
Some specific technical comments were also offered and we have incorporated their comments where applicable.
A package including the same information provided herein was forwarded to the Committee to Review Generic Requirements (CRGR) on November 28, 1989, for information only with the suggestion that a formal CRGR review was not needed j
since the Generic Letter Supplement (Enclosure 2) does not contain any new requirements for licensees.
Informal comments were received from CRGR members and incorporated into Enclosures 1 and 2.
Negative consent was received from CRGR on February 5, 1990 (Enclosure 6).
Since this package has received a thorough review, we believe it is now ready for distribution to individual utilities.
By copy of this memorandum, we are also sending a copy for information to each of the regional administrators.
Item 8 of the January 2, 1990, Staff Requirements Memorandum directed the staff to:
"Adviss the Commission if the Containment Improvements Program, the External Events Assessment Program, and the Accident Management Program slip i
significantly from current schedules." The original schedule for issuance of the subject Generic Letter Supplement was December 1989. Due to delays in. the review of the letter and attachment (Enclosures 2 and 1), issuance has been delayed until the present time. We anticipate issuing the Generic Letter Supplement and NUREG/CR-5474 by March 30, 1990.
Original Signed By:
1 James M. Taylor.'
James M Taylor Executive Director j
for Operations
Enclosures:
1 As stated 4
cc:
See next page.
- Previously concurred i
RPSB/DSR RPSB/6F J/DSR D/DSR D/DSIR-
"" 'R ben /
- Tlee
- LSho w 'JMurphy *BSheron *WMinners ks
/*GN 90 02/8/90 02/13/90 02/13/90 02/13/90 02/15/90 02//1 0
D/b PRAB/DREP PRAB/DREP D/DREP OGCB/D0EA A(l N
/
D R 02/20/90 'pCBerlinger
- FCongel j
j yos
' rd
- RPalla
- RBarrett 02g9 0
90 02/13/90 2/15/90 02/p/90 D/ DST [
I
[ MAS A
NRR EDdel7 AThadani s le F ia niezek "M
ey ylor 7
2g90 02 90
/f/90 02/'#90 02g 02 0
c J
~ _ _ _ _...... _ _ _ _ _... _ _ _ _
i l
l.
~ The Comissioner 3
i j'
y i-
. cc:
SECY
't l-OGC-j.
W. Hodges, RI l
W. Russell, RI t
S. Ebneter, RII A. Davis, RIII i
l R. Martin, RIV i
J. Martin,: RV j
1 i
i i
t
}-
i I
1 l
i.
{
k.
l i
l
)
t E
h I
t t
i I
I 1
l 1
i e
i, l
i 1.
i l
1 4
i 1
I
,'Th'e Commissioners 4
- ~ Distribution:
RES Circ /Chron DSR Chron NUDOCS wJIBRis EDO r/f.
ASummerour PDorm MBridgers-DMossburg RPSB r/f GNLauben r/f GNLauben LShotkin JMurphy BSheron.
WMinners
'TSpeis Dross EBeckjord RPalla RBarrett FCongel CBerlinger AThadani FGillespie FMiraglia JSniezek TMurley
-HThompson JB1aha EJordan RScroggins
- F.
JTaylor JRoe 4
i::
RErickson-RCJones GZech MTayl or..
JRosenthal.
KRaglin-WBeckner FCoffman MCunningham-FCostanzi DHouston RES NO:
893538 (#3, 8)
. WITS NO:
9000003
-j
E w closuv e. i 04L '
1 NUREG/CR-5474 BNL-NUREG-52221 l
ASSESSMENT OF CANDIDATE ACCIDENT MANAGEMENT STRATEGIES i
l W.J. Luckar; -Jr., J.J. Vandenkieboom* and J.R. Lehner w
i FEBRUARY 1990
{
I
==
Department of Nuclear Energy Brookhaven National Laboratory Upton, NY 11973
Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission
'ushington, DC 20555 FIN A-3964 u
a
~
z i
o j.
'g '
ABSTRACT r
4 A set of candidate accident management strategies, whose purpose is to prevent or mitigate in-vessel core damage, were identified from various NRC and j
industry reports.
These strategies have been grouped in this report by the challenges they are intended to meet, and assessed to provide information which l
may be useful to individual' licensees for consideration when they perform their Individual' Plant Examinations.
Each assessment focused on describing and ex-plaining the strategy, considering its relationship to existing requirements and practices as well as identifying possible associated adverse effects.
t
/
4 4
i r-'
1 N
- 9 O
i i
/
111
+&+e m
.s--..
,y----
.--,y q
l.
J fj
(
EXECUTIVE
SUMMARY
i Recognizing the risk reduction potential associated with' accident manage-ment, the NRC has initiated an accident management program aimed at promoting i
the most effective use of available utility resources to prevent 'and. mitigate severe accidents.
This' report contains an assessment of selected' candidate accident management strategies developed from information obtained from the
. NUREG-Il50 - analysis, NUREG/CR-4920, NUREG/CR-5132, other PRAs, and industry a
reports and articles pertinent to accident management.
The strategies were j
grouped according to the challenges they are intended to meet.
Some of the strategies reported on apply to BWRs or PWRs only, others apply to both types i
of plants.
The strategies described herein focus primarily on preventing or i
mitigating in-vessel core damage.
Strategies aimed at preventing containment failure and/or mitigating the release of fission. products to the environment are the subject of ongoing research and will be reported on in the future.
The assessment focused on describing and explaining each strategy, deter-i mining its relationship to existing requirements and practices as well as
~
j identifying possible associated adverse effects.- The reactor vendor developed generic emergency procedure guidelines antPthe emergency operating procedures
~
of sneral plants were examined to determine the extent to which these strate-gies may already be implemented in light of existing regulation and NRC/ industry 4
activities.
I This report provides licensees with a more complete description of selected ~
l accident manage:nent strategies as well as information that might.be useful to i
the licensees in assessing the feasibility of tha strategies for their plants.
1 The set of strategies discussed in this report is not meant to be complete or exhaustive. It is anticipated that other strategir.s 'important to the provention or mitigation of core damage may be identified by the licensees through the conduct of their Individual Plant Examinations.
i While all of the candidate strategies assessed in this report are believed to offer some benefit in terms of either prevention or mitigation of core damage accidents, some go beyond the traditional thinking which established the licens-ing design basis, and fall into the category of "last resort" measures.
It should be kept in mind that there is no recommendation made herein for changes to the current safety practices.
It is intended that the implementation of the strategies discussed herein serve to strengthen the defense-in-depth against severe accidents by possibly
. extending the existing emergency operating procedures beyond design basis situations without compromising plant safety. Hence, in reviewing these strate-gies for applicability to their plants, licensees should give careful considera-tion to the possible adverse affects.
/
v
[
i
- s*
a TABLE OF CONTENTS Pace 4
ABSTRACT................................................................
iii EXECUTIVE
SUMMARY
v ACKNOWLEDGMENTS.........................................................
ix i
l 1.
INTRODUCTION........................................................
1 1.1 Background........................................
1 i
i 1.2 Objective / Scope................................................
1 1.3 Relationship to Individual Plant Examinations..................
4 1.4 Organization of Report.........................................
4 i
2.
STRATEGIES RELATED TO INSUFFICIENT C00LANT..........................
8 2.1 Strategy to Reduce Containment Spray Flow Rate to Conserve Water for Core Injection (PWR).................................
8 2.2 Strategy to Enable Early Detection, Isolation, or Otherwise Mitigate the Effects of an Interfacing Systems LOCA (BWR and PWR)...........................................................
9 2.3 Makeup to Emergency Storage Tank...............................
11 i
2.3.1 Strategy to Refill Condensate Storage Tank (BWR)........
11 1
2.3.2 Strategy to Refill Refueling Water Stora
... Borated Water (PWR).....................ge Tank With 12 2.4 Strategy to Ensure Appropriate Recirculation Switchover and i
Manual Intervention Upon Failure of Automatic Switchover (PWR).
14 2.5 Strategy to Ensure Adequate Plant Heat Removal Capability by Emergency Connection (s) of Er.i ning or Alternate Water Sources l
(BWR and PWR)..................................................
14 3.
STRATEGIES RELATED TO UNAVAILABLE INJECTION SYSTEMS.................
16 3.1 Strategy to Extend Emergency Core Cooling System Availabilit by Switching Pump Suction (BWR).............................y 16 3.2 Emergency Bypass or Change of Pump Protective Trips............
17 3.2.1 Strategy to Enable Emergency Bypass or Change of Pro-tective Trips for Injection Pumps (BWR and PWR).........
17 3.2.2 Strategy to Extend Reactor Core Isolation Cooling System Availability by Pump Trip Function Bypass or Change (BWR)............................................
18 3.3 Core Injection by Non-Safety Rel ated Pump......................
19 3.3.1 Strategy to Use Control Rod Drive Pumps for Core Injection (BWR).........................................
19 3.3.2 Strategy to Use Non-Safety Related Char Core Injection (PWR)...................ging Pumps for 20 vii
I 3
{
S ~.
v I
p..
1 1
M 4
3.4 Strategy to Use Alternate Seal Injection (e.g., Hydrotest Pump When Reactor Coolant Pump Seal Cooling is Lost (PWR)..........)
21 l
~
3.5 Strategy to Use Condensate Pumps or Startup Feedwater Pumps 4
for Steam Generator Injection (PWR)............................
22 4
4.
STRATEGIES RFU4TED TO LOSS OF P0WER.................................
24 l
[
4.1 Strategy to Conserve Battery Capacity by Shedding Non-Essential
{
Loads (BWR and PWR)............................................
24 4.2 Strategy to Ufe P'ortable Battery Chargers or Other Power j
Sources to Recharge Station' Batteries (BWR and PWR)............
25 4.3 Strategy to Enable Emergency Rep %nishment 'of the Pneumatic l
l Supply for Safety Related Air Operated Components (BWR and PWR)...........................................................
26 i
(
4.4 Strategy to Enable Emergency Bypass or Change of Protective
. Trips for Emergency Diesel Generators (BWR and PWR)............
-27 I.
i l
4.5 Strategy to Enable Emergency Crosstie of AC Power Between Two 1 -
Units or to an Onsite Gas Turbine Generator'(BWR and PWR)......
28 l'
l 4.6 Strategy to Use 'a Diesel Generator or Gas Turbine Ger:erator to b
Power a Control Rod Drive or Other Appropriate Pump for Core
~;
1 Injection (BWR)................................................
29 4.7 Strategy to Use Diesel-Driven Firewater Pump for BWR Core Injection, PWR Steam Generator Injection or Containment Sprays j
(BWR and PWR)..................................................
30 5.
STRATEGIES RELATED TO LOSS OF HEAT SINK.............................
32 5.1 Strategy to Reopen Main Steam Isolation Valves and Turbine Bypass Valves to Regain the Main Condenser as a Heat Sink
[
(BWR and PWR)..................................................
32 6.
STRATEGIES RELATED TO SHUTDOWN FAILURE (REACTIVITY CONTROL).........
34
[
6.1 Strategy to Provide Additional Supply _of Borated Makeup Water
{
for Long-Term Accident Control (BWR and PWR)...................
34 6.2 Strategy to Inject Borated Water in Case of Potential Core Damage and to Guard Against Boron Dilution in the Core (BilR)... 35 f
j 7.
REFERENCES..........................................................
37 viii i
4 ACKNOWLEDGMENTS The authors would like to acknowledge the contributions of_ a number of individuals who provided appreciated assistance in this effort; from NRC Office of Nuclear Regulatory.Research
.T.M. Lee, G.N. Lauben, L.M. Shotkin and J.A.
Murphy for their support,-review and technical guidance; from the NRC Office of l!
the Executive Director for Operations - M.A. Taylor for his overall review and comments; from NRC Office of Analysis and Evaluation of Operational Data - K.A.
Raglin, S.D. Roessler, and S.K. Showe for the review and comments by 'ae Techni-cal Training Center staff and D.G. Marksberry for his constructive comments.
R. Tokarz, W. Scott and their support personnel at Pacific Northwest Laboratory should be acknowledged for their efforts in developing early versions of some of the strategies presented in this report.
Also, the Nuclear Management and Resources Council staff should be ac-knowledged for their tifnely response to the NRC's request to review and comment on a recent draft of this report.
The resultant comments, compiled responses by the four vendor owners' groups, were found to be very helpful to this effort.
A special note of appreciation is due to R.L. Palla, from the NRC Office of Nuclear Reactor Regulation, for his steadfast review resulting in valuable and timely comments on each of several draft versions of this report.
The authors are especially grateful to S.L. Flippen for her patience and exceptional effort in producing this report from its numerous draft versions.
O l
/
l ix l
[,
.g i
'l.
INTRODUCTION
'1.1 Backaround
.. The concept of defense-in-depth, the.high standards used for the design and construction of nuclear power plants,-and the training of the operating staff t
all contribute to the low risk associated with nuclear power plants. Neverthe-l less, experience obtained from the NUREG-11501 analysis, as well as other PRA analyses, has shown that changes in plant operating procedures and/or relatively l
i
. minor hardware modifications can reduce severe accident risk even further. This can often be accomplished by innovative use of existing plant equipment and has i
the added advantage of being cost effective when compared to risk reduction i
achieved as a result of major hardware addition to or modification of plant
[
systems. Accident management measures in the form of incremental improvements which extend existing Emergency Operating Procedures (EOPs) somewhat further into severe accident regimes and make the most effective use of available j
utility resources can effer the potential for such a risk reduction.
l The set of generic accident management strategies listed in Table I was identified and grouped from information obt.ained from the NUREG-1150 analysis, NUREG/CR-4920,2 3
NUREG/CR-5132 and other PRAs.
j i
NUREG-1150, " Severe Accident Risks:
An Assessment for Five U.S. Nuclear i
Power Plants" (Second Draft), documents a current assessment of the severe acci-dent risks of different plant designs.
Th'e report provides summaries of the i
risk analysis results of the-studied plants and perspectives on these results.
3-
- NUREG/CR-4920, Volumes 1 through 5, " Assessment of Severe Accident Preven-i tion and Mitigation Features," identifies plant features and operator actions which were. found to be important in either preventing or mitigating severe accidents'. These. features and actions were developed from insights derived from.
reviews of risk assessments performed on particular reference plants illus-trative of the.five different containment designs used.in U.S. plants.
NUREG/CR-5132,"SevereAccidentInsightsReport,"describestheconditions and events that nuclear power plant personnel may encounter durirg the latter stages of a severe core damage accident and what the effects and consequences i
might be due to actions they may take during these latter stages.
The report also describes what can be expected of the performance of the key barriers to fission product release (primarily containment systems), what decisions the operating staff may face during the course of a severe accident, and what could i
result from these decisions based on the current state of knowledge of severe I
accident phenomena.
{'
1.2 Ob.iective/ Scone L
This report is intended to provide licensees with an assessment of a set i
of candidate accident management strategies as well as information that might i
be useful to the licensees in evaluating the feasibility of the implementation of one or more strategies for their plants. The relationship with existing re-i quirements and practices as well as possible adverse effects associated with the strategy use are also identified. While the applicability of each strategy 1
is likely to be plant specific, these strategies could be useful to utilities j
in their consideration of developing a capability to respond to accidents.
i, k
~
t 1
J e
m,
.c,v...--,
)
n Appropriate sections of the reactor vendor developed generic emergency procedure guidelines *' and E0Ps from a number of plants were used to examine the extent
~
to which these strategies may already be in place as a result of existing regulation and NRC/ industry activities. These plants are Calvert Cliffs, Grand Gul f, LaSalle, Limerick, Oconee, Peach Bottom, Seabrook, Sequoyah, Surry, Susquehanna, Trojan and Zion.
While many of these strategies are implemented in existing E0Ps and other plant-specific operating procedures and instructions, j
the degree to which this is done varies widely across the industry. Therefore, l
this report provides generic information that may be useful for enhancing some procedures in order to take advantage of existing backup systems and components which could be made available during certain accidents.
This document does not include information on the risk reduction that might be achieved with implementation of individual strategies assessed herein since the actual risk reduction potential of a given strategy is highly plant speci-fic.
