Protecting People and the EnvironmentUNITED STATES NUCLEAR REGULATORY COMMISSION
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555-0001
September 30, 1996
NRC GENERIC LETTER 96-06: ASSURANCE OF EQUIPMENT OPERABILITY AND CONTAINMENT
INTEGRITY DURING DESIGN-BASIS ACCIDENT CONDITIONS
Addressees
All holders of operating licenses for nuclear power reactors, except for those
licenses that have been amended to possession-only status.
Purpose
The U.S. Nuclear Regulatory Commission (NRC) is issuing this generic letter to
(1) notify addressees about safety-significant issues that could affect con-
tainment integrity and equipment operability during accident conditions,
(2) request that all addressees submit certain information relative to the
issues that have been identified and implement actions as appropriate to
address these issues, and
(3) require that all addressees submit a written response to the NRC relative
to implementation of the requested actions.
Background
As a result of recent NRC inspection activities, licensee notifications, and
event reports, several safety-significant issues have been identified that
have generic implications and warrant action by the NRC to assure that these
issues have been adequately addressed and resolved. In particular, the
following issues are of concern:
(1) Cooling water systems serving the containment air coolers may be exposed
to the hydrodynamic effects of waterhammer during either a loss-of-
coolant accident (LOCA) or a main steamline break (MSLB). These cooling
water systems were not designed to withstand the hydrodynamic effects of
waterhammer and corrective actions may be needed to satisfy system design
and operability requirements.
(2) Cooling water systems serving the containment air coolers may experience
two-phase flow conditions during postulated LOCA and MSLB scenarios. The
heat removal assumptions for design-basis accident scenarios were based
on single-phase flow conditions. Corrective actions may be needed to
satisfy system design and operability requirements.
9609250096. GL 96-06
September 30, 1996
Page 2 of 10
(3) Thermally induced overpressurization of isolated water-filled piping
sections in containment could jeopardize the ability of accident-
mitigating systems to perform their safety functions and could also lead
to a breach of containment integrity via bypass leakage. Corrective
actions may be needed to satisfy system operability requirements.
The sections that follow contain additional background information about each
of these issues.
Waterhammer
On February 13, 1996, the Pacific Gas and Electric Company (PG&E, the licensee
for Diablo Canyon Units 1 and 2), determined that component cooling water,
which is circulated through the containment air coolers, could flash to steam
in the cooler unit cooling coils during a design-basis LOCA with a concurrent
loss of offsite power (LOOP) or with a delayed sequencing of equipment. This
condition was reported to the NRC in Licensee Event Report (LER) 1-96-005,
dated April 26, 1996.
The Diablo Canyon units have five containment air coolers in each containment,
these are typically used during normal plant operation to prevent excessive
containment temperatures. The containment air coolers are also automatically
initiated engineered safety features that are relied upon to help maintain
containment integrity by performing their heat removal function during
postulated accident conditions. The air coolers in the Diablo Canyon units
transfer heat from the containment to the respective unit's component cooling
water system (a closed-loop system).
PG&E reported that, during a postulated design-basis LOCA with a concurrent
LOOP, the component cooling water pumps and the air cooler fans will tempo-
rarily lose power (an expected condition). The component cooling water flow
stops almost immediately, while the fans coast down over a period of minutes.
The first air cooler fan will restart on slow speed approximately 22 seconds
after the LOOP and the component cooling water pumps will restart 4 to
8 seconds later. In this scenario, the high-temperature containment atmo-
sphere will be forced across the containment air cooler's cooling coils for up
to 30 seconds with no forced component cooling water flow through the coolers.
PG&E determined that the stagnant component cooling water in the containment
air coolers may boil and create a substantial steam volume in the component
cooling water system. As the component cooling water pumps restart, the
pumped liquid may rapidly condense this steam volume and produce a water-
hammer. The hydrodynamic loads introduced by such a waterhammer event could
be substantial, challenging the integrity and function of the containment air
coolers and the associated component cooling water system, as well as posing a
challenge to containment integrity. As corrective action, PG&E has installed
a nitrogen pressurization system on the component cooling water head tank to
increase the margin to boiling.
