ML17244A833
| ML17244A833 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 08/24/1979 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| To: | |
| Shared Package | |
| ML17244A831 | List: |
| References | |
| NUDOCS 7909210420 | |
| Download: ML17244A833 (25) | |
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0, UNITED STATES NUCLEAR REGULATORY COMMISSION
'ASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 29 TO PROVISIONAL OPERATING LICENSE NO.
DPR-18 ROCHESTER GAS AND ELECTRIC CORPORATION R.
E.
GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244 Introduction By letter dated December 18, 1972, the Atomic Energy CommissiqIt's* Regulatory staff requested Rochester Gas and Electric Corporation (RG8E)'(licensee) to submit a detailed evaluation to substantiate that the R.'D Ginna Nuclear Power Plant (Ginna) could withstand the effects of a postulated rupture of any high energy fluid piping outside the primary containment, including the double ended rupture of the largest line in the main steam and feedwater systems.
It was further requested that, if the results of the evaluation indicated changes to'he facility were necessary to assure safe plant shutdown, information on the design changes and plant modifications be provided.
Criteria for performing this evaluation were included in our December 18, 1972 letter.'RC and RG8E representatives met in Bethesda,
- Maryland, on February 1, July 18 and September 18, 1973, to discuss the NRC request and the scope of the expected RG&E analyses.
In response to our request, RG8E submitted a letter dated November 1, 1973, that included a sumary report "Effects of Postulated Pipe Breaks Outside the Containment Building" dated October 29, 1973.
The results of this pipe whip and building pressurization analysis indicated that the intermediate building structure at Ginna was generally incapable of resisting pipe whip and pressuriza-tion effects of most postulated main steam and feedwater breaks within this building and from the adjacent turbine room.
The licensee determined that modification of the structure
~r pipe encapsulation to provide the required protection was not practical<3 and an extensive volumetric examination program**
to provide added assurance that the postulated piping system breaks yquld not occur was later proposed(4),
initiated in 1973 and finally approved(o>
by NRC in 1975.
- Currently known as the Nuclear Regulatory Commission (NRC).
- In accordance with the requirements of 10 CFR Part 50, Section 50.55a, paragraph (g),
RG8E submitted by letter dated 7/2/79, the "Ginna Station In-Service Program for the 1980 through 1989 Interval".
Additional information was subnitted by RGSE's letter dated May 24, 1974.
This information was responsive to NRC concerns for postulated high energy line breaks outside containment and potential effects on safety related equipment that might
$ q required to cool the core.
The licensee later submitted a schedule~'~ for analysis and plant modifications.
As a result of the High Energy Liny )~eak Outside of Containment evaluation, plant changes have been made's summarized below:
An augmented In-Service-Inspection Program has been initiated to (4) further reduce the probability of a main feedwater or steam line rupture.
A Standby Auxiliary Feedwater System.has been added to further improve=steam generator feedwater reliability-and specifically to substitute for the auxiliary feedwater in the low probability that auxiliary feedwater pumps are damaged due to near by high energy pipe breaks within the intermediate building.
Check valves have been added to existing auxiliary feedwater lines near the connections
.to the main feedwater lines to minimize the auxiliary feedwater piping that is pressurized during normal operation.
Two parallel remotely operated valves have been added to a crossover line between the motor driven pump discharges to provide additional auxiliary feedwater makeup capability.
A large metal plate jet shield has been installed underneath the main steam header in the Intermediate Building to protect the service water piping from a postulated crack in the main steam line.
Jet Impingement Shields have been added to protect vital equipment including containment isolation valves, motor generators, transfer switches, cable trays, terminal boxes and wiring, pressure transmitters and reactor trip breakers.
Also jet shields have been added to protect main steam bypass valves and piping and other locations listed by RG8E.
Instrument cabling has been relocated to areas that will not be affected by postulated high energy pipe breaks; The heating and ventilation system has been modified to withstand postulated high energy pipe breaks without further endangering the capability to safely shut down the plant.
The east end of the cable tray that connects the Intermediate Building and the Relay Room of the Control Building has been sealed to prevent damage that could result from a postulated high energy line break.
Openings around pipes and cable trays that pass through the areas required for safe shutdown of the plant have been sealed to prevent steam leakage into these areas in the unlikely event of steam or feedwater line breaks in the Turbine Building.