Examples of quantitative plant specific risk reductions can be found in such references as NUREG/CR-5263,8 "The Risk Management Implications of NUREG-1150 Methods and Results," and Supplement No. 4 to NUREG-0979,8 the GESSAR II Safety Evaluation Report.
This information may be helpful in evaluating the potential implementation of one or more oF-these strategies.
An essential part of each strategy considered by the licensee should be an evaluation of how a strategy may affect plant equipment and operators during both normal and accident conditions.
Included in this evaluation process, may be the following examples of operational implementation considerations across various accident conditions:
j Hardware considerations:
- anticipated accident conditions which would influence equipment operability:
- pressure,
- temperature, and
- radioactivity levels.
- actual capabilities of existing hardware which might be used to backup failed safety systems:
- water supply, (e.g., available tank, sump, pool inventory),
- flow rates of alternate piping configurations to supply coolant, and
- water quality (e.g., borated versus unborated, raw versus treated).
Operator considerations (human factors):
- added burden placed on operators and other plant staff,
- adequacy of existing instrumentation,
- need to bypass or change trip setpoints, and
- habitability of areas which need to be accessed.
While all of the candidate strategies assessed in this report are believed to offer some benefit in terms of either prevention or mitigation of severe core damage accidents, some go beyond the traditional thinking which established the licensing design basis, and fall into the category of "last resort" measures.
2
f r
It must be kept in mind that there is no recommendation made here for changes i
to the current safety practices.
It is the intent that the implementation of the strategies discussed herein should serve to strengthen the defense-in-depth against severe accidents by possibly extending the existing E0Ps beyond design basis situations without compromising plant safety.
It should also be noted in considering these strategies that there are some possible adverse effects associated with most, if not all, of them.
Examples are the additional burden on operators and the possible drawbacks arising from the use of non-safety grade equipment on safety-related systems. Some of these adverse effects may be minimized by taking sufficient preparatory measures for certain strategies in the form of preparing cables, adaptors, jumpers, spool l
pieces, developing procedures for their use and training using the procedures.
Some licensees may even want to train selected operating personnel and/or shift supervisors to deal with severe accident situations. The most successful acci-dent management program would be one which makes maximum use of the existing j
human and hardware resiiu'rces at the plant to maximize the effectiveness of accident prevention and mitigation while at the same time keeping costs and adverse effects to a minimum.
r The strategies listed in Table 1 were rearranged and combined into twenty candidate strategies grouped by safety objectives and challenges they are intended to meet such as insufficient coolant, loss of power and loss of heat i
sink as shown in Figure 1.
This arrangement is helpful for recognizing possible relationships among the strategies.
Some strategies apply to Boiling Water Reactors (BWRs) or Pressurized Water Reactors (PWRs) only, others apply to both types of plants.
The strategies described focus primarily on preventing or mitigating in-vessel enre damage.
Strategies aimed at preventing containment failure and/or mitigating the release of fission products to the environment will be addressed as part of ongoing NRC research and completed assessments of such strategies will be documented as appropriate.
The logic of Figure I should be helpful to utilities undertaking a sys-tematic assessment of their accident management capabilities.
An attempt has been made in each section to group the strategies in the order in which it is felt that the plant staff would most likely implement them, starting with conservation or improvement of normally used supplies or systems and ending with attempts to use alternate sources or systems. This is only a general grouping and the actual order of implementation would be accident scenario and plant dependent.
Many of the strategies will be more effective when used together with one or more of the other strategies, and under certain conditions might only be applied when another strategy is also implemented. All of the strate-gies are meant to preserve the two safety objectives of maintaining core cool-ing, and reactivity control.
Threats to these safety objectives are grouped into challenges, and these challenges are addressed by the accident management strategies. The safety objective of maintaining core cooling can be challenged by insufficient cooling, unavailable injection systems, power loss, and/or heat sink loss. The other safety objective of reactivity control can be challenged by the failure of the reactor to shut down or by a core damage accident that results in recriticality.
f It is also important to realize that the set of strategies discussed in this report is not meant to be complete or exhaustive.
Other strategies 3
0 (v\\
V important to the prevention or mitigation of core damage may be identified by licensees through the conduct of their Individual Plant Examinations (IPEs) or during the course of their own accident management assessment.
1.3 Relationshio to Individual Plant Examinations Generic Letter (GL) 88-20 issued by the NRC staff for the IPEs emphasized the importance of accident management. The strategies discussed in this report are presented as information which may be useful to individual licensees when they perform their IPEs.
These strategies may aid the nuclear industry in taking cost effective, useful steps to further reduce risk from severe accidents by maximizing the use of existing resources. The strategies can be considered as an adjunct to the ones which may be identified through the IPE.
1.4 Oraanization of Report Figure 1 shows the logic structure of the strategies and gives an overview i
of the organization of the rest of the report. The remaining sections describe the strategies in the order as indicated in the figure.
Section 2 discusses strategies that deal with the challengen f insufficient coolant, Section 3 e
contains strategies which are concerned with the uravailability of injection i
systems, Section 4 addresses power loss strategies, and Section 5 assesses one strategy related to heat sink loss. The strategies contained in Section 6 are meant to meet the challenge of reactor shutdown failure.
9 i
l
/
1 4
i
.-- ~. -
~
Table 1 Generic Accident Management Strategies I.
Conserving and Replenishing Limited Resources Refill refueling water storage tank (RWST) with borated water, or condensate storage tank (CST) with condensate. Assure adequate supply of boron on site.
Maintain emergency core cooling system (ECCS) suction to condensate systems to avoid pump failure due to high suppression pool temperature.
4 Throttling containment sprays to conserve water for core injec-tion.
j Conserve battery capacity by shedding non-essential loads.
Use of portable battery chargersTr other power sources to re-charge batteries.
Enable emergency replenishment of gas supply, or otherwise ensure operability of air operated components.
4 Enable early detection, isolation, or otherwise mitigate the effects of an interfacing systems loss of coolant accident (LOCA).
II. Use of Systems / Components In Innovative Applications
. - Strategies to enable emergency use of available pumps to 1
accomplish safety functions.
Use of diesel fire systems for injection to the containment sprays, a BWR core, or the PWR steam generators (SGs).
Use of control rod drive (CRD) pumps in BWRs or charging pumps in PWRs for core injection.
Use of alternate injection (e.g., hydro test pump) when reactor coolant pump (RCP) seal cooling is lost (seal failure concern).
Enable emergency crosstie of service water and closed (component) cooling water (CCW) to residual heat removal (RHR) in BWRs or feedwater in PWRs.
Use of condensate, or startup pumps for feedwater injection.
5
l' V
i
{
- ,f 4
Table 1 (Cont'd)
?-
Strategies (and hardware) to enable emergency connection of available-AC power sources to meet critical safety needs.
Use of diesel generator or gas turbine generator to drive CRD pumps for core injection.
1 Enable emergency crosstie of AC power between two units or to onsite gas turbine generator.
).
Strategies to enable emergency connection of injection systems j-to alternate water sources.-
I Ensureappropriaterecircu5a'ionswitchoverandcopewith t
j the failure to switch over.in LOCA.
I Enable emergency connection ofservice water or feedwater systems to rivers, reservoirs or municipal water systems.
. ' Strategies for Reactivity Control.
Initiate standby liquid control system (SLCS) in case of potential core damage and guard against boron dilution when 2
core injection is restored.
1 Ensure abunda'nt supply of borated makeup for long-term accident control.
4 III. Defeating Interlocks and Component Protective Trips in Emergencies.
Reopen main steam isolation valves (MSIVs) and turbine bypass valves to regain the condenser as a heat sink.
Extend reactor core isolation cooling (RCIC) availability by 3
either raising the turbine exhaust pressure trip set point, or overriding the trip function.
Enable emergency bypass of protective trips for diesel generators and injection pumps.
n i
i 1
-6 1
-,,--e
O
\\
~
V
(-
4 ELATED TO INSUFFICIENT COOLANT Reduce Containment Soray Flow Rate to Conserve W t T
4R) a er for Core
,Ig is strategy is to conserve refueling water stor
. d core injection, if emergency coolant sump recircul ti tducing containment spray (CS) flow rates age tank (RWST)
=
cne or more of the following:
a on is This strategy can te or more totally redundant spray traithrottling the CS discharge T
e discharge flow back to the RWST ns, and/or recirculat-led in Subsection ?.3.2, " Strategy to Refill R f'ht be via a test line e
Another i torated Water."
erve RWST e ueling Water heir design values during accidents.f the CS system is to ma pressure and tdundant spray trainw-each with its owIn most PWRs, this is
! natic initiation on high containment pressure,p, valves and n pum e RWST and ty headers. pump on the order of several thousand gallons per the CS pumps 1p level in a few plants), emergencyIf a low RWST level is reached
' reral plants examined if long-term CS is coolant recirculation iR) pumps can be used to divert a portiorequired the resid-o supply several spray h n of the coolan i the RHR heat exchangers.eaders with suction from the cont ng a CS pump to takt suction from and discharge back tA CS pump p, ". " """a" o the e
irawdown rates associated with full flow CS i flow desirable in certain accidents operation may those LOCA sequences where containment presbut n 4
In particular, this sure is high
\\
all redun-r Otherwise Mitigate the Effects of an Interfacing S
(
to Ensure Appropriate Recirculation Switchover arly ystems
}
l lure of Automatic Switchover " a id 61 i
and Manual
- .~
ply of Borated Makeup for Long,-Term Accident Cont
=
" Strategy to E"""
rol."
int Reouirements and Practices ncy procedure guidelines of the domestic PWR vendo se i Procedures (EOPs) of many PWRs were examined to d t ch this strategy has been implemented.
rs and e er-ow rates was found in many of the plant E0Ps examia The use of a ned.
ng containment pressure as a guide.
e containment fan coolers was also.specified to r din some E0Ps the or w,
1 e plant examined, the ability to achieve multiple flo e uce w
8
i t
i p.
rates in the CS trains was indicated.
The actual procedural steps for ac-l complishing this were not given.
While none of the E0Ps were found which addressed the recirculation of a i
portion of the CS pump discharge back to the RWST, the. ability to accomplish this may exist in some plants, via an existing pump flow test line. The ability I
and desirability to run reduced flows through the CS test line is plant and accident specific and would need to be examined individually for each plant, 3
i Possible Adverse Effects i
Reduced CS flow rates could result in higher containment pressures and L
temperatures due to reduced spray coverage and spray droplet atomization at lower spray header pressures. If the accident has reached the fuel damage stage i
.there would also be less scrubbing of fission product aerosols from the contain-ment atmosphere with reduced sprays.
The o pressure and increase CS flow, if necessary.perators would have to monitor the The ability to reduce CS flow rates effectively by valve throttling appears limited in most plants that have gate vakes rather than globe valves.
- Also, the ability to reduce CS flow rates by pump stoppage may introduce the pos-sibility of restart failure.
2.2 Strateav to Enable Early Detection. Isolation. or Otherwise Mitiaate the Effects of an Interfacina Systems LOCA (BWR and PWR)
'Strateav Descriotion The aim of this strategy is to limit the effects of an interfacing systems LOCA (ISL) by early detection and isolation or if isolation is unsuccessful, with additional actions to mitiqate the consequences. An ISL involves the loss of isolation between high and ' ow pressure interfacing systems, and overpres-surization of the low pressure system. The resultant breach of the low pressure system outside of containment results in a LOCA which bypasses the containment.
Early detection and recognition of such :n event is an important first step for achieving possible mitigation, if not isolation, of the ISL.
Isolation of the failure may be possible and would halt the progress of this kind of accident.
If isolation cannot be achieved, various other actions may be of use in mitigat-ing the effects of an ISL.
The primary indicators of an ISL would be abnormal pressure, temperature and radiation measurements in different areas outside of containment.
Where
'available, the correlation of information from valve position indicators and line flow rates, pressures, and temperatures could also aid in the identifica-tion of an ISL.
Decreasing inventory levels of the reactor coolant system or the refueling water storage tank (RWST) along with a lack of corresponding increase in containment sump level are othe* nnssible indicators of an ISL in a PWR. In a BWR, the reactor coolant inver!
And pressure along with conden-sate storage tank (CST) and suppression gn levels would be important in-dicators.
/
The isolation of some ISLs may be possible because a number of valves exist in many lines which can be closed to compensate' for the break.
However, in order to isolate an ISL, the operators must be able to pinpoint its location.
9
I
,f t
j
\\v>
v Therefore, isolation will depend in some measure on plant instrumentation, e.g.,
pressure indication and alarms in key lines, and on the operator's ability to accurately detect the break location.
Additional training may improve the operator's ability to detect and isolate an ISL.
In some cases where isolation has failed, mitigation of the effects of an ISL may be possible by manual actions which might be proceduralized or described in guidance called out by the E0Ps, e.g., flooding the location of the break in the low pressure system. The submergence of the break will provide some scrub-bing of fission products and mitigate releases to the environment. The use of sprays can also help reduce the concentration of fission product aerosols; existing spray systems in the auxiliary building (i.e., fire sprays) are pos-sibilities.
The depressurization of the reactor vessel may also mitigate the effects of an ISL by reducing the mass flow rates out of the break.
In some cases, the judicious use of available pumps to manage the drawdown rate from the RWST or CST could be appropriate in delaying the onset of core uncovery and damage. See the related strategies listed below.
The NRC has a related ongoing program that will investigate in detail the issue of ISL. More insight on this issue-4hould result from the completion of this program.
Other related strategies include Subsections 2.1,
" Strategy to Reduce Containment Spray Flow Rate to Conserve Water for Core Injection," 2.3.I,
" Strategy to Refill Condensate Storage Tank," 2.3.2, " Strategy to Refill Refuel-ing Water Storage Tank With Borated Water," 3.3.1, " Strategy to Use Control Rod Drive Pumps for Core Injection," 3.3.2, " Strategy to Use Non-Safety Related Charging Pumps for Core Injection," and 4.7,
" Strategy to Use Diesel-Driven i
Firewater Pump for BWR Core Injection, PWR Steam Generator Injection or Contain-ment Sprays."
Relationshio With Current Reouirements and Practices The generic emergency procedure guidelines of the domestic BWR and PWR vendors and Emergency Operating Procedures (EOPs) of several plants were found 3
to have some procedures which addressed ISLs.
These procedures include steps which typically check the major containment isolation valves in a systematic manner by opening and closing (cycling) each valve individually while monitoring for a reactor coolant system pressure increase.
All of the E0Ps reviewed for PWR plants were found to address accidents involving steam generator tube rupture (SGTR), a specific ISL that is often addressed as a separate issue.
j Possible Adverse Effects i
Attempts to isolate an ISL can lead to an aggravation of the accident if the wrong valves are operated / cycled or vital systems shutdown in an attempt to contain the leakage.
In particular, the cycling of valves to diagnose and locate an ISL which then fail to close (or reclose) may add to the leakage.
Mitigative actions such as flooding the break location or using auxiliary building sprays may impact the performance of other systems located in that environment, e.g., shorting of electrical systems, etc.
10
lA 4
V y/
2.3 Makeuo to Emeroency Storaae Tank 2.3.1 Strateov to Refill Condensate Storaae Tank (BWR)
Strateav Descriotion The aim of this strategy is to supply additional water to the condensate storage tank (CST) to help avoid or at least delay depletion of the tank. This strategy would augment the CST water capacity and therefore reduce the risk of core damage in events such as extended station blackouts or LOCAs involving i
failures which render the suppression pool (SP) unavailable as a supply for reactor injection.
This strategy is accomplished by refilling the tank with treated water from other onsite sources.
In the event that sources of treated water are not available, other sources could be considered.
Replenishing CST water may be accomplished by normal. plant operating pumping systems, by gravity drain, by manual operation, or by using pumping systems independent of station AC or DC power supplies. Possible treated water sources might be: the deminer-alized water storage tank, main condenser hotwell, or the fuel pool. In plants with multiple units where cross connections exist, treated water could be drawn from storage tanks maintained for the seeend unit.
Possible untreated sources might be:
the plant firewater system, a community fire pumper truck, or a municipal water supply.
By design, turbine-driven high pressure core injection / spray (HPCI/HPCS) or reactor core isolation cooling (RCIC) pumps supply CST water to the reactor vessel injection systems in the event of loss of normal high pressure cooling water sources (e.g., main feedwater).