On June 20, 1996, Westinghouse Electric Corporation issued Nuclear Safety
Advisory Letter NSAL-96-003, "Containment Fan Cooler Operation During a Design
Basis Accident," to alert its customers to the potential safety issue that was
. GL 96-06
September 30, 1996
Page 3 of 10
identified by PG&E (Westinghouse is the reactor vendor for the Diablo Canyon
units). In NSAL-96-003, Westinghouse recommended that licensees review their
containment cooling systems to determine if their safety-related containment
air coolers are susceptible to waterhammer.
On July 22, 1996, the Connecticut Yankee Atomic Power Company (CYAPC, the
licensee for the Haddam Neck nuclear power plant) declared all four of the
containment air coolers at the Haddam Neck plant inoperable and initiated a
plant shutdown in accordance with Technical Specification requirements. The
containment air coolers at the Haddam Neck plant are the only components that
are credited for post-accident containment heat removal, and station service
water (an open-loop system) is the cooling medium for the containment air
coolers. The containment air coolers were declared inoperable after CYAPC
completed its review relative to Westinghouse NSAL-96-003. The licensee's
analysis predicted hydrodynamic loads in the service water system from
waterhammer that exceeded piping and support structural limits.
On August 12, 1996, the staff issued Information Notice (IN) 96-45, "Potential
Common-Mode Post-Accident Failure of Containment Coolers," to alert addressees
to the potential failure mode of the containment air coolers and their
associated cooling water systems. IN 96-45 discussed the information that was
reported by PG&E and CYAPC relative to the Diablo Canyon and Haddam Neck
plants, respectively, and attached a copy of Westinghouse letter NSAL-96-003.
Two-Phase Flow in Safety-Related Piping and Components
In July 1996, the NRC issued Inspection Report 50-213/96-201, "Special
Inspection of Engineering and Licensing Activities at Haddam Neck-Connecticut
Yankee." Among other things, the report identified an issue relative to two-
phase flow in the station service water system. The inspection team reviewed
the service water system flow models, calculations, and operational data and
found that some steam may be produced in the service water system as the
service water flows through the containment air coolers during design-basis
accident conditions. However, the licensee's service water system model and
calculations only assumed single-phase flow conditions (liquid phase only) and
did not consider two-phase flow conditions (both steam and liquid present).
The licensee is currently evaluating the system to determine whether or not
corrective actions are needed.
On July 23, 1996, the Wisconsin Electric Power Company submitted information
regarding two-phase flow in the service water system at the Point Beach
nuclear plant during a design-basis LOCA. The licensee's preliminary evalua-
tions concluded that after the cooling water is heated via heat transfer from
the containment air coolers, some steam could be formed at the air cooler
outlet throttle valves. This two-phase mixture (steam and water) would result
in a higher frictional pressure drop in the service water return piping and
would ultimately affect the service water flow and the heat removal capabili-
ties of the containment air coolers. Steam formation due to low pressure and
high temperature in the service water system could reduce the service water
flow rates through the containment air coolers to values below those needed to
. GL 96-06
September 30, 1996
Page 4 of 10
satisfy design-basis heat removal requirements. The licensee is completing
more detailed analyses to determine if immediate corrective action is war-
ranted.
On August 20, 1996, the Public Service Electric and Gas Company (the licensee
for Salem 1 and 2) notified the NRC of a condition that is not bounded by the
existing design basis for the Salem nuclear power plants (EN 30900). The
licensee reported that because the service water isolation valves for the
nonsafety-related turbine loads do not start to close until approximately
30 seconds into the emergency loading sequence, the service water system may
not be able to supply sufficient flow for the containment cooling function
during accident conditions. The licensee determined that the initial heat
transfer rates through the containment air coolers could result in additional
"flow restrictions" in the air cooler tubes, further decreasing the flow of
service water through the containment air coolers as a result of the higher
frictional pressure drop caused by two-phase flow. At the time of the
licensee's notification, the Salem units were shut down for refueling.