In a station blackout or certain LOCAs, high pressure injection into the core could be maintained for long periods of time by replenishing the water in the CST.
In some plants, the HPCI and RCIC systems might be manually operated under acceptable radiation levels thus allowing high pressure injection to be maintained in those cases after the loss of station AC power. These turbine-driven pumps can maintain rated flow at very low vessel operating pressures.
Other related strategies include Subsections 3.3.1,
" Strategy to Use Control Rod Drive Pumps for Core Injection," and 4.7, " Strategy to Use Diesel-Driven Firewater Pump for BWR Core Injection, PWR Steam Generator Injection or Containment Sprays."
Relationshio With Current Reauirements and Practices The generic emergency procedure guidelines of the domestic BWR vendor and the Emergency Operating Procedures (EOPs) of several plants were examined to determine the extent to which this strategy has been implemented.
The ability to provide CST makeup during an accident was found in the E0Ps of several plants.
In several of these instances, makeup is provided via a cross-connect to another unit on site.
Possible Adverse Effects Using any firewater pumping system could impact the ability of the fire-water system to respond to a fire. The use of a plant firewater system may also limit its capability to be used to control radionuclides that may be released into reactor buildings during the late stages of a severe accident.
11
j p
i V
4 The use of any untreated water source has a certain potential for plugging
~
the injection pathways to the reactor core and may inhibit complete closure of related valves when the injection is terminated.
If CST refill rates were significantly greater than emerger.cy depletion rates, the possibility of tank overpressure may exist.
Likewise, overfill of the SP may occur if other source (s) of injection are used in addition to the CST. This-increased SP level may adversely effect containment venting.
Drawing water from a cross-connected unit could adversely impact the unaffected unit.
Refilling the CST from a municipal water system raises a possibility of backflow of contaminated water'to the municipal water supply.
2.3.2 Strateay to Refill Refuelina Water Storace Tank With Borated Water (PWR)
Strateav Descriotion t
The aim of this strategy is to supply additional borated water to the refueling water storage tank (RWST) to help avoid or at least delay the deple-tion of the water in the tank. The water may be required to respond to certain sizes and types of loss of coolant accidents (LOCAs, if emergency coolant sump recirculation is not available, where other sources)of water are unavailable or less desirable to supply the requirements of core injection and possibly con-tainment spray.
The possible sources of water having sufficient boron con-centration to maintain an appropriate reactor safe shutdown margin might be:
normal RWST makeup (limited capacity), borated water holdup tank (possibly limited capacity), spent fuel pool (above fuel ass'emblies), unaffected unit's RWST (for multi-unit plants) via cross-connect (assuming appropriate measures being taken with the unaffected unit to compensate). Most, if not all, of these existing sources require AC power to pump the water.
i The RWST (alternately called the refueling water tank or borated water storage tank) was designed as the initial source of borated water for the emergency core cooling system (ECCS) and containment spray system during a LOCA.
The RWST will supply all of these emergency flow requirements until the tank is almost empty.
If there is sufficient coolant discharged into the containment sump by that time, then there can be either an automatic or manual transfer of ECCS and containment spray pump suctions to the sump. This ECCS mode is called either containment sump recirculation or em.ergency coolant recirculation.
Depletion of the RWST is of concern when the design capability of the ECCS and contai.nment spray pumps to take and maintain suction from the containment recir-culation. sump may not be achieved.
For example, sump recirculation ~would not be available if equipment malfunctions or sump blockage were to render all redundant recirculation trains inoperable, or if the water inventory accumulated j
in the sump were insufficient.
i This refill strategy taken by itself addresses those LOCA situations where i
both, (1) the break is large enough that the RWST is in jeopardy of being emptied while still needed for emergency injection, and (ii) there is concern that sump recirculation will not work (e.g., interfacing systems LOCA to outside containment). This strategy would be ineffective for large break LOCAs in which the emergency flow requirements may be much greater than the refill capability.
12
0\\
U v
m,'
' tre Acorooriate Recirculation Switchover and Manual inter-lure of Automatic Switchover (PWR) pr 1
strategy is to assure that a recirculation flow path exists ecoolingsystem(ECCS)andthecontainmentspray(CS)when storage tank (RWST) supply reaches its required switchover event, and/or the water level in the containment recircula-
.pecified level. This strategy is accomplished by assuring recirculation switchover and to cope with an automatic nen required by manual intervention.
u nside containment, the water lost from the reactor coolant ow into the containment sump. During the initial phase of for core injection is supplied from the RWST.
When the intainment sump and/or RWST reach (es) a specified level, the y be switched over either remotely or locally from the RWST water sump.
When sump water is recirculated (emergency ulation), the remairekg water inventory in the RWST is
strategies include Subsections 2.1,
" Strategy to Reduce Icw Rate to Conserve Water for Core Injection" and 2.3.2, Refueling Water Storage Tank With Borated Water."
arrent Reauirements and Practices ergency procedure guidelines of the domestic PWR vendors and
{"""*"'""'+
ting Procedures (EOPs) of several plants were examined to t to which this strategy has been implemented.
Procedural operator (s) to assure the switchover were found in the E0Ps ined including manual backup to automatic failure.
fects
'er (outside the control room) to emergency coolant sump 3 resent problems associated with valve locations being in tion areas.
[
"~
nsure Adeauate Plant Heat Removal Capability by Emeroency b
of Existina or Alternate Water Sources (BWR and PWR) h er s strategy is to ensure an adequate long-term supply of water
..,i.
. inventory and remove heat from the reactor and other plant
- egy could be implemented during an accident in which all
.er supplies and systems are unavailable or inadequate. This lish:d by providing backup emergency connections such as:
vice water (SW) from existing or alternate sources including
' ^ " -
tervoirs, municipal water systems, ocean, etc.; SW supply Idwater (or condensate) system. Actual hard-piped crossties 14
~
J between systems needed to implement this strategy are unlikely to exist in most plants.
The alternative would be to utilize a temporary hose connection ar-rangement.
Although such an arrangement would depend on specific plant con-figuration, it is likely that some plants have a penetration or blank flange that could be adapted by a hose connection. The proposed connected systems most probably will require AC power and sufficient pumping capability to deliver adequate supplies of cool water. Therefore this strategy may be affected by a station blackout.
I The SW system takes suction from an adequate source of water such as a river, lake, ocean or cooling tower basin and provides cooling to all plant loads during reactor shutdown. During normal reactor operation, most heat loads are handled by the main turbine generator, feedwater heaters and the main con-denser via the circulating water system. The SW system handles other loads such
?
as pump cooling and spent pool cooling.
I Other related strategies include Subsections 3.5, " Strategy to Use Conden-l sate Pumps or Startup Feedwater Pumps for Steam Generator Injection," and 4.7,
" Strategy to Use Diesel-Driven Firewater Pump for BWR Core Injection, PWR Steam Generator Injection or Containment Sprayst!L Relationshio With current Reouirements and Practices The generic emergency procedure guidelines of the domestic BWR and PWR vendors and the Emergency Operating Procedures (EOPs) of several plants were examined to determine the extent to which this strategy has been implemented.
Procedural steps for parts of this strategy exist in a number of plants.
Possible Adverse Effects The alternate water supply selected by this strategy may not be adequately 4'
filtered, therefore, in cases where this water is used for reactor core or steam generator injection, the injection path may become partially obstructed with debris especially in the feedwater spargers.
The impurities in untreated water pumped into a reactor core or steam generator could potentially lead to somewhat higher radiation levels in the system. Connecting the plant systems to rivers, reservoirs or municipal water systems could open these latter systems to possible contamination.
Service water supply directly to the feedwater (or condensate) system may reduce cooling water supply to normal SW loads.
/
15
r
(
O/
Gi C
~
3.
STRATEGIES RELATED TO UNAVAILABLE INJECTION SYSTEMS 3.1 Strateay to Extend Emeroency Core Coolina System Availability by Switchino Pumo Suction (BWR)
Strateov Descriotion i
The aim of this strategy is to extend the availability of the emergency core cooling system (ECCS) function by switching the suction cf the associated pumps from the suppression pool (SP) to an alternate condensate inventory. The transfer of the suction from the SP to a cool alternate source of ELC; water may be appropriate in response to some types of events, such as loss of coolant accident (LOCA), long-term station blackout and anticipated transient without scram (ATWS).
In these events, the SP temperature may become high erough to risk loss of emergency pumps due to accelerated wear or inadequate net positive suction head.
To accomplish this strategy, possible sources of cool uater to the suction of the ECCS pumps might be:.the condensate storage tank (CET), the main condenser hotwell via the main condensate or condensate transfer system, spent fuel pool or any other large quantity of cool water which can be accesad by either permanent connections or temporary hookups including hose connections.
Most of these sources require AC power to pump the water to the ECCS pump suction.
To accomplish this switchover, valve interlocks may have to be by-passe 3 or changed.
In most BWRs, the high pressure ECCS pumps, namely, high pressure coolant injection or core spray (HPCI or HPCS) and reactor core isolation cooling (RCIC)
{
initially take suction from the CST (preferred source) and later switch to the SP. The low pressure ECCS pumps, namely low pressure coolant injection (resid-ual heat removal) and low pressure core spray (LPCI and LPCS) normally take i
their suction from the SP with limited (or no) ability to readily use the CST j
as an alternate.
The condensate transfer system, usually in conjunction with.
the condensate system, link the main condenser hotwell with the CST thus provid-ing the ability to transfer water both ways between the hotwell and the tank.
In addition, in at least one plant, there is piping whic.h allows the condensate transfer system to provide a limited supply of hotwell water to the suction of the low pressure ECCS systems.
Other related strategies include Subsections 2.3.1,. " Strategy to Refill Condensate Storage Tank," 3.3.1, " Strategy to Use Control Rod Drive Pumps for Core Injection," and 4.7, " Strategy to Use Diesel-Driven Firewater Pump for BWR Core Injection, PWR Steam Generator Injection or Containment Sprays."
4 Relationshio With Current Reouirements and Practices The generic emergency procedure guidelines of the domestic BWR vendor and the Emergency Operating Procedures (E0Ps) of several plants were examined to determine the extent to which this strategy has been implemented.
In the E0Ps examined, there were no procedural steps found that directed the switch back of any high pressure ECCS pump suction from the SP to the CST.
i i
16 J
V Possible Adverse Effects The rising SP water level associated with the suction transfer back to the CST could be of concern as long-term containment performance might be diminished if it is not corrected.
3.2 Emercency Bvoass or Chance of Pumo Protective Trios 3.2.1 Stratecy to Enable Emeraency Bvoass or Chance of Protective Trios for In.iection Pumos (BWR and PWR)
Strateay Descriotion The aim of this strategy is to enable continued injection pump operation beyond the point where they would normally trip, with the intention of prevent-ing or mitigating an accident before the pumps fail.
This strategy is ac-complished by bypassing certain protective trips or changing trip setpoints on injection pumps unless this could result in early failure of the pumps.
The following identifies those injee64on pumps by their plant system and possibly related trips which might be considered to be of some benefit if bypassed or changed during an accident condition:
In BWR plants, the pumps are associated with the following systems, namely, reactor feedwater, high-pressure core injection (HPCI), high-pressure core spray (HPCS), low-pressure coolant injection (LPCI) mode of RHR, low-pressure core spray (LPCS) and control rod drive (CRD).
Examples of trips are high turbine exhaust pressure, high reactor water level, low steam supply pressure, low pump suction pressure, low lube oil pressure, low control oil pressure, thrust bearing wear, low oil tank level, high bearing vibration, electrical trips, in-line valves not full open and high steam line flow.
In PWR plants, the pumps are associated with the following systems, namely, charging and high and low pressure safety injection, main feedwater, auxiliary feedwater and condensate (plus condensate booster or heater drain).
A few i
examples of associated trips are high turbine exhaust pressure, low pump suction pressure, high steam generator level, low steam supply pressure and high steam line flow.
The trips mentioned above for BWRs and PWRs may not be all inclusive, how-ever, they are representative of those at most plants.
One or more of these trips may be considered for potential bypassing under emergency conditions. An assessment of each trip considered for bypass should be performed as part of the strategy evaluation process.
This assessment should include detailed informa-tion on the original design requirements for each trip and analysis of potential accidents in which these trips might be bypassed. The assessment should include beneficial attributes as well as detrimental aspects of bypassing individual trips.
Other related strategies include Subsections 3.2.2, " Strategy to Extend Reactor Core Isolation Cooling System Availability by Pump Trip Function Bypass or Change," and 4.4, " Strategy to Enable Emergency Bypass or Change of Protec-i.ie Trips for Emergency Diesel Generators."
17 i
i
O (O
i.
V
()
Pelptignship With Gyrrent Requirement $
vendors and the Emergency OperatinThe generic gnd Practices
\\
examined to determine the extent to g Procedures (EOPs) of sever For the plants reviewed, no pro which this strategy has been impl n
PWR cedural steps utilizing this strateg Possible Adverse Effect n s were emented.
l y were found.
detrimental and would possibly rBypassing protectiv lance and dependence on the adeq esult in the need for constant ope
~
expected to provide needed short te lead to extended loss in the long ould be uacy of existing instrumentationrator vigi-er term and risk substantial pump This is I
3.2.2 Strateoy to Extend Reactor Core Is l y could PUmo Trio FunctdM3yna1S_qr Chtnne (BWo amage.
Strateov Description n
R)
Y The aim of this strategy is to en bl
,ij cooling
-s.
intention (RCIC) system operation beyond which the a
ji complished by bypassing cof preventing or mitigating an a pump would trip with the more trip setpoints, (e.g.ertain RCIC pump protective tri on c dent.
in early failure of the pum,p or itturbine exhaust pressu This 1
strategy is ac-changing RCIC pump trip setpoint(s) s steam turbine.
strategy of pump trip bypass The s could result is treated separa. strategy of bypassing /
reduction potential is perceive (d to b Subsection 3.2.1) because the as tely from the broader e greater.
vessel to cool the core should the vesThe RCIC sy sociated risk k.
normally takes suction from the cond sel be isolated.a ntain sufficient w y
turbine to maintain pump flow and thmain feedwa ensate storage tank and discharges i tits e
r The turbine is designed to trip aut o the vessel.
i e exhausts to the suppression pool (S noa such as pressure,high turbine exhaust pressure,tically on vario oma pump suction pressure and turbine ovehigh steam flow, 4
RCIC system isolation (e.g ons e in various locations,,etc.), lowlow steam rspeed.
(SBO) situation where continued RCICDuring an anti
_(
water level, thereby preventing core u operation is(ATWS) or a sta out scram be beneficial to bypass or change ncovery and possible core damageneeded For example eventually r,equire controlled reactorthe high SP temperatures one or more present turbine trip set
, it may rises to satisfy the SP heat capacity li ated with an ATWS or SB0 will points.
reduced reactor pressure may cause a RCIC pressure reduction as the SP temper t limits are reached. exhaust pressure or low steam pressu mit.
turbine trip from either high turb a ure account the accident conditions and this significan This is true re before appropriate RCIC operabilit ne These operability limits should t E
urbine exhaust pressure y
e into mage.
18
I- (, ?.
l s
l Another related strategy includes Subsection 3.2.1, " Strategy-to Enable Emergency Bypass or Change of Protective Trips for Injection Pumps" which o
addresses pumps in other systems other than the RCIC system.
i.
Relationshio With Current'Reauirements and Practicei The generic emergency procedure guidelines of.the domestic BWR vendor and
~ the Emergency Operating Procedures (EOPs) of several plants were examined to determine the extent to which this strategy has been implemented. For the E0Ps examined, no procedural steps were found which used this strategy.
i Possible Adverse Effects Bypassing protective trips or changing the setting of trip values could be detrimental and would possibly result in the need for constant operator vigi-lance and dependence on the adequacy of existing instrumentation.
This is-expected to provide needed short-term pump availability, but ultimately could lead to extended loss in the longer term and risk substantial RCIC pump or turbine damage.
A 3.3 Core In.iection by Non-Safety Related Pumo 3.3.1 Strateav to Use Control Rod Drive Pumos for Core In.iection (BWR)
Strateav Descriotion_
R The aim of this strategy is to inject water into the reactor vessel to prevent or mitigate reactor core damage. This strategy is accomplished by the use of the control rod drive (CRD) pumps for core ihjection of unborated water when other methods are not available and during an anticipated transient without scram (ATWS), when control of reactor vessel water level is required to minimize reactor power, e.g., reactor water level / power control.
The CRD pump sub-system. consists of two AC operated pumps that while supplying motive force for control rod movement during normal operation also inject water into the vessel lower head via the CRD mechanisms to cool the i
mechanism drive piston seals. At least one of these pumps is operating at all times during normal operation and is aligned to draw water from the main con-denser hotwell reject line or the condensate storage tank..The flow of water into the vessel from the CRD system through the seals is regulated from the i
control room by throttling system control valves. After a. scram, the seal flow is increased substantially by the increased pressure directed to each mechanism i
by the opening of scram inlet valve. This is true unless there is a CRD system failure such as loss:of the pumps during a station blackout. Therefore, until the scram is reset, the CRD pump (s) provide a source of core injection-(above i
normal seal cooling flow) when other methods Reactor Core Isolation Cooling and High Pressure (e.g.,
Condensate /Feedwater, 1
Injection or Spray) are not available. This CRD flow can be maximized by using a pump. test bypass line 20 or opening up the CRD pressure contro1 ~ valves for a total injection flow of no more than several. hundred gallons per minute.
f In addition to using the CRD pumps as a backup to other methods of core injection there' are two special cases when their use might be considered, 4
19
o o
namely, for reactor water level / power control during an ATWS when boron is not
.available, and then again when core uncovery is suspected.
During an ATWS event in which boron is not available, reactor water level /
power control may be a viable means of preventing or mitigating core damage.
.In-this situation,:CRD pump (s) in combination with another injection pump may be useful as a source of controlled core injection to minimize reactor power.
In the event of an accident involving suspected core uncovery and possible core damage, the possible use of CRD pumps is covered in Subsection 6.2, "Strat-
.egy to Inject Borated Water in Case of Potential Core Damage and to Guard Against Boron Dilution in the Core."
Other related strategies include Subsections 4.7, " Strategy to Use Diesel-Driven Firewater Pump for BWR Core Injection, PWR Steam Generator Injection or Containment Sprays," and 6.1, " Strategy to Provide Additional Supply of Borated Makeup Water for Long-Te~rm Accident Control."
Relationshio With Current Reauirements and Practices The generic emergency procedure guidelines of the domestic BWR vendor rnd the_ Emergency Operating Procedures (EOPs) of several plants were examined to determine the extent to which this strategy has been implemented.
Procedural steps to use the CRD pumps to restore and maintain reactor vessel water level were found in the E0Ps of the plants examined.
In addition, under extreme low water level conditions without-an ATWS, several E0Ps reviewed include step (s) to maximize CRD flow.
Possible Adver[e' Effects
' Controlling water level correctly during an ATWS may prove to be difficult even when using the CRD pumps in combination with another injection pump. This level control concern may-be compounded by misleading or conflicting indica-tion (s) of vessel level and reactor power.
3.3.2 Strateav to Use Non-Safety Related Charaina Pumos for Core Iniection (PWR) i Strateay Descriotion i
The aim of this strategy is to supply water to the reactor vessel by using non-safety related charging puns when other sources of borated water are i
unavailable for high pressure em.gency core injection requirements. Note that a non-safety related charging pump as used in this strategy refers to any in-stalled high head charging pump whose electrical power does not come from an emergency bus and which has not been qualified to safety-related standards. To accomplish this strategy, electrical power to the motor-driver (s) of the pump (s) should be assured via the normal non-emergency bus or possibly by provisions for connecting to a more reliable alternate AC source.
During normal operations, a redundant charging pump supplies borated water to the reactor vessel at the proper high pressure flow and boron concentration for volume and chemical control. In most PWRs, the charging pump discharge flow
. splits before reaching the reactor vessel, with some flow going to each reactor L
20
G k
I the majority goes
. c'oolant pump (RCP) gland L.!) assembly (seal injection)
Most C ustion Engineering into the reactor coolant system (RCS) cold leg (s).
l t
all PWRs do not use seal injection to provide seal cooling; in these p an s, charging flow goes to the cold leg (s).
It is important to note that in many PWRs the design and implemen i l the high head charging pumps is such that they operate as part of the F
In an emergency, these safety related pumps and volume control system (CVCS).
Given a switch th' sir function to provide high pressure emerg ignal, these pumps will automatically start or continue to run (wit h
automatic restart upon restoration of power). charging pump i
In contrast, there emergency source, the refueling water :;torage tank (RW d
,i are only operated as normal CVCS components.(including p
-l if av=" ?'r, for emergFncy core injection.
This strategy addresses emergency situations where the reactor co t
system (RCS) remains st high pressure as-fer instance due to loss D,
in the steam generators (SG), anticipated transient with I
head of most high head safety injection pumps, only the charging pump small break LOCA.
provide injection to the RCS.
sate or Startup Feedwater Pumps for Steam Ge "Strat-PWR Steam egy to Use Diesel-Driven Firewater Pump for BWR Core Injection, j
Generator Injection or Containment Sprays."
Relationshio With Current Reouirem2nts and Practices The generic emergency procedure guidelines of the do Procedural j
the Emergency Operating Procedures determine the extent to which this strategy has been implemented.
ft f
steps directing the operator (s) to initiate core injection with a non l
t y
related charging pump were not identified in the E0Ps examined.
so that this strategy would not apply to them.
Possible Adverse Effects _
Other than generic concerns discussed in the Introduction of this there were no specific concerns identified for this strategy.
Iniection (e.o..
Hydrotest Pumo) When Strateov to Use Alternate Seal Reactor Coolant Pumo Seal Coolino is Lost (PWR) 3.4 Strateav Descrintiori The aim of this strategy is to regain reactor coolant pump (R To accomplish this strategy, cooling by an alternate means of seal injection.a suitable d
One 21
~
p f's V
O alternative might be an instailed hydrotest pump.
This strategy only applies to those PWRs which normally ust seal injection.
1 During normal operation, one of (at least) two redundant charging pr.,ps i
supply borated water to the reactor vessel at the proper high pressure fic., and boron concentration.
In those PWRs using seal injection, the charg%g pump discharge f'ow splits before reaching the reactor vessel with some ',ow going to each RCP shaft seal assembly (seal injection) while the majority goes into the RCS cold leg (s). The seal injection flow in turn splits in the seal assem-bly with some of the flow going down into the RCP pushing relatively cool water past the RCP thermal barrier, thus preventing RCS water from entering the seal assembly. The remainder of the seal injection flow passes through the pressure breakdown seal stages and exits to the volume control tank and other leakoff paths.
If seal injection flow is lost (or does not exist as part of the RCP seal assembly design like in most Combustion Engineering plants), the seals are cooled by low temperature RCS water flowing up through the seal assembly. This RCS water is cooled by'a' heat exchange.r in the RCP thermal barrier supplied by component cooling water (CCW).
Alternate seal injection cannot be used in plants which do not normally use seal injection.
Plants which have seal in-jection, would have to use an alternate %en the normal seal injection and thermal barrier seal cooling are not effectively cooling the RCP seals.
This alternate RCP seal injection strategy specifically addresses those situations where the safety related charging pumps and the CCW flow to.the RCP thermal barrier heat exchangers are not adequately cooling the RCP seals in PWR plants with RCP seal injection.
A related strategy is included in Subsection 3.3.2, " Strategy to Use Non-Safety Related Charging Pumps for Core Injection.",
Relationshio With Current Reouirements and Practices The generic emergency procedure guidelines of the domestic PWR vendors and the Emergency Operating Procedures (EOPs) of several plants utilizing seal injection were examined to determine the extent to which this strategy has already been implemented. Procedural steps directing the operator (s) to lineup and initiate alternate seal injection were not identified in the E0Ps of the I
plants examined. ThesedomesticplantE0Ps,directtheoperator(s)totripthe RCPs on loss of seal cooling. However, the reactor safety literature revealed i
at least two foreign plants (in different countries) which have made provisions to use a hydrotest pump for alternate seal injection.
Possible Adverse Effects Other than generic concerns discussed in the Introduction of this report, there were no specific concerns identified for this strategy.
3.5 Strateov to Use Condensate Pumos or Startuo Feedwater Pumos for Steam Generator Iniection (PWR)
Strateov Descriotion The aim of this strategy is to provide steam generator (SG) feedwater injection when normal station AC power is available, and main feedwater (MFW) 22
i.-
and auxiliary (emergency) feedwater (AFW) pumps are unavailable. To accomplish i
i
'this strategy, a suitable alternate for MFW and AFW should be used.
Suitable L
alternates might be the. main condensate pumps, the heater drain pumps or a startup feedwater pump.
For whichever alternate is used, the SG pressure must be reduced below the shutoff head of the pump and any interlock preventing the 1
use of these pumps in this strategy would have be bypassed or overridden. Mis strategy cannot be implemented in a station blackou', because the pumps reisire p
AC power for their operation.
[
The condensate pumps normally pump water from the main condenser hotwell j
to the suction side of the MFW pumps via booster pumps (or heater drain pumps, j
if appropriate).
Where startup feedwater pumps exist in PWRs, they are of various capacities.
Generally they take sttion from the condensate storage
[
-tank (CST).and inject into the SG.
During an accident where the MFW pumps and the AFW pumps are unavailable, l
the SG pressure can be'ieduced and the MFW isolation valves could be reopened to permit the condensate pumps to inject directly into the SGs bypassing the MFW pumps.
The flow of these pumps at low pressures may be sufficient to provide i
cooling to the reactor coolant system ifWertain accident situations.
Simi-larly, other pumps could be used. Specific plant analysis would be required to i
determine when these pumps could provide sufficient cooling to handle decay heat loads.
i Other related strategies include Subsections 2.5,
" Strategy to Ensure -
l
~ Adequate Plant Heat Removal Capability by Emergency Connection (s) of Existing or Alternate Water Sources," and 4.7, " Strategy to Use Diesel-Driven Firewater Pump. for.BWR Core Injection,- PWR. Steam Generato.r Injection = or Containment i
Sprays."
Relationshin With Current Reauirements and Practices j
}
The generic emergency procedure guidelines of the domestic PWR vendors and the Emergency Operating Procedures (EOPs) of several plants were examined to 2
l determine the extent to which this strategy has been implemented.
Procedural j
steps to use condensate pumps for' low pressure SG injection were found in the i
procedures of several plants examined.
i l
Possible Adverse Effects During restoration of a secondary heat sink, it may become necessary to inject relatively cool water into a hot dry SG even after core damage.
This j
injection from the main condenser hotwell or the' CST may result in excessive thermal stresses in the SG, possibly leading to SG tube failure. Reestablishing injection will result in SG repressurization and may lead to SG pressure greater than the shutoff head of one or more of the lower head pumps considered for this strategy (e.g., main condensate pumps, heater drain pumps).
1
/
23
J,. 2 L.
['
- 4.
STRATEGIES RELATED TO LOSS OF POWER 4.1 Strateay to Conserve Battery Canacity by Sheddina Non-Essential loads
[
L Strateay Descriotion The aim of this strategy is to conserve station battery power for essential loads as long as possible-in the event of a station blackout (S80).. The essen-tial loads are those related to maintaining control of_ the systems needed to bring the plant to a safe shutdown and maintain it. To accomplish this strat-egy, non-essential DC loads -should be shed.
During an emergency, the plant's DC power system provides a reliable supply of power to DC and vital AC bus loads required by the emergency equipment.
Its design.is plant specific and may be influenced more by the architect / engineering firm involved in the phnt' design and construction _than by the plant or contain-ment type.
For redundancy at least two divisions are used and newer plants typically have four divisions.
Each division has its own battery and one or more battery chargers.
Normally the DC=% ads are supplied by the installed battery chargers which keep the batteries up to full charge. During an SB0, the installed chargers stop charging and the DC and vital AC bus loads start to discharge the batteries.
Other related strategies include Subsections 4.2, " Strategy to Use Portable Battery Chargers or Other Power Sources to Recharge Station Batteries," and 4.5,
" Strategy to Enable Emergency Crosstie of AC Power Between Two Units or to an Onsite Gas Turbine Generator."
Relationshin With current Reauirements and Practices The generic emergency procedure guidelines of the domestic BWR and PWR vendors and the Emergency Operating Procedures (EOPs) of several plants were examined to determine the extent to which this strategy has been implemented.
All. plants were found to have some provisions for load shedding in their E0Ps.
The NRC's 1988 SBO' Final Rule required. that all plants be capable of withstanding a total loss of AC electrical power for a specified duration while maintaining both reactor core cooling and containment integrity. According to the rule, the capability for coping with an SB0 of specified duration may be determined by an appropriate analysis in lieu of providing an additional onsite emergency AC power source.
The analysis should include a description of the procedures and a list of equipment' modifications that will be implemented.
It is likely that most plants will implement procedures for load shedding to help respond to the rule.
Possible Adverse Effects Implementing this strategy allows for the possibility of inadvertently shedding the wrong loads and/or delayed shedding of correct loads, hence prema-turely depleting the station batteries during an SBO.
24
p p (.
1 4.2 Strateav to Use Portable Battery Charaers or Other Power Sources to Re-i charae Station Batteries (BWR and PWR) 5 Strateav Description The aim of this ' strategy is to _ recharge the station batteries during ~a 4
station blackout (SBO) thereby providing prolonged DC power supply for vital safety functions in the plant. To accomplish this s';rategy, a power source such i-as a suitably sized portable gasoline, engine drive.n battery charger might be used to recharge a station battery. The chargers would be placed into operation during an SB0 when the return of AC power does'not seem imminent and time to accomplish the portable hook up task is available. This strategy should reflect i
consideration of the need and priorities for power to vital ECCS related func-tions and a minimum set of plant sensors which adequately monitor' plant status.
The DC power system provides a reliable supply of power to DC and vital AC l
bus loads required by the emergency equipment. Its design is plant specific and may be influenced more by the architect / engineering firm involved in the plant design and construction than by the plant or containment type.
For redundancy i
at least two divisions are used and newer $ ants typically have four divisions.
i Each division has its own battery and one or more battery chargers.
Normally the DC loads are supplied by the installed battery chargers which keep the batteries up to full charge. During an SBO, the installed chargers cease charg-l ing and the DC and vital AC bus loads start to discharge the batteries.
Other related strategies include Subsections 4.1, " Strategy to Conserve y
Battery Capacity by Shedding Non-Essential-Loads," and 4.5, " Strategy to Enable Emergency Cross. tie of AC Power Between Two Units or-to an Onsite Gas Turbine Generator."
1-l Relationshio With Current Reouirements and Practices The generic emergency procedure guidelines of the' domestic BWR and PWR vendors along with the E0Ps of several plants were reviewed to determine'the extent to which this strategy has been implemented.
In one instance, the use j
'of-portable battery chargers was found.
i j
The NRC's 1988 SB0 Final Rule required that all plants be capable of withstanding a total loss of AC electrical power for a.specified duration while maintaining both reactor core cooling and containment integrity. According to l'
.the rule, the capability for coping with an SB0 of specified duration may be f
determined by an appropriate analysis in lieu of providing an additional onsite j
emergency AC power source.
The analysis should include a description of the procedures and a list of equipment modifications that will be implemented.
It 1
i is possible.that some plants may propose to add portable battery chargers to help respond to the rule.
i Possible Adverse Effects 1-Implementing this strategy involves the use of non-safety grade equipment 1
on a safety related system which could jeopardize the remaining battery capacity if the portable charger is' not properly isolated or properly used.
i' 25 1
d
..,p'*
)
P
' 4.3 Strateay to Enable Emeraency Reolenishment of the Pneumatic Sunolv for-
. Safety Related Air Ooerated Components (BWR and PWR) i Strateav Descriotion.
[
The aim of this strategy is to mitigate an accident 'by preventing the premature functional loss of critical equipment requiring ' instrument air (IA).
-This strategy is accomplished by replenishing the air supply with an appropri-ately filtered and dried alternate supply to ensure that safety related air-operated valves and instruments will. be able to operate as necessary during an i-extended severe accident. Options for additional air supplies include:~ service air (SA) systems (which is typically non-safety related), diesel air compressors F
(typically used as a backup to the SA system), and additional onsite storage of j:
bottled air systems.