Overpressurization of Isolated Piping Sections
On July 3, 1996, Duquesne Light Company (the licensee for Beaver Valley
Units 1 and 2) notified the NRC that during surveillance testing of a compo-
nent cooling water inlet valve to the RHR heat exchanger on Unit 1, the motor-
operated butterfly valve located inside the containment would not open
(EN 30833). The licensee found that pressure in the piping section between
this valve and a closed manual butterfly valve located outside the containment
measured slightly higher than the system design pressure. After the pressure
in this isolated section of piping was relieved by opening a drain valve, the
remotely operated butterfly valve was opened without any trouble. The
licensee concluded that pressure in the isolated section of piping increased
when the trapped water was heated up by increased ambient temperatures. The
section of piping was isolated in the spring when the unit was shut down and
ambient temperatures were much lower than temperatures that existed in the
summer after the plant was returned to power operation and ambient tempera-
tures reached about 32 øC [90 øF].
On July 19, 1996, the Maine Yankee Atomic Power Company (MYAPC, licensee for
the Maine Yankee nuclear plant) notified the NRC of a condition that was
outside the plant design basis (EN 30769). The primary component cooling
water (PCCW) system at the Maine Yankee plant has a nonsafety-related subdivi-
sion that serves the containment fan coolers (not needed for accident mitiga-
tion), and a safety-related subdivision that serves ECCS equipment. The
nonsafety-related subdivision of PCCW has a swing-check valve at the contain-
ment inlet (supply) penetration, and an air-operated valve at the containment
outlet (return) penetration. During a design-basis LOCA, the containment
isolation logic initiates closure of the air-operated outlet valve, thereby
stopping the flow of water. The licensee has determined that heat from the
containment accident environment could cause the PCCW in the containment fan
coolers between the inlet check valve and closed air-operated outlet valve to
expand, rupturing this portion of the PCCW system. Water from the PCCW system
is then able to flow through the supply check valve for the containment fan . GL 96-06
September 30, 1996
Page 5 of 10
coolers and out the rupture, rendering the PCCW system inoperable and jeopar-
dizing safety-related equipment that is cooled by the safety-related division
of the PCCW system. Upon recognizing this postulated scenario, the licensee
promptly shut down the Maine Yankee plant. To correct this, the licensee
plans to install a pressure relief valve on each of the six containment fan
cooler PCCW branch lines downstream of the supply check valves.
On August 20, 1996, the staff issued Information Notice (IN) 96-49, "Thermally
Induced Pressurization of Nuclear Power Facility Piping," to alert addressees
to the potential for safety-related piping to become overpressurized during
accident conditions. IN 96-49 discusses the information reported by Duquesne
Light Company and MYAPC relative to Beaver Valley Unit 1 and the Maine Yankee
plant, respectively.
Discussion
The issues discussed in this generic letter pertain to situations that may not
be bounded by the applicable system design capabilities and for which correc-
tive actions may be needed to satisfy equipment design and operability
requirements. The sections that follow contain additional discussion about
each of these issues.
Waterhammer
At many plants, containment air coolers satisfy a significant safety function
by removing heat from the containment and reducing post-accident containment
pressure. The hydrodynamic loads imposed by waterhammer can be substantial,
challenging the integrity and function of the containment air coolers and the
associated cooling water system, as well as posing a challenge to containment
integrity. Waterhammer in cooling water systems associated with nonsafety-
related containment air coolers can also challenge containment integrity by
creating a containment bypass flow path, and interfacing safety-related
systems can be affected. During this accident scenario, the steam that is
produced in the containment air coolers may accumulate in other parts of the
cooling water system, restricting flow as well as causing waterhammer damage.