1 i
The accumulators pf-safety-related air-operated valves at plants considered i
in this study provide air pressure for a certain number of valve cycles after loss of supply-air pressure. However,-during an extended accident, more valve l
actuation cycles may be needed to ensure 4hutdown of the plant and to provide
}
long term cooling. Valves not normally considered in design basis accidents may 1
be considered if they can help prevent or mitigate an extended accident.
Relationshio With Current Recuirements and Practices C
l The generic emergency procedure guidelines of the domestic BWR and &
vendors and the Emergency Operating Procedures (EOPs) of several plants w examined to determine the extent'to which this strategy has been implemental.
i-Modifications of systems or additional equipment currently in place at plants reviewed include:
a crossover connection between the SA and IA systems-to backup IA, a portable diesel compressor available for connection to an SA 1
connection, nitrogen gas bottle banks as needed to provide long term actuation L
of safety-related air operated valves, and backup such as a liquid nitrogen truck.
Some modifications to the air supply systems at many plants have oc-j-
curred in compliance with Generic Issue 43 (Instrument Air), Generic Issue A-44 1
(Station Blackout) and Generic Issue B-56 (Diesel Reliability).
In the plants reviewed, the modifications to provide backup bottled nitro-F i
gen for critical safety-related valves and crossover lines were found in most E0Ps. To a lesser extent, the remainder of the proposed modifications were also j
referenced in the E0Ps.
Possible Adverse Effects The use of SA in components normally supplied by IA without a filter / dryer k,
system may result in malfunction and even failure of these components.
Cross-connections of nitrogen systems in BWR plants with inerted contain-ments with the air systems may compromise containment inerting requirements.
/
1 26 1
[
7, V'
4.4 Strateay to Enable Emeroency Byoass or Chance of Protective Trios for Emeroency Diesel Generators (BWR and PWR)
Strateov Descriotion The aim of this strategy is to enable continued emergency diesel generator i
1 (EDG) operation beyond the point where they would normally trip, with the inten-tion of preventing and mitigating an accident before the EDGs fail. This strat-3 egy is accomplished by bypassing certain protective trips or changing their trip setpoints unless this selective bypassing or changing could result in early failure of the EDGs.
The EDGs in most plants have been designed with an automatic bypass of some protective trips during an emergency start.
Examples of the types of trips typically bypassed during emergency starts are: hi high vibration, low turbocharger lube oil pressure,gh jacket water temperature, main bearing high tempera-ture, and connecting rod bearing high temperature. Other trips which are found to be automaticall crankcase pressure,y bypassed in some plants are low lube oil pressure, high and generator-differential, i
If automatic bypass of any of the above trips is not presently part of the system dedgn in a particular plant, i.e., within its design basis, they may still be candidates for manual bypass..The current regulatory guidance allows for bypassing an EDG trip under accident conditions provided that the operator has sufficient time to react appropriately to an abnormal EDG unit condition.
An assessment of each trip considered for bypass should be performed. This assessment should include detailed information on the original design require-ments for each trip and analysis of potential accidents in which the trip might be bypassed.
If trips are bypassed in an accident condition, the need for continuous or frequent monitoring of parameter readings should be assessed.
Other related strategies include Subsections 3.2.1, " Strategy to Enable Emergency Bypass or Change of Protective Trips for Injection Pumps," 3.2.2,
" Strategy to Extend Reactor Core Isolation Cooling System Availability by Pump Trip Function Bypass or Change," and 4.5, " Strategy to Enable Emergency Crosstie of AC Power Between Two Units or to an Onsite Gas Turbine Generator."
Relationshio With Current Reouirements and Practices The generic emergency procedure guidelines of the domestic BWR and PWR vendors and the Emergency Operating Procedures (EOPs) of several plants were examined to determine the extent to which this strategy has been implemented.
For the plants reviewed, no procedural steps utilizing this strategy were found.
The NRC's 1988 SB0 Final Rule required that all plants be capable of withstanding a total loss of AC electrical power for a specified duration while maintaining both reactor core cooling and containment integrity. According to the rule, the capability for coping with an SB0 of specified duration may be determined by an appropriate analysis in lieu of providing an additional onsite emergency AC power source.
The analysis should include a description of the procedures and a list of equipment modifications that will be implemented. This strategy could be part of such a response to the rule.
m
~.-
f.
4 j
Possible Adverse Effects Bypassing EDG protective trips or changing their trip setpoints could be i
detrimental and would possibly result in the need for constant operator vigi-l-
lance and dependence on the adequacy of existing instrumentation.
This action is expected.to provide needed short-term EDG availability, but ultimately could lead to extended loss in the longer term and risk substantial EDG damage.
4.5 Strateay to Enable Emeroency Crosstie of AC Power Between Two Units or to an Onsite Gas Turbine Generator (BWR and PWR) l Strateav Description i
i The aim of this strategy is to provide an alternate source of AC electrical power to the unit's emergency buses to help recover from a station blackout i
I (SBO) when the unit's normal and emergency AC power sources are lost.
This permits continued operation of safety-related equipment.
This strategy is accomplished by establishing an emergency crosstie capability -(AC switchyards j
and/or diesel generators) between equivalent AC power systems of two units at j
a multi-unit site, or by connecting an afe41able onsite gas turbine generator j
to the AC power system to provide an alternate AC power source. Plant electri-cal systems are usually compatible and only require minor desigr and planning i
to accomplish crosstying of electrical equipment at multi-unit facilities, 4
through the use of switchgear and controls, but this would need to be addressed
~
as part of strategy assessment.
Implementation:of this strategy may not be possible at single unit loca-tions unless another source of independent offsite AC power exists, e.g., a gas i
turbine' generator. Several-plants examined were found to have large gas turbine j
generators onsite.
i For this strategy, a gas turbine generator or other AC power' source, must be capable of developing plant emergency bus voltage.
If a gas turbine gener-l ator is considered, black start capability is desirable.
Another related strategy includes Subsection 4.6, " Strategy to Use a Diesel Generator or Gas Turbine Generator to Power a Control Rod Drive or Other Ap-l propriate Pump for Core Injection."
Relationshio With Current Reauirements and Practices j
The generic emergency procedure guidelines of the domestic BWR and PWR vendors and the Emergency Operating Procedures (EOPs) of several plants were i
examined to determine the extent to which this strategy has been implemented.
j' Procedural. steps to implement this strategy were not found.
i The NRC's 1988 SB0 Final Rule required that all plants be capable of withstanding a total loss of AC electrical power for a specified duration while maintaining both reactor core cooling and containment integrity. According to the rule, the capability for coping with an SB0 of specified duration may be i
determined by an appropriate analysis in lieu of providing an additional onsite f
i emergency AC power source'.
The analysis should include a description of the procedures and list of equipment modifications that will be implemented. This j
strategy could be part of such a response to the rule.
k j
28 i
~
a j.f' I
Possible Adverse Effects Electrical system crossties at multi-unit sites may compromise the emer-
.gency AC power reliability of the unit sharing its power, e.g.. fault propaga-l tion.
4.6 Strateav to Use a Diesel Generator or Gas Turbine Generator to Power a j
Control Rod Drive or Other Anoropriate Pumo for Core-Iniection (BWR)-
2' l'
Strateay Description j'
. The aim. of this strategy is to supply alternate electrical AC power to i
drive.a control _ rod drive (CRD) or another appropriate pump for. core injection l
L and/or emergency boration. This strategy is accomplished by supplying emergency
[
power from a mobile diesel generator or a gas turbine generator to provide the appropriate AC power source to drive the pump (s). Other appropriate pumps might L
be residual heat removal (RHR) pumps and condensate / motor-driven feedwater pumps i
assuming the generator has sufficient capacity.
4 I
The use of this generator could prteent or mitigate a station blackout (SBO) accident. An alternate AC generating. unit used for driving the CRD pumps i
may be beneficial in other selected accident scenarios as well. Several plants 1
j examined were found to have large gas turbine generators onsite.
j l-For this strategy, a gas turbine generator or other AC power source, must be capable of developing the required. pump bus voltage and adequate capacity.
j-If a gas turbine generator is considered, black start capability is desirable.
i i
Other related strategies include Subsection's 3.3.1,
" Strategy to Use l
Control Rod Drive Pumps for Core Injection," 4.5, " Strategy to Enable Emergency
)
Crosstie of AC power Between Two Units or to an Onsite Gas Turbine Generator,"-
l and '6.1, " Strategy to Provide Additional Supply of Borated Makeup Water for j
Long-Term Accident Control."
d i
j Relationshio with Current Reauirements and Practices I
The generic emergency procedure guidelines of the domestic BWR vendor and j-the Emergency Operating Procedures (EOPs) of several plants were examined to determine the extent to which this strategy has been implemented.
Procedural-steps to perform this strategy have not been found.
3 The NRC's 1988 SB0 Final. Rule required that all plants be capable of 3-j withstanding a total loss of AC electrical power for a specified duration while maintaining both reactor core cooling and containment integrity. According to a
the rule,' the capability for coping with an SB0 of specified duration may be determined by an appropriate analysis in lieu of providing an additional onsite i
emergency AC. power source.
The analysis should include a' description of'the j
procedures and a list of equipment modifications that will be implemented. This j
strategy could be part of such a response to the rule.
EI Possible Adverse Effects i
Providing a mobile-AC generator, connecting it to the appropriate bus and I
operating the injection pump may increase the need for operator vig:hnce.
i 29
(
l 2
.~
~.
\\.,..;
O Or O
4.7 Strateay to Use Diesel-Driven Firewater Pumo for BWR Core Iniection. PWR Steam Generator In.iection or Containment Sorays (BWR and PWR)
Strateav Descriotion The aim of this strategy is to provide an alternate source of BWR core injection, PWR steam generator (SG) injection, or containment spray (CS) in both BWRs and PWRs. To accomplish this strategy, a diesel-driven firewater pump will l
be used as a source of the water.
Actual hard-piped crossties from the fire-water system to provide these functions do not exist in many U.S. plants. The alternative would be a temporary hose connection arrangement with the necessary connectors (e.g., spool piece).
Although such an arrangement would depend on specific plant configuration, it is likely that most plants have a penetration or blank flange that could be adapted to a hose connection from the firewater system.
Injection to a BWR core or to a PWR SG is more important than spraying the drywell of a BWR or the containment of a PWR.
The firewater supply system typically consists of one or more electrically
~
driven pump (s) and a backup pump driven by a dedicated diesel engine.
These pumps feed a firewater main which is tapped at various locations around the plant site. For plants located near fresh water rivers or lakes, suction to the fire pumps is usually taken directly from these sources and therefore has an unlimited supply. For other plants, the firewater pumps are supplied by one or more storage tanks with capacities of several hundred thousand gallons each.
This strategy addresses accident sequences involving a loss of all feed-water (both main and auxiliary) or a loss of CS.
Since the diesel-driven fire pump is independent of station AC power, this st,rategy may also be used in station blackout scenarios.
Also the use of the diesel firewater system to supply the spray headers could possibly prevent or delay containment overpres-sure failure during accidents such as LOCAs involving a loss of containment heat removal.
A related strategy is Subsection 3.5, " Strategy to Use Condensate Pumps or Startup Feedwater Pumps for Steam Generator Injection."
Relationshio With Current Reauirements and Practices The generic emergency procedure guidelines of the domestic BWR and PWR vendors and the Emergency Operating Procedures (EOPs) of several plants were ex.
amined to determine the extent to which this strategy has been implemented.
In some plants the E0Ps call for the use of the diesel-driven fire pumps as an alternate source of injection into either a BWR core or the PWR SGs. However, its use for either BWR or PWR containment sprays was not found.
The NRC's 1988 SB0 Final Rule required that all plants be capable of withstanding a total loss of AC electrical power for a specified duration while maintaining both reactor core cooling and containment integrity. According to the rule, the capability for coping with an SB0 of specified duration may be determined by an appropriate analysis in lieu of providing an additional onsite i
emergency AC power source.
The analysis should include a description of the procedures and a list of equipment modifications that will be implemented. This strategy could be part of such a response to the rule.
1 30
Possible Adverse Effects The use of the diesel-driven firewater pumps for BWR core injection, PWR SG injection or containment spray is a reduction in the flow available for the actual fire suppression systems in the unlikely event these are needed in the same time frame.
The use of non-filtered firewater for CS may result in clogged nozzles in the spray headers.
If appreciable flow can be achieved through the CS headers in a PWR, the addition of this unborated water into the containment sump may i
i pose possible reactivity problems due to boron dilution when the sump is used for core cooling during emergency sump recirculation.
The use of CS in the later stages of an extended accident when the contain-ment atmosphere may contain significant amounts of steam, air and hydrogen, would condense steam and could result in more readily combustible mixtures.
e 18ts l
I
/
31
6.
.o l.c.
j 1
i 5.
STRATEGIES RELATED TO LOSS OF HEAT SINK-i 5.1 Strateav to Recoen Main ~ Steam Isolation Valves and Turbine Bvoass Valves-
)
[
to Reaain the Main Condenser as a Heat Sink (BWR and PWR)
.{
Strateav Descriotion The aim of this strategy-is to regain the main condenser as a heat sink by 1
i reopening the main steam isolation valves (MSIVs) (or -their drain bypass l
headers) and the turbine bypass valves To-
' accomplish this strategy, condenser vacuum (TBVs) after they have closed.
must be maintained or reestablished and circulating water must be available.- Then main steam line (MSL) pressure
-on both sides of the MSIVs may be equalized while the MSLs are being drained and warmed.
If the MSIVs are to be opened, the isolation signal input (s) which j
closed the MSIVs must be cleared and reset, or the isolation interlocks bypassed or defeated. This strategy addresses those valve isolation situations where the i
main condenser is available with its vacuum maintained or able to be. rees-i tablished easily. Therefore, circulating water, turbine gland sealing steam and
}
the vacuum pumps must be available.
Also, the circumstances which caused the i
isolation must be corrected or toleratedM the isolation is to be overridden.
r l
If overridden, the isolation function will probably be defeated thus eliminating p
any further automatic reisolation.
a 1
Note that this strategy does not include reopening BWR HSIVs based on closure caused by a MSL break or fuel damage associated with high radiation.
-Likewise, it does not include reopening PWR MSIVs based on closure caused by a MSL break located downstream of the MSIVs. Those PWR MSIV closures associated with isolating a steam generator tube rupture or MSL break from that steam generator (SG) may also be excluded.
In almost all BWRs, there are four MSLs each containing two redundant
~
MSIVs,. both near the primary containment, one inside and one outside.
Each valve is designed to rapidly close on abnormal conditions. The closure of one MSIV in each MSL will prevent the release of extraordinary amounts of radioac-tive materials to the turbine building and/or the plant stack in the event of abnormal fuel failure.
Further, the MSIV closure will limit reactor vessel inventory loss in the event of~a MSL break outside primary containment.
The closure is initiated by containment isolation logic to all valves simultane-ously, or manually from the main control room.
In almost all PWRs, there is one MSIV (or equivalent set of check valves) on each of the MSLs near the containment on the outside.
These MSIVs are designed also to rapidly close on abnormal conditions. The closure of all MSIVs will prevent the rapid cooldown of the reactor coolant system (RCS) in the event of a MSL break outside of containment.
Also, the closure of all valves will reduce the containment pressure buildup due to a MSL break inside containment by preventing backflow through the intact MSLs. Further, it will limit loss of reactor coolant inventory and possible radioactive release in the event of a steam generator (SG) tube rupture.
The closure is initiated automatically or manually from the main control room by isolation logic to all valves simul-taneously.
On both BWRs and PWRs, drains are located on either side of the MSIVs.
These' sets of drains permit drainage to a drain system or the main condenser j
l 32
)
4
+
.."I o i
i hotwell. The drain system is utilized to drain water out of, and warm up, the MSLs.
It.also equalizes pressure across the MSIVs.
This may have to be done along with resetting or defeating the MSIV isolation signal prior to attempting the reopening'of an isolated MSIV.
The TBVs on most BWRs and PWRs are designed and used to provide the normal means of controlled cooldown of the plant via MSL pressure control prior to the use of residual heat removal RHR. The TBV isolation logic is normally based i
on the loss of condenser vacuu(m. )This logic should not be defeated as i
this strategy since the integrity of the main condenser and its hotwell could be jeopardized if it were pressurized.
Using the main condenser as a heat sink avoids dumping steam to the sup-pression pool in a BWR or to the atmosphere by the steam generator atmospheric dump valves 'in a PWR, especially during an ATWS.
If the reactor is shutdown, the use of the MSL drain headers without reopening MSIVs in some plants may be adequate.
Relationshio With Current Reauirements and Practices
)
The generic emergency procedure guidelines of the BWR and PWR domestic vendors and the Emergency Operating Procedures (EOPs) of many plants were examined to determine the extent to which this_ strategy has been implemented, For several BWRs examined, there were detailed procedural steps for reopening the MSIVs and TBVs to regain the main condenser as a heat sink, if available, provided there is no indication of a MSL break or gross fuel failure. Among the PWRs, several have procedural steps to reopen the MSIVs and TBVs.
Possible Adverse Effects Defeating the MSIV isolation function logic and reopening the MSIVs poten-tially can have a negative influence on the state of the plant, e.g., failure of the condenser due to overpressure is a possibility.
There is also the possibility of the operator (s) error when the action requires many relatively complicated and unfamiliar steps to be accomplished.
Even if the defeated logic is performed correctly, the automatic isolation capability is most probably lost thus significantly increasing the need for operator vigilance and possible manual isolation.
i
/
I l
Q
.S
.?l o 6.
STRATEGIES RELATED TO SHUTDOWN FAILURE (REACTIVITY CONTROL) 6.1 Strateov to Provide Additional Supply of Borated Makeup Water for lono-Term Accident Control (BWR and PWR)
Strateov Descriotion The aim of this strategy is to supply adequate borated makeup water for long-term accident control. Consideration must be given to the potential needs for borated water that may result from the wide range of plant specific acci-dents.
To accomplish this strategy, a sufficient supply of boron must be accessible on site or readily available.
The amount of borated water and its concentration needs to be identified and steps necessary to prepare it for supply to the reactor vessel specified.
In PWRs, borated water is used in the reactor coolant system (RCS) and i
maintained at proper concentration for long-term reactivity control. A loss of j
coolant accident (LOCA) could place a significant demand for borated makeup
~
water that may exceed the refueling water storage tank (RWST) capacity and if containment sump recirculation fails or tv-unavailable, additional sources of 1
borated water would be required.
This source may be a large tank of concen-l trated boric acid, such as the boric acid storage tank, combined with a large source of water, such as the demineralized water storage tank, at the suction of the charging pumps.
In BWRs, boron is only used when control rods.are not available for reac-tivity control, as in the case of an anticipated transient withou't scram (ATWS).
Reactor water level / power control is used during an ATWS while arranging for boron injection to shutdown the reactor.
Boron injection is normally ac-complished by the standby liquid control system (SLCS).
Other than the SLCS tank capacity, most BWRs only have a limited supply of borated water available and that quantity only exists in an unprepared form.
j l
Other related strategies include Subsections 2.3.2, " Strategy to Refill Refueling Water Storage Tank With Borated Water," and 6.2, " Strategy to Inject Borated Water in Case of Potential Core Damage and to Guard Against Boron Dilution in the Core."
Relationshio With Current Reouirements and Practices The generic emergency procedure guidelines of the domestic BWR and PWR 4
vendors and the Emergency Operating Procedures (EOPs) of several plants were examined to determine the extent to which this strategy has been implemented.
In the E0Ps reviewed for PWRs, there exist methods and practices for mixing boron and water and then injecting the mixture into the RWST at a limited rate.
For BWRs, the E0Ps examined provided for additional supplies of boron and alternate methods of injection in response to the ATWS rule.
They may not be sufficient for accident conditions where core damage is anticipated.
Possible Adverse Effects Other than generic concerns discussed in the Introduction of this report, there were no specific concerns identified for this strategy.
m
l<
G
.L*
e, r
6.2 Strateov to In.iect Borated Water in Case of Potential Core Damaae and to Guard Aaainst Boron Dilution in the Core (BWR)
Strateov Description The aim of this strategy is to ensure that proper concentrations of boron l
can be injected and maintained in the reactor core when core uncovery and 1
possible damage are suspected. This strategy is accomplished by the appropriate use of the standby liquid control system (SLCS) or an alternate injection method such as the control rod drive (CRD) system, the reactor water cleanup (RWCU) system, etc., using a limited capacity source of borated water (see Subsection 6.1, " Strategy to Provide Additional Supply of Borated Makeup Water for Long-Term Accident Control").
Borated water supply for an alternate injection method i
may be accomplished by temporary connections. Most plants require AC power for boron injection.
A source of independent power for boron injection may be desirable in the case of a station blackout that results in core damage.
The strategy for such a power source is covered in Subsection 4.6, " Strategy to Use a Diesel Generator or Gas Turbine Generator to Power a Control Rod Drive or Other Appropriate Pump for Core Injection."
A l
In the event of an accident involving core uncovery, the control rods are predicted to begin to melt prior to the fuel rods, thus core reflood could result in possible recriticality.
To avoid or reduce the chances of recriti-cality and possible associated rapid power generation, borated water could be injected using the systems mentioned above.
Once the boron has been injected into the vessel, adequate boror, concentra-tions need to be maintained in the core. The reactor vessel refill rate should i
be controlled so that boron is not lost from the core. As long as water ese:p::
the vessel in the form of steam above the water level, significat emuunts of boron should not escape; boron dilution, therefore, should not be an important-If, however, it escapes below the water level boron may be lost from concern.
the core and dilution may provide a recriticality concern.
Until a source of borated water is again available and injected to shut down the reactor, the i
potentially damaged core will have to be cooled by unborated water injection at controlled rates. A balance of reactor water level / power control will have to be achieved, which will have to be maintained until some method can be found to inject boron.
This might be accomplished by CRD pump (s) possibly used in conjunction with another injection pump, the combination of which would be sufficient to control and maintain an appropriate water level.
The use of CRD pump (s) is covered in Subsection 3.3.1, " Strategy to Use Control Rod Drive Pumps for Core Injection."
If power level can be controlled at a low level, by maintaining liquid at appropriate levels in the core, damage may be stopped and core cooling may be achieved, although it is not assured.
The unborated water injection part of this potential core damage related strategy is assessed as being beyond the design basis of a plant and requires further investigation including consideration of planning and training if such a situation is anticipated, f
Other related strategies are mentioned above.
35
a
' 'g h
h) 1.: $
- G
.o Relationshio With Current Reouirements and Practices The generic emergency procedure guidelines of the domestic BWR vendor and the Emergency Operating Procedures (EOPs) of several plants were examined to determine the extent to which this strategy has been implemented.
Procedural steps to inject borated water from SLCS and alternate injection paths were found in several plant E0Ps.
The initiation of borated water found in the E0Ps was in response to an anticipated transient without scram (ATWS).
No procedures were found that address boration in the event of core damage.
Possible Adverse Effects During an accident involving core damage with a very low vessel water level, the injection of unborated water at a high flow rate may significantly increase reactor power with detrimental effects (including the possibility of
'1 increased fuel rod failure).
Controlling water level and flow appropriately after the onset of core damage may be difficult even when using the CRD pumps.
This is especially true when unborated water is injected and boron dilution is a reactivity concern.
i i
l
/
36
_g
- k' g.
e i.
9 7.
REFERENCES l.
" Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants,"
(Second Draft for Peer Review), NUREG-Il50, June 1989.
- 2. =" Assessment of Severe Accident Prevention and Mitigation Feat'ures," Brook-I haven National Laboratory, NUREG/CR-4920, Volumes 1-5, July 1988.
i 3.
"Severo Accident Insights Report'," Brookhaven National Laboratory, NUREG/
CR-5132, April 1988.
I i-4.
" Emergency Procedure Guidelines," General Electric Owners Group, OEI Docu-
- ~
ment 8390-4, Revision 4.
I 5.
" Emergency Response Guidelines (both High and Low Pressure Versions),"
Westinghouse Owners Group, Revision 1, September 1983.
i 6.
" Emergency Procedure Guidelines," Combustion Engineering Owners Group, GEN-1 152, Revision 1.
A 7.
" Abnormal Transient Operating Guidelines," Babcock & Wilcox, No. 74-1123297-00, March 1982.
8.
"The Risk Management Implications of NUREG-1150 Methods and Results,"
Sandia National Laboratories, NUREG/CR-5263, September 1989.
9.
" Safety Evaluation Report Related to the Final Design Approval of the GESSAR !! BWR/6 Nuclear Island Design," Gener#1 Electric Company, NUREG-0979, Supplement No. 4.
- 10. - "Tne Effect of Small-Capacity, High-Pressure Injection _ Systems on TQUV Sequences at Browns Ferry Unit One," Oak Ridge National Laboratory, NUREG/CR-3179, September 1983.
/
37 v-
-&,,r
-w
-yc.,
-g,~.t, r-4
,e-,, - - - -
f%
^x ENetosvag 3, b
UNITED STATES
[. sw ' j NUCLEAR REGULATORY COMMISSION 7
WASHINGTON, D. C. 20555
%, v f TO:
ALL HOLDERS OF OPERATING LICENSES AND CONSTkUCTION PERMITS FOR NUCLEAR POWER REACTOR FACILITIES SUE >J ECT: ACCIDENT MANAGEMENT STRATEGIES FOR CONSIDERATION IN THE INDIVIDUAL DLANT EXAMINATION PROCESS - GENERIC LETTER 88-10, SUPPLEhENT N0. 2 1.
SUMMARY
Over the past several years, the NRC has performed and reviewed numerous probabilistic risk assessments (PRAs) and severe accident stuuies. From this experience, it has become evident that it is possible to implement certain actions, or accident management strategies, that have significant potential for recovering from a wide variety of accident scenarios. These accident management strategies typically involve the use of equipment that already exists at plants. The NRC staff has compiled a list of such accident manage-ment strategies. The purpose of this letter is to forward these strategies to
.inaustry so that licensees can evaluate these or similar strategies for applicability and effectiveness at each of their plants as part of conducting the Individual Plant Examination (IPE) called for in Generic Letter 88-20:
" Individual Plant Examination for Severe Accident Vulnerabilities." This generic letter supplement also transmits for information the enclosed NUREG/CR-5474, which contains a technical assessment of these accident management strategies.
This generic letter supp k.sent does not establish any requirements for licensees to take the specific accident management strategies into account as part of the IPE or to implement any of the strategies. Adoption on the part of a licensee of any accident management strategies in response to this supplement is voluntary.
2.
DISCUSSION Generic Letter 88-20 directs each utility to perform an IPE to identify any plant-specific vulnerabilities to severe accidents and report the results to hRC. One purpose of the IPE is to determine whether modifications to hardware and procedures are necessary to reduce the frequency of severe accidents or to i
mitigate their consequences. An effective way of achieving that goal may be through the implementation of accident management procedures, that is, procedures that promote the most effective use of available plant equipment and staff in the event of an accident.
In parallel with the IPE program, NRC is preparing to issue a generic letter in 1991 forwarding guidance to each licensee on the development of an accident management framework. The guidance will adoress identification and implemen-tation of accident management procedures and associated hardwara, development Technical
Contact:
R. Palla, NRR (301) 492-1076
T j
l
{
2 of training programs, definition of decisionmaking responsibilities, and development of technical guidance for operational staff such as technical support center personnel. The staff will work with the Nuclear Management and 4
Resources Council (NUMARC) to define the scope and content of a utility accident management framework or plan, and the means of implementing such a framework.
l It is expected that the accident management framework each licensee develops will ensure that procedures are implemented in an effective and integrated j
manner, and that due consideration will be given to potential negative impacts on plant safety.
Generic Letter 88-20 states that in the course of the IPE, utilities "may identify operator or other plant personnel actions that can substantially l
reduce the risk from severe accidents and that should be issnediately implemented in the form of emergency operating procedures or similar formal guidance," and encourages each licensee to "not defer implementing such actions until a more I
structured and comprehensive accident management program is developed on a i
longer schedule, but rather to implement such actions issnediately within the constraints of 10 CFR 50.59."
The staff guidance document concerning the IPE submittal (NUREG-1335, " Individual Plant Examination: SubmittalGuidance")
requests that licensees document any strategies that were developed as part of i
the IPE process to prevent or mitigate the detrimental effects of severe accidents.
As a result of experience with numerous PRAs and severe accident studies, the NRC staff has identified several accident management strategies that have significant potential for reducing plant risk. These accident management j
strategies can be grouped into three categories:
i Conserving and/or replenishing limited resources during the course i
of an accident. These resources would include, for example, battery capacity, borated water, and compressed air.
Using plant systems and components for innovative applications during an accident. This usage would include enabling crossties of support systems l
or the use of fire systems, or contrcl rod drive (CRD) pumps in the case of a boiling water reactor (BWR), for decay heat removal.
In addition, this category includes procedures to connect alternate electrical power i
j sources to meet critical safety needs during an accident.
Defeating appropriate interlocks and overriding component protective trips in emergency situations. An example of this strate the ability to reopen main steam isolation valves (MSIVs)gy would be i
in a BWR anticipatedtransientwithoutscram(ATWS) event.
These three categories, and others as appropriate, can be applied to eich of the major safety functions of the plant such as reactivity control, coolant inventory control, heat removal, and containment performance, as well as to the principal support functions such as electric power, equipment cooling, and air systems. Table 1 contains a list of examples of strategies derived from PRAs categorized into one of the three categories above. The NRC believes that a significant risk reduction benefit can be achieved, with reasonable resource expenditure, by implementation of emergency procedures and/or operating guidance associated with selected accident management strategies.
~
3 i
We encourage evaluation of accident management strategies in conjunction with
?
the IPE that each utility is expected to perform pursuant to Generic i
Letter 88-20. The IPE process is expected to disclose plant-specific design l
.and operational information that will guide the evaluation of candidate i
strategies and the implementation of corresponding procedures or guidance, if appropriate.
It is not intended that the evaluation of potential accident management measures be limited to the specific strategies identified in Table 1.
For instance, strategies to maintain containment function and to delay or prevent possible early containment failure can also be assessed in this context.
Accident management strategies related to external events would be considered at a later date.
As part of the strategy evaluation process, we encourage licensees to consider the potential drawbacks or negative aspects of each strategy as well as the risk reduction potential. A detailed technical assessment of several accident management strategies is provided in the enclosed NUREG/CR-5474. This document provides evaluation guidance and cautions for each strategy. Licensees may wish to review the information provided in this document as part of the strategy vivaluation and implementation process to provide added assurance that use of the strategy will not detract from overall plant safety.
In the course of evaluating potential operator actions and accident management strategies, licensees may identify certain aspects of existing regulations or regulatory guidance that preclude, conflict with, or otherwise hinder the implementation of etfective accident management measures, and that may not be in the interest of overall safety. Licensees are encouraged to inform the NRC of such situations so that NRC can consider the need for further clarification or modification of these regulations or guidance.
3.
LICENSEE RESPONSE Licensees are encouraged to consider accident management strategies, such as those identified in Table 1 and asessed in NUREG/CR-5474, for applicability and effectiveness as part of conducting the Individual Plant Examination called for in Generic Letter 88-20.
In accordance with guidance provided in NUREG-1335, licensees "should document any strategies to further prevent or mitigate the detrimental effects of severe accidents that were developed as part of the IPE
~
process and for which credit has been taken in the analysis." : For those plants with an existing PRA or IPE study, licensees may wish to consider the accident management strategies described herein, and document any procedures adopted as a result, as a follow-on to submittal of the IPE documentation. No documen-tation beyond that requested by Generic Letter 88-20 is necessary unless the IPE analysis is modified as a result of this evaluation.
This generic letter supplement does not establish any reporting requirements.
If reports to the NRC result from actions suggested herein, they are covered
_~_ -
)
us y
4
-s i
i-by Office of Managenent and Budget Clearance No. 3150-0011, which expires on January 31, 1991.. The estimate of burden on licensees is covered by and unchanged from that presented in Generic Letter 88-20.