Plant vulnerability to the postulated waterhammer scenario depends on a number
of factors, such as piping configuration, how long it takes for the flow of
cooling water to stop, the coastdown rate of the fans in the containment fan
coolers, the operating pressure and pressure decay rate of the cooling water
system, how long it takes to establish forced cooling water flow, the contain-
ment temperature profile, and other site-specific parameters.
The postulated failure scenario is applicable to both LOCA and MSLB events
that involve a loss of offsite power, a loss of cooling water flow to the
containment air coolers (e.g., one train of cooling water inoperable), or the
sequencing of equipment that can affect the containment cooling function.
Steam formation and waterhammer in cooling water systems associated with
safety-related and nonsafety-related containment air coolers may not require a
loss of offsite power for this scenario to be valid.
. GL 96-06
September 30, 1996
Page 6 of 10
Two-Phase Flow in Safety-Related Piping and Components
Two-phase flow (i.e., both steam and liquid) in cooling water systems associ-
ated with the containment air coolers can significantly interfere with the
ability of the containment air coolers to remove heat under design-basis
accident conditions, and can interfere with the cooling of other safety-
related components. These cooling water systems were designed assuming
single-phase flow conditions (i.e., liquid only) and containment heat transfer
analyses are based on this assumption. Two-phase flow is a much more complex
situation to deal with analytically than single-phase flow and involves
additional hydrodynamic loading considerations as well as flow, heat transfer,
systems interaction and erosion considerations. Additionally, the steam that
is formed during two-phase flow can accumulate in the cooling water system,
restricting flow and resulting in a waterhammer as discussed above.
Overpressurization of Isolated Piping
Because of its thermal expansion, water heated while it is trapped in isolated
piping sections is capable of producing extremely high pressures. This
phenomenon is typically a design consideration. Piping design codes as far
back as U.S.A. Standard (USAS) B31.1 (1967), have explicitly recognized the
need to consider the effects of heating fluid that is trapped in an isolated
section of piping. The potential for thermally induced expansion of fluid
trapped in valve bonnets was one reason for issuing Generic Letter (GL) 95-07,
"Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate
Valves." In addition, several information notices (INs) have been issued
discussing the pressurization of water trapped in valve bonnets, including
IN 95-14, "Susceptibility of Containment Sump Recirculation Gate Valves to
Pressure Locking," IN 95-18, "Potential Pressure-Locking of Safety-Related
Power-Operated Gate Valves," IN 95-30, "Susceptibility of LPCI and Core Spray
Injection Valves to Pressure Locking," and IN 96-08, "Thermally Induced
Pressure Locking of a HPCI Gate Valve."
The potential for systems to fail to perform their safety functions as a
result of thermally induced overpressurization is dependent on many factors.
These factors include leak tightness of valve seats, bonnets, packing glands
and flange gaskets; piping and component material properties, location and
geometry; ambient and post-accident temperature response; pipe fracture
mechanisms; heat transfer mechanisms; relief valves and their settings; and
system isolation logic and setpoints. Engineering design and modification
evaluations, which include systematic evaluation of heat input to systems and
components with consideration of factors such as those just noted, can detect
conditions which may influence system operability under normal operating,
transient, and accident conditions.
Under the "single-failure concept," failure due to overpressurization does not
preclude consideration of additional active and passive failures in the same
and other systems in evaluating plant response to a postulated accident. If
relief valves are installed to prevent overpressure conditions, consideration
must be given to the effects of stuck-open relief valves and associated
environmental flooding and radiation hazards.
. GL 96-06
September 30, 1996
Page 7 of 10
Requested Action(s)
Addressees are requested to determine:
(1) if containment air cooler cooling water systems are susceptible to either
waterhammer or two-phase flow conditions during postulated accident
conditions;
(2) if piping systems that penetrate the containment are susceptible to
thermal expansion of fluid so that overpressurization of piping could
occur.
In addition to the individual addressee's postulated accident conditions,
these items should be reviewed with respect to the scenarios referenced in the
generic letter.
With regard to waterhammer, addressees may find Volumes 1 and 2 of NUREG/CR-
5220, "Diagnosis of Condensation-Induced Waterhammer," dated October 1988,
informative and useful in evaluating potential waterhammer conditions.
If systems are found to be susceptible to the conditions discussed in this
generic letter, addressees are expected to assess the operability of affected
systems and take corrective action as appropriate in accordance with the
requirements stated in 10 CFR Part 50 Appendix B and as required by the plant
Technical Specifications. GL 91-18, "Information to Licensees Regarding Two
NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming
Conditions and on Operability," dated November 7, 1991, contains guidance on
the review of licensee operability determinations and licensee resolution of
degraded and nonconforming conditions.
Requested Information
Within 120 days of the date of this generic letter, addressees are requested
to submit a written summary report stating actions taken in response to the
requested actions noted above, conclusions that were reached relative to
susceptibility for waterhammer and two-phase flow in the containment air
cooler cooling water system and overpressurization of piping that penetrates
containment, the basis for continued operability of affected systems and
components as applicable, and corrective actions that were implemented or are
planned to be implemented. If systems were found to be susceptible to the
conditions that are discussed in this generic letter, identify the systems
affected and describe the specific circumstances involved.
Required Response
Within 30 days of the date of this generic letter, addressees are required to
submit a written response indicating: (1) whether or not the requested
actions will be completed, (2) whether or not the requested information will
be submitted and (3) whether or not the requested information will be sub-
mitted within the requested time period. Addressees who choose not to . GL 96-06
September 30, 1996
Page 8 of 10
complete the requested actions, or choose not to submit the requested informa-
tion, or are unable to satisfy the requested completion date, must describe in
their response any alternative course of action that is proposed to be taken,
including the basis for establishing the acceptability of the proposed
alternative course of action and the basis for continued operability of
affected systems and components as applicable.
Address the required written reports to the U.S. Nuclear Regulatory Commis-
sion, ATTN: Document Control Desk, Washington, D.C. 20555-0001, under oath or
affirmation, under the provisions of Section 182a, Atomic Energy Act of 1954,
as amended, and 10 CFR 50.54(f). In addition, send a copy to the appropriate
regional administrator.
Backfit Discussion
Title 10 of the Code of Federal Regulations (10 CFR) Part 50 (Appendix A) and
plant licensing safety analyses require and/or commit that the addressees
design safety-related components and systems to offer adequate assurance that
those systems can perform their safety functions. Specifically, 10 CFR
Part 50, (Appendix A, Criterion 38) specifies a "system to remove heat from
the reactor containment. The safety function of this system is to rapidly
reduce pressure and temperature in the containment following any loss-of-
coolant accident and to maintain them at acceptably low levels." Addition-
ally, Criterion 44 of Appendix A specifies a "system to transfer heat from
structures, systems, and components important to safety. The system safety
function shall be to transfer the combined heat load of these structures,
systems and components under normal operating and accident conditions." The
heat load values as defined in final safety analysis reports are based on
single-phase flow assumptions for the containment air cooler cooling water
systems. The potential for waterhammer and two-phase flow raises concerns
that these systems will not meet their design-basis requirements as specified
in 10 CFR Appendix A, Criteria 38 and 44. Further, 10 CFR Part 50 Appendix A,
Criteria 1 and 4 specify that safety-related systems be designed to offer
adequate assurance that those systems can perform their safety functions under
accident conditions. Accordingly, licensees are required to ensure that the
containment air coolers and their associated cooling water systems that may be
affected by waterhammer or by two-phase flow are capable of performing their
required safety functions and that containment integrity will be maintained.