Sincerely, 1
Jares G. Partlow L
Associate Director for Projects
]
Office of Nuclear Reactor Regulation
Enclosures:
1.
Table 1, " Generic Accident Management Strategies" l
2.
NUREG/CR-5474, " Assessment of Candidate Accident Management Strategies" l
3.
List of Recently Issued NRC Generic Letters i
I a
i 4
1 1
J i
i i
i j
Table 1 Generic Accident Management Strategies 1.
Conserving and Replenishing Limited Resources Refill refueling water storage tank (RWST) with borated water, or condensate storage tank (CST) with condensate. Assure adequate supply of boron on site.
Maintain emergency core cooling system (ECCS) suction to condensate systems to avoid pump failure as a result of high suppression pool temperature.
Throttle containment sprays to conserve water for core injection.
Conserve battery capacity by shedding non-essential loads.
Use portable battery chargers or other power sources to recharge batteries.
Enable emergency replenishment of gas supply, or otherwise ensure operability of air-operated components.
Enable early detection, isolation, or otherwise mitigate the effects of-an interfacing systems loss-of-coolant accident (LOCA).
- 11. Using Systems and Components in Innovative Applications Strategies to enable emergency use of available pumps to accomplish safety functions.
Use diesel-driven fire systems for injection to the containment sprays, a BWR core, or the PWR steam generators.
Use control rod drive (CRD) pumps in BWRs or charging pumps in PWRs for core injection.
Use alternate injection (e.g., hydro test pump) when reactor coolant pump seal cooling is lost.
Enable emergency crosstie of service water and component cooling water to residual heat removal in BWRs or feedwater in PWRs.
j Use condensate or startup pumps for feedwater injection.
/\\
p t
+\\
~
U V
s Table 1 (Cont'd)
Strategies to enable emergency connection of available ac power sources to meet critical safet:: needs.
Use diesel generator or gas turbine generator to drive CRD pumps for core inject an.
Enable emergency crosstie of ac power between two units or to onsite gas turbine generator.
Strategies to enable emergency connection of injection systems to alternate water sources.
Ensure appropriate recirculation switchover and cope with the failure to switch over in LOCAs.
Enable emergency connection of service water or feedwater systems to rivers, reservoirs, or municipal water systems.
Strategies for Reactivity Control.
Initiate standby liquid control system (SLCS) in case of potential core damage and guard against baron dilution when core injection is restored.
Ensure abundant supply of borated makeup for long-term accident control.
Ill. Defeating Interlocks and Component Protective Trips in Emergencies Reopen main steam isolation valves (MSIVs) and turbine bypass valves to regain the condenser as a heat sink.
Extend reactor core isolation cooling (RCIC) availability by either raising the turbine exhaust pressure trip setpoint or overriding the trip function.
Enable emergency bypass of protective trips for diesel generators and injection pumps.
[(
ENCLOSURE 3
~
gaarcy l-
?$
POLICY ISSUE January 18, 1989 o a ion Vot@
SECY-89-012 For:
The Commissioners From:
Victor Stello, Jr.
Executive Director for Operations
Subject:
STAFF PLANS FOR ACCIDENT MANAGEMENT REGULATORY AhD RESEARCH PROGRAMS
Purpose:
The purposes of this paper are:
1.
To describe the major goals, framework, ano elements of NRC's accident management program, and the approach of implementing the Accident Management portion of the integratio!. plan for severe accident closure (SECY-88-147), ano 2.
To summarize the research efforts planned or now underway to furnish confirmatory inforraation and technical support to the staff in support of Accident Management, 3.
To obtain approval of the Accident Management research program through FY90.
=
Background:
Accident Management encompasses those actions taken during the course of an accident by the plant operating and technical staff to: (1) prevent core damage, (2) terminate the progress of core damage if it begins and retain the core within the reactor vessel, (3) maintain containiaent integrity as long as possible, and (4) minimize offsite releases.
Accident management, in effect, extenos the defense-in-depth principle to plant operating staff by extending the operating procedures well beyond the plant design basis into severe fuel damage regimes, with the goal of taking advantage of existing plant equipment and operator skills and creativity to find ways to terminate accidents beyond the design basis or to limit offsite releases.
CONTACT:
L. Shotkin, RES, 49-23530 R. Earrett, hER, 49-21089 (Z?C
[
n L)
\\
~
~-
The NRC staff has concluded, based upon PRAs and severe
{
accident analyses, that the risk associated with severe core damage accidents can be further reduced through effective accident management.
In this context, effective accident management would ensure that optimal and maximum safety benefits are derived from available, existing systems and plant operating staff through pre-planned strategies.
Furthermore, the International Nuclear Safety
{
1 Advisory Group (INSAG) in its report on Basic Safety Principles for Nuclear Power Plants concluded.that accident management and mitigation measures can i
significantly reduce risk. Accordingly, accident management is considered to be an essential element of the j
severe accident closure process described in the Integration Plan for Closure of Severe Accident Issues (SECY-88-147) and the Generic Letter on the Individual
{
Plant Examination (Generic Letter 88-20).
In the IPE Generic Letter, the staff deferred the requirement to develop an accident management plan, stating that we are currently developing more specific guidance on this matter and are working with huMARC to (1) define the scope and content of acceptable accident j
management programs, and (2) identify a plan of action that will ultimately result in incorporating any 4
plant-specific actions deemed necessary, as a result of the IPE, into an overall severe accident management program.
Since that time we have made considerable progress towards development of an Accident Management Program that would lead to enhanced accident management capabilities in the nuclear industry. This program will be supported by an Accident Management Research Program, which will be coordinated with and use the results of other programs, including the Containment Performance Improvement Program (CPI), the Severe Accident Research Program (SARP), and the Human Factors Research Program. A summary description of the Accident Management Program and supporting research programs is presented below, and in Attachment 1.
Discussion:
The fundamental objective of the proposed Accident fianagement Program is the following:
4 Each NRC licensee shall implement for each nuclear plant an " Accident Management Plan" which provides a framework for evaluating information on severe accidents, including that developed through conduct of the Individual Plant Examinations (IPEs), for preparing and implementing severe accident operating procedures, and for training operators and managers in these procedures.
2
p w
L/
The " Accident 14anagement Plan" developed by-licensees for each plant will be expected to have four subsidiary objectives which are:
(1) Developing technically sound strategies for maximizing the effectiveness of personnel and equipment in preventing and mitigating potential severe accidents.
This includes ensuring-that guidance and procedures to implement these strategies are in place at all plants, (2) Assuring that installed instrumentation and equipment called for in the diagnosis and control of accidents beyond the design basis are. identified and assessed to determine their availability and capabilities, and the need for incremental improvements to existing systeus to assure their availabiiity is assessed, (3) Assuring that nuclear plant staff are' trained in the procedures, and guidance to follow in the event of an accident beyond the design basis of the plant, and utility management is trained and prepared to deal with severe accidents, and (4) Providing a technical basis for assessing the effectiveness of specific accident management j
strategies and capabilities.
I s
The NRC Accident fianagement Program is' aimed at promoting the'most effective use of available utility. resources (people and hardware) to prevent and mitigate severe accidents.
This would largely be achieved through incremental improvements in the existing emergency procedures and training programs, and by additional planning for severe accidents that could strengthen the support provided to the plant operating staff in case of a severe accident.- Hardware changes or other plant-modifications to reduce the frequency of severe accidents
.are not a central aim of this. program, although limited, minor modifications may be identified during the process of developing an Accioent lianagement Plan.
Elements of a Utility Accident 14anagement Plan To varying degrees, accident management capabilities diready exist dt all U.S. reactors, largely in response to regulatory requirements, implementation guidance, and review criteria, such as that set forth in 10 CFR 50.47 and NUREG-0737 Supplement 1.
However, these capabilities are not as strong and comprehensive as they might be 3
%c
-r
- - -. ~.
4 N
V V
through use of a more oisciplined approach. The risks associated with severe reactor accidents can be further reduced through implementation of utility accident management plans which incorporate improvements to current utility capabilities in five general areas:
Accident Management Procedures On the basis of existing PRAs the staff has identified several generic accident management strategies that can greatly enhance a licensee's ability to cope with the accident scenarios that tend to dominate risk in PRAs. This information will be previded to each licensec with the request that thej evaluate its benefits for their plants.
Training In Severe Accidents Operators, technical support staff, and managers responsible for responding in the event of an accident should be generally aware of the progression of severe accidents (i.e., their symptoms and timing) and shculd be proficient in potential response strategies. Licensees
- accident management plans are expected to be aimed at upgrading existing training programs as needed to ensure that training for these personnel includes an appropriate treatment of severe accident management. This will be done with as little impact as possible on the current training curricula for licensed plant operators.
Accident Management Guidance Each licensee will be expected to make available for the technical support staff and managers a set of guidance for diagnosing the progress of severe accidents and planning the appropriate response. The NRC staff will work with industry (e.g., NUMARC/EPRI) towards the development.of generic guidance to licensees in this area.
Instrumentation Licensees will be expected to review instrumentation changes that might be needeo at their plants in order i
to implement their accident management procedures.
Currently, the NRC Office of Nuclear Regulatory Research and EPRI are independently assessing the need for and availability of instrumentation during varicus accident scenarios. When comparing the 4
i results of this assessment to Regulatory Guide 1.97 which describes acceptable standards for post accident monitoring instrumentation, it is expected that the impact on existing instrumentation will be minimal.
Decisionmaking Responsibilities Each licensee's " Accident Management Plan" will be expected to include a review, and modification if necessary, of the plant's current decisionmaking authority for accident managemen? Strategies, to assure:
well-established, clear lines of authority and communications for severe accident conditions, assigned responsibilities for_ specific key decisions and established authority ano criteria for procedural I
overrides and ad-hoc equipment / procedure modifications.
Approach For Accident Management Implementation The staff intends to work with industry to define the scope and attributes of a utility accident management plan which meets the four major plan objectives, and to develop guidelines which oescribe the plant-specific implementation of such a plan. The principal interaction to date with industry on accident management has been through the Nuclear Management and Resources Council (NUMARC). A working group has been established by NUMARC to address the matter of severe accidents, with accident management being a high priority.
In addition, we expect NUMARC to provide industry's perspective and bring about the riecessary industry-supported initiatives on accident management.
The staff also plans to interact through NUMARC with the owners' groups for each reactor vendor since the prospective accident management procedures and equipment improvements are closely related to the emergency procedures guidelines that have already been developed by the owners' groups.
The regulatory mechanism for obtaining improvements in i
industry accident management capabilities will be through issuance of a generic letter. A draft of the generic letter will be circulated to utilities through NUMARC and to the public for comment. Additionally, the contents of the generic letter will be discussed at workshops 1
associated with the IPE generic letter. An outine of the eneric letter is attached to this Commission Paper ).
5
)
G It is the staff's view that many elements of accident
. management are sufficiently well understood and separate from the IPE analysis, that improvements in accident management c.n abilities can be realized prior to completion of an IPE.
Examples of such improvements include near term implementation of certain accident management procedures (to be included in the Generic Letter on accident management) and the revision of training curricula for emergency response personnel to include current insights on severe accident progression and phenomena.
In this regard, industry will be encouraged to implement accident management improvements as soon as practical.
The staff intends to be flexible with regard to implementation of aspects of accident _ management requiring plant-specific information to be learned from the IPE, recognizing that some licensees are well advanced in PRA-and severe accident studies for their plants, whereas other licensees are only beginning to consider severe accident analysis in response to the IPE Generic Letter.
Indeed, some licensees have already made significant advances in several of the general areas of accident management mentioned previously.
Thus, the details of the oevelopment and implementation of plant-specific accident management plans will be pursued with each licensee on an individual basis.
Implementation of the Accident Management Program is expected to be accomplished through an extension of existing industry programs, and evaluated through existing' regulatory mechanisms.
For instance, new training requirements could be integrated with the existing INPO training program. Generic accident management procedures coula be incorporated into emergency proceaure guidelines by the vendor owners groups and assessed by NRC according to existing channels of review. Other aspects of the program, such as hardware modifications ano implementation of computational tools, could be implemented through internal utility mechanisms such as their 10 CFR 50.59 process, ano would be subject to NRC audit and inspection in the usual way.
NRC Accident Management Research Program The NRC research program has an important role to play in contributing to the Accident Management Program. Our understanding of the physical progression of severe core damage accidents is incomplete, ana the NRC will rely on the research program to supply neeceo information and provide insights for dCCident management, partiCularly in the area of limiting 6
i
p i
)
(
i G
v' s
potential radioactive releases and stabilizing conditions should the reactor vessel be breached.
Research information j
will be oeveloped in this severe accident regime for many years, and as a result we can expect that licensees' accident management plans will be further upgraded as new research information is evaluated and a consensus reached that revisions to severe accident strategies can reduce risks further.
Research activities will center on assessing'the feasibility of various strategies that might be implemented by utilities to prevent or mitigate severe accidents, and on identifying those which should be considered for inclusion in utility accident management plans.
This will include an investigation of specific accident management strategies applicable to the period before the core would penetrate the reactor vessel (in-vessel accident management) and those applicable following postulated reactor vessel penetration (containment and release management).
In all cases, the design and operational requirements for strategy execution will be evaluated, but emphasis will also be given to examining potential circumstances under which certain operator actions could worsen !
accident consequences or adversely impact the ability to achieve a long-term, stable state. Nuch of the information needed for accident management research will be orawn from several existing NRC programs (e.g., the CPI and Severe Accident Research Programs), as well as existing programs in other countries.
l Research activities are divided into a short-term effort and a long-term program.
The short-term effort will support two principal elements in the closure plan:
(1) definition of example severe accident management strategies (primarily preventative) which can be formulated from existing insights reports, PRA studies, and completed research; and (2) development of a framework which defines the necessary components of a functioning utility severe accident management plan.
The products will be documents to be included as appendices to the generic letter on accident management.
The long-term research program includes those activities of a confirmatory nature, which are not required for closure.
The short-term program to achieve closure is i
proceeding effectively with the present knowledge base.
The long-term program can be considered as an augmentation i
to the existing Severe Accident Research Program (SARP) i and will draw heavily from the results of SARP.
In fact, SARP will provide improved knowledge and phenomenological modeling for such complex processes as steam explosions, 7
l
Y(3 Q
degraded core coolability, crust formation, hydrogen behavior, vessel depressurization, direct containment heating, etc., which will then be used to evaluate both the benefits and the adverse effects of candidate strategies. The research program has been formulated to provide the NRC staff with an organized, comprehensive basis for evaluation of generic accident management strategies. The program will define supporting studies needed from related programs (e.g.,
SARP, Human Factors research, PRA studies).
It will also integrate all appropriate results from these programs, and others related to accident management (e.g., CPI, IPE and international research), into a practical assessment for generic accident management strategies.
It is anticipated that these strategies will be primarily mitigative in nature. Upon completion, it will provide a reasonably detailed evaluation of both the instrumentation and training capabilities needed to support these strategies.
The long-term program will also evaluate the potential adverse effects that could occur if and when these strategies are applied.
Examples are strategies such as water addition to a degraded core and use of special emergency equipment.
Each must be analyzed carefully tc fully understand the consequences of these strategies.
Some of the specific issues to be addressed include:
instrumentation needs for proper diagnosis of the course of events and their observable symptoms; the capability of existing equipment to bring the reactor to a long-term stable state; the effect c,f timing of human action on the success of candidate strategies; the consequences of adding water to a degraded core; and the uncertainties in phenomenological knowledge and its consequences on strategy development.
Even though this research will concentrate or ti>
consequences of potential actions, it is currr expected that the ultimate guidance to the operating staff is to always add water during the course of a severe accident.
The NRC resources associated with reaching closure on accident management will be approximately 5 FTE of staff effort and about $500K of contractor assistance over the next twelve months. Resources for the long-term research p)rogram are $1.4M in FYB9 and $4M in FY90 (See Attachment 1.
Resource needs beyond FY90 are expected to be similar, and will be submitted for Commission approval as part of the normal budgetary process.
8
,s 1
1 Recommendation:
That the Commission note the approach proposed herein for implementing the accident management portion of the Integration Plan for Closure of Severe Accident Issues (SECY-88147), and approve the accident management research projects for FY89 and FY90.
Scheduling:
This paper is scheduled to be considered at an open meeting on January 23, 1989.
i
/Y ExecutiveDirect[or ctor Stell J
for Operations Attachments:
I 1.
NRC A/M Research Projects for FY89/90 i
2.
Outline of Generic Letter t
i Commissioners' corrents or consent should be prcvided directly to the Office of the Secretary by c.o.b. Frida".
1 1909.
Februar*.* 3, Commission Staff Office comments, if any, should be submitted to the Commissioners ULT Friday. Januarv 27, 1989, uith an information copy to the Office of the Secretary.
If the j
paper is of such a nature that it requires additional time for analytical revieu and comment, the Commissioners and the-Secretariat should be apprised of when comments may be expected.
DISTRIBUTIO'h Commissioners I
OGC OIA GPA 4
REGIOL*AL OFFICES i
ASLAP SECY 9
m--
N P
I, m
ATTACHMENT 1 HRC A/M RESEARCH PROJECTS FOR FY 89/90 A.
Ongoing Projects 1.
Depressurization to minimize direct containment heating.
An analytic study of station blackout (TMLB') at the Surry plant, using depressurization through the PORV.
Results indicate that the pressurizer surge line could overheat and rupture, leading to reasonably early depressurization.
2.
LWR recriticality.
An analytic study of injecting unborated water, as well as water with increasing levels of boron concentration, into an LWR core:
(1) when the control blades have melted, but the fuel rods are intact; and (2) when the core geometry has degraded.
3.
Instrumentation requirements.
Development and demonstration of a structured methodology to determine adequacy of existing plant instrumentation to monitor l
severe accidents.
1 1
4.
Report on candidate A/M strategies; Appendix I to the Generic Letter.
5.
Report on framework of an A/M program; Appendix II to the Generic Letter.
B.
NewProjects(FY89/90)
NRC has a longer-term program to conduct research directly related to A/M.
This longer-term research would focus on assuring integration of completed ongoing resarch (from CPI, SARP, PRA, and international cooperative programs) to effectively evaluate A/M guidance, while simultaneously -
evaluating potential disaovantages of seemingly advantageous accident management actions. Our longer-term research program has been thought out in detail and discussed with the various contractors scheduled to conduct the research.
The following tasks will be covered by this program in FY 89/90.
In-vessel A/M candidate procedures Ex-vessel A/M candidate procedures Information needs for A/M A/11 framework implementation Bounds of coolability of a degraded core Uncertainties affecting A/M BWR A/ft insights (MARK-I, III)
Strategies related to DCH Diagnostic Computational Aids Guidelines for industry audit i
Support for Plant Exercises The total funding required to complete these tasks is $1370K in FY 89 and
$3935K in FY 90.
Outline of Generic Letter
SUBJECT:
Accident Management (Supplement to Generic Letter 88-20, " Individual Plant Examination.")
j I.
Objective Licensees shall certify to the NRC that they have put in place an
" Accident Management Plan" providing a framework for evaluating frfe9 nation on severe accidents, for preparing and implementing severe accident procedures, and for training operators and managers in those procedures.
4 2.
Elements of a utility Accident Management Plan should include the following:
Procedures: Each licensee should, in the context of performing their IPE, evaluate and implement severe accident procedures. The staff has identified general strategies which should be included and has compiled a list of " example" procedures (see Appendix I) which should be considered for early implementation (e.g., prior to completion of the IPE where the licerisee determines this to be of benefit to his plant protection and to public safety).
Training in severe accidents should be provided for operators, technical staff and managers responsible for responding in the event of a severe accident.
- . Guidance and computational aids for diagnosing and responding to accidents should be provided to technical support staff and managers.
An assessment of the need for and availability of instrumentation should be performed.
A review of utility decisionmaking processes for severe accident response should be conducted.
3.
NRC/NUMARC Interaction NRC staff will work closely with industry groups (e.g., owner's groups ard EPRI) through the NUMARC Severe Accident Working Group (SAWG) to better define the attributes of an Accident Management Plan and to develop guidelines that define a generic framework for a utility to evaluate and organize its resources to prepare for and respond to severe accidents (to be supplied as Appendix II). These <;cidelines, to be developed as part of an industry initiative, shouid address each of the key attributes of a utility Accident Management Plai and provide a basis fur responding to the generic letter.
l
l t
i 4.
Utility Response Within 90 days of' receipt of this letter, licensees should submit a response certifying their comitment to implement an Accident Management Plan that provides a framework for evaluating and implementing severe accident procedures.
In each case, a best estimate schedule should be included for implementing their Accident Management Plan.
5.
Regulatory Basis 10 CFR 50.54(F) 6.
Appendices:
I.
Candidate Procedures for Near Term Implementation (draft attached)
II. Framework of an Accident Management Program (under develapment)
1
?.
i APPENDIX 1 l
i Each utility.should systematically seek to identify and implement effective procedures (and associated hardware) to optimize the plant's accident management !
s resources.
These procedures should at least address three global strategies derived from operational experience and PRA insights which have significant i
l potential for reducing,nlant risk:
f i
- Procedures for conserving and/or replenishing limited utility resources during the course of an accident.
These resources would include, for example battery capacity, borated water, and compressed air.
i i.
- Procedures for using plant systems and components for innovative applications during an accident. This would include enabling crossties of support systems or the use of fire systems, or CRD pumps (in the case of a BWR), for decay heat removal.
In addition, this cM1ory includes 4
i proceaures to connect alternate electrical power sources to meet
, critical :afety needs during an accident scenario.
' ' Procedures for defeating interlocks and overriding component protective trips in emergency situations. An example of this strategy would be the ability to reopen MSIV's in a BWR ATWS event.
These three strategies, and others as apprzpriate, should be applied to each of the applicable major safety functions of the plant:
reactivity control, coolant inventory control, heat removal and containment perfomance; as well as to the prihciple support functions: RCS depressurization, electric power, i
I equipment cooling, and air systems.
In the context of performing the IPE, i
each utility should be alert to identify and implement effective procedures i
associated with these strategies.
Your evaluation should not be
]
limited to the strategies given in Table 1.
For instance, procedures to maintain containment function and delay or prevent possible early containment i
failure should also be assessed.
Table I contains a list of " example" procedures for each global strategy.
These procedures are categorized according to the safety functions they relate to and the types of plants they apply to. These procedures are to be evaluated on a plant-specific basis and considered for near tem implementation prior to completion of the IPE (absent sound arguments to the contrary).
This letter assumes implementation of existing EPGs addressed in the ATWS ard Station Blackout rules, in addition to feed-and-bleed and BWR containment venting provisions of the EPGs.
There may be some areas of overlap between these existing emergency operating procedures and the procedures described in Table I, however, the NRC believes the benefits from additional. utility review and strengthening of existing procedures would include both reduction of risk magnitude and uncertainty from severt accidents and moreover would outweigh the costs of any duplication of effort.
TABli I Genes ic Acciden,' Managment Procedures f or t
ficar Term Evaluation and implementation Attected Plant Adicability Principal Safety Objective F
---'0WR PRR
~PiivintibiI-~Mitigit ijn-SafetL unttion i;lohol Strate.y xameple Procedure j
1.
Conser ving end Iteplenishing iiniited Resounes Procedure to refill LW51 with Luciant Inventory X
X X
X boroted water, or C5T with condensate.
Containn. erit Performance Assure adequate supply of boron on site.
- Maintain itL5 suction to condensate toolant Inventory X
X systems to avoid pump failure due tohighsupp{essionpool tecipera tu s e.
Procedures for throttlino Coolant Inventory X
X X
containment sprays to conserve water for.. ore injection.
Proceduces to conserve battery llu tric Power X
X X
X capacity he shedding non-essential loads.
" Procedures for use of porteble Electric Power X
X X
X battery chargers or other power sources to recharge batteries.
I lhis muy require re-examilnation to defeat existing interlocks
P T ABI C_ I_(bHilHilt DJ Affected PlantApdicability Principal Safety Objectice
~~ digali6n~
1 unction
~ ~~ BUR ~PSR
~Prevenil6Ti M
F Globa l Strotegy lxample Procedure Safet Procedure:. tu enable energency A i r / h.,
X X
X X
replenishmi rit of gas supply, or otherwise assure operability of air operated compnents.
' Peutedures to enet,le early detection, Cuolant leiveritory X
X X
X isolation, or otherwise mitigste the Contairunent Perf.
effects of an iriterfacing LOCA.
- 11. Ilse ut Syste pn/ Loa.ponents in innovative Applications
- Piutedure, to esieble (en:rgency use of available pumps to accomplish sdfety functions
- lite of diesel f ire systems f or Coolont inventory X
X X
X injer t ion to UWR core, PWR Heat Renoval steam. generators, or containnient Coritainment Perf.
sprays.
- lise i.I CRD pumps (BWR) or toolant Inv(ntory X
X X
chaiying pumps (PWR) for core injection.
- Ilse sit altertiate injection (e.g.,
Coolant triventury X
X X
hydrn test 1 ump when RCP seal cooling is lost
- Protedures (end associated hardware) lleet Removal X
X X
to enable emergency crosstie of Coolant Inventory setvice water and CLW to RHR (BUR) or feedwater (PWR).
Risk signf icence of seal f ailure is strongly deper. dent on the seal design.
]
TABLE I (CCHTINUED)
Affected Plant A py cabilig P rir:
al Safet Objective i.loba l Strategy imangile_ Procedure Salety funcyon
9
_PER ~
~ Preg.lich---
I g ti6{
M h
- Use of condensate, or starttip pumps llea t l<en. ova l X
X for fordwater injection.
Procedures (and hardware) to enable liedt Remaval X
X X
X emergency ti,nnection of available AC power soones to nieet critical salety needs.
- Ilse of diesel generator or gas Coulant Inventory X
X turbine generotor to drive CHD pumps for cure injection.
- Procedures to enable eniergeticy Electric Power X
X X
X crosstie of AC power between two units or to onsite gas turbine generator.
- Procedures to enable eniergency connection of injection systems to alternate water sources
- Procedures to assure oppropriate toolant Inventory X
X X
recirculation switchover and to llea t Renioval cope with the failure to switchover in thrA.
- Procedures to enable emergency lleat renioval X
X X
(unnection of service water or Equipnient Cooling X
X X
X feedwoter systems to rivers, reservoirs or mainicip61 wate ' systtais, l'roceduees for Reactivity Control
'.4 TABL E I (00t:llHUED)
Alfected Plant A litability Princi Latet Objtctive t.lobol Straterjy (xair.[ile Procedure 5.fety functi6rt
"~~ BR
~PWR Hiven on lilda Hoh-X X
X
- Procedure to initiate SLLS it.
Penttivity Conttul case of potential core damage and to guard agaltist baron dilution when core injection is restored.
- Ensun obundalit siipply of Reattivity Control X
X X
X boreted makeup for long-team atCidt tet Control.
111. I.efeatisig leiterlotLs and losponent Pratettive Trips iri linergencies.
- Procedures to reopen MSIVs acid litat Removal h
X X
lurbine Bypess Valves to regair.
the t.ondeit,tr as a heat sink.
X X
" Procedure to extered RCIC Cuolant triveritory availability by citiier raistrig the turbine exhaust piessure trip set point, or overriding the trip function.
X X
- l s uedure, tu es.oble eniergency All byp6ss of protettive trips for diesel generato, es.d injection pumps.
p
/]
ENCLOSURE h[ a ntcy*':,
~
UNITED STATES
! g.,.,7 7 j NUCLEAR REGULATORY COMMISSION j
W ASmN GTON. D.C. 20555
%..v4/
February 28, 1989 OFFICE OF THE SECRETARY MEMORANDUM FOR:
Victor Stello, Jr., Executive Director for Operations g ;.15 Msamuel J. Chilk, Secretary a
FROM:
SUBJECT:
STAFF REQUIREMENTS - SECY-89 STAFF PLANS FOR ACCIDENT MANAGEMENT REGULATORY AND RESEARCH PROGRAMS This is to advise you that the Commission, with all Commissioners agreeing, supports the principles of accident management and has approved your recommendation to proceed with the FY 1989 and 1990 accident management research program.
However, the Commission requests that, in the future when the staff seeks Commission approval for year to year expenditure of resources, the staff provide the Commission with a breakdown of the dollars budgeted for each of the individual projects.
The Commission provides the following direction with respect to accident management.
1.
The implementation of the accident management program should be fully integrated into the Individual Plant Examination (IPE) process.
Staff should provide reactor licensees with lemons learned information on risks associated with severe core damage accidents and generic accident management strategies which may be appropriate.
Staff should ensure, to the extent possible, that suggested generic accident management strategies are not likely to detract from safety.
Further, staff should caution licensees to ensure that plant specific implementation of suggested generic accident management strategies does not detract from safety.
This information should be provided on a schedule so that licensees have these insights to consider in conjunction with development of their IPEs.
This information should also be updated as we gain further insights from our research program.
- f h 0 [f 0 h 0 + %
)
(m
\\
3 4
2 4
2.
Staff should establish a well defined mechanism for exchange of-information between the NRC and industry groups regarding developmental activities and research.
Staff should continue to work with NUMARC to define the scope and content of an acceptable accident management program *and to deve16p a plan for incorporating plant-'
specific actions into such a program.
This program should t
be developed on a schedule such that the accident management framework will be available for implementation by 1.icensees inparallel with conducting their IPE.
The ACRS should be kept currently informed of the progress of this program so that its comments can be considered.
j 3.
Staff's plan for implementation of the accident management i
program should be presented for Commission review after the staff has more thoroughly defined the scope and content of such a program, and before licensees are requested to implement the program.
4 (EDO)
(SECY SUSPENSE:
6/30/89) t l
4.
Staff should incorporate the commission's directives in the May 1989 update of the Five Year Plan.
t j
(EDO)
(SECY SUSPENSE:
5/89) 1 Copies:
Chairman Zech Commissioner Roberts Commissioner Carr Commissioner Rogers Commissioner Curtiss OGC GPA l
y ENCLOSURE 5
t UNITED STATES P
/o NUCLEAR REGULATORY COMMISSION y,
I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20555 November 20, 1989 The Honorable Kenneth M. Carr Chairman U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Chairman Carr:
SUBJECT:
DRAFT SUPPLEMENT NO. 2 TO GENERIC LETTER 88-20. " ACCIDENT MANAGEMENT STRATEGIES FOR CONSIDERATION IN THE INDIVIDUAL PLANT EXAMINATION PROCESS" During the 355th meeting of the Advisory Comittee on Reactor Safeguards, November 16-18, 1989, we discussed the subject document with the NRC staff. We also reviewed a draft NUREG/CR report entitled, " Assessment of Candidate Accident Management Strategies," that the staff proposes to send as an enclosure with the supplement to the generic letter.
We had the benefit of these documents which are referenced.
Our Subcommittee on Severe Accidents met on September 20, 1989 to discuss this matter.
We conclude that the information in these two dor.uments will be useful to licensees in the process of performing Individual Plant Examinations, and we agree that the documents should be issued.
The draft NUREG/CR report referred to describes strategies for accident management that are said to be PRA based.
However, the report does not include information on the risk reduction that might be attributed to the i
strategies.
This information would be useful to those considering the strategies.
We recommend that this information be added if it is rea-sonably retrievable from existing sources.
We observe that a number of the strategies described in the draft NUREG/CR report either overlap or are very similar to the content of the emergency operating procedures that are either being developed or are already in place in many plants.
We believe that labelling these procedures as accident management strategies where others label them as emergency operating procedures is likely to lead to confusion on the part of both i
the NRC staff and the industry.
Sincerely, l
Forrest J. Remick Chairman k] ll ?CC:l E//0
I
/
/
The Honorable Kenneth M. Carr 2-November 20, 1989 References i
1.
U.S. Nuclear Regulatory Comission, " Accident Management Strategies for Consideration in the Individual Plant Examination Process " Draft Supplement No. 2 to Generic Letter 88-20, dated November 8,1989 (Predecisional) 2.
U.S. Nuclear Regulatory Comission, " Assessment of Candidate Accident Management Strategies," Draft NUREG/CR Report (Unnumbered), Prepared by BNL, October 1989 6