Licensees are also required either by their commitment to USAS B31.1 or the
American Society of Mechanical Engineers (ASME) Code for piping design or by
virtue of 10 CFR 50.55a, which endorses various editions of the ASME Boiler
and Pressure Vessel Code, to comply with design criteria which specify that
piping systems which have the potential to experience pressurization due to
trapped fluid expansion shall either be designed to withstand the increased
pressure or shall have provisions for relieving the excess pressure. The
potential for overpressurization raises concerns that these piping systems
will not meet their design code criteria.
The actions requested in this generic letter are considered compliance
backfits under the provisions of 10 CFR 50.109 and existing NRC procedures to . GL 96-06
September 30, 1996
Page 9 of 10
ensure that containment integrity will be maintained and that safety-related
components and piping systems are capable of performing their intended safety
functions and satisfying their licensing-basis code criteria, respectively;
and that containment integrity and these safety-related piping systems and
components will not be adversely affected by the occurrence of waterhammer,
two-phase flow, or thermal overpressurization that may occur in safety-related
and nonsafety-related systems that penetrate containment. In accordance with
the provisions of 10 CFR 50.109 regarding compliance backfits, a full backfit
analysis was not performed for this proposed action; but the staff performed a
documented evaluation which stated the objectives of and reasons for the
requested actions and the basis for invoking the compliance exception. See
also 10 CFR 50.54(f) . A copy of this evaluation will be placed in the NRC
Public Document Room.
Federal Register Notification
A notice of opportunity for public comment was not published in the
Federal Register because of the urgent nature of the generic letter. However,
comments on the actions requested and the technical issues addressed by this
generic letter may be sent to the U.S. Nuclear Regulatory Commission, ATTN:
Document Control Desk, Washington, D.C. 20555-0001.
Paperwork Reduction Act Statement
This generic letter contains information collections that are subject to the
Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). These information
collections were approved by the Office of Management and Budget, approval
number 3150-0011, which expires on July 31, 1997.
The public reporting burden for this collection of information is estimated to
average 300 hours per response, including the time for reviewing instructions,
searching existing data sources, gathering and maintaining the data needed,
and completing and reviewing the collection of information. The U.S. Nuclear
Regulatory Commission is seeking public comment on the potential impact of the
collection of information contained in the generic letter and on the following
issues:
(1) Is the proposed collection of information necessary for the proper
performance of the functions of the NRC, including whether the informa-
tion will have practical utility?
(2) Is the estimate of burden accurate?
(3) Is there a way to enhance the quality, utility, and clarity of the
information to be collected?
(4) How can the burden of the collection of information be minimized, includ-
ing the use of automated collection techniques?
. GL 96-06
September 30, 1996
Page 10 of 10
Send comments on any aspect of this collection of information, including
suggestions for reducing this burden, to the Information and Records Manage-
ment Branch, T-6F33, U.S. Nuclear Regulatory Commission, Washington, D.C.
20555-0001, and to the Desk Officer, Office of Information and Regulatory
Affairs, NEOB-10202 (3150-0011), Office of Management and Budget,
Washington, D.C. 20503.
The NRC may not conduct or sponsor, and a person is not required to respond
to, a collection of information unless it displays a currently valid OMB
control number.
If you have any questions about this matter, please contact one of the
technical contacts listed below or the appropriate Office of Nuclear Reactor
Regulation (NRR) project manager.
signed by B.K. Grimes
Thomas T. Martin, Director
Division of Reactor Program Management
Office of Nuclear Reactor Regulation
Technical contacts: Laura Dudes, NRR James Tatum, NRR
(301) 415-2831 (301) 415-2805
Email: lad@nrc.gov Email: jet1@nrc.gov
John Fair
(301) 415-2759
Email: jrf@nrc.gov
Lead Project Manager: Beth Wetzel, NRR
(301) 415-1355
Email: baw@nrc.gov