Steam generator blowdown lines have been rerouted through the sub-basement to minimize the potentially detrimental effects of breaks in these lines within the Intermediate Building.
Sufficient floor grating has been installed at manholes to guard against flooding of safety related equipment in the Intermediate Building resulting from an assumed feedwater line break.
Steam line pressure and feedwater flow transmitters have been relocated away from the locations that could be affected by postulated high energy line breaks.
Pressure shielding steel diaphragm walls are being, installed at selected locations in the Turbine Building to assure continued operability of safety related equipment following a postulated high energy pipe break in the Turbine Building.
RG&E coranitted, by letter dated June 27, 1979, to provide jet shielding for one atmospheric steam dump valve, all steam generator code safeties, and the two main steam bypass valves and their associated 3-inch piping.
This shielding would be provided in conjunction with the Systematic Evaluation Program (SEP).
Furthermore, modifications to the Intermediate Building wall resulting from analysis of high energy line breaks in the Turbine Building will be made as necessary upon completion of the SEP.
Discussion Ginna is a pressurized water reactor that utilizes a reinforced concrete containment which contains the entire primary coolant
- system, including the steam generators.
The criteria and requirements used by the licensee and the staff for evaluy]jng the high energy line breask outside containment are summarized as follows:'quipment and structures necessary to shut down the reactor and maintain it in a safe shutdown condition, assuming a concurrent and unrelated single active failure of essential equipment, should
be protected from all effects of ruptures in pipes carrying high energy fluid, up to and including a double-ended rupture of such
- pipes, where the service temperature and service pressure conditions of the fluid exceed 200'F and 275 psig; Breaks should be assumed to occur in those locations specified in the "pipe whipe criteria".
The rupture effects to be considered include pipe whip, structural (including the effects of jet impingement),
and environmental.
2.
In addition, equipment and structures necessary to shut down the reactor and maintain it in a safe shutdown condition, assuming a
, concurrent and unrelated single active failure of essential equip-
- ment, should be protected from the environmental and structural effects (including the effects of. jet impingement) resulting from a single open crack at the most adverse location in pipes normally carrying high energy fluid routed in the vicinity of this equipment.
The postulated size of the cracks was either 1/2 the pipe diameter in length and 1/2 the wall thickness in width (critical crack size) or equivalent pipe flow cross section in area.
The licensee evaluated all piping outside containment that contains high energy fluid and is in the same building with or in the proximity of safety related equipment required for safe shutdown.
These lines are:
Hain Steam Feedwater Auxiliary Feedwater Steam Supply to Auxiliary Feedwater Pump Turbine Steam Generator Blowdown Charging Line Plant Steam The licensee's evaluation postulated longitudinal and circumferential breaks at high stress locations specified by the NRC criteria for piping break locations and considered the effects of pipe whip, jet impingement, pressurization, environment and flooding.
For the evaluation of piping cracks, effects of pipe whip and pressurization were not applicable.
The licensee described the course of events following various size breaks of the main steam and feedwater lines at different reactor operating conditions.
The equipment necessary to bring the plant to a safe shutdown was listed.
The licensee's analyses indicate that the Intermediate Building, through which the main steam and feedwater lines pass from the Containment Building to the Turbine Building, cannot withstand most main steam line and feedwater line breaks.
This results from the pressurization of the building following the postulated high energy pipe break exceeding the design pressure for the concrete block walls and the roof, and from the structural capabilities not being sufficient to withstand the effects of pipe whip.
Some equipment that is used to maintain the reactor in a safe shutdown condition is located in this building and might be rendered inoperable.
This equipment includes instrument channel
- cables, service water piping, and the auyj'diary feedwater system.
A number of alternatives to the final plan) modifications'ere evaluated and considered by the licensee to be impractical
Evaluation The Au mented In-Service Ins ection Pro ram proposed and implemented'y the licensee consists of radiographic examination of all welds at the design basis break locations in the main steam and feedwater lines and at other locations where a failure would result in unacceptable consequences.
The examination techniques, procedures, and inspection intervals are based on the requirements of Class 2 components of Section XI of the ASME Boiler and Pressure Vessel Code.
The program* is based on ten year inspection intervals with the first interval running from 1973,to 1982.
The extensive in-service inspection program is designed to preclude design bases or consequential main steam or feedwater pipe breaks.
During each third of the first inspection interval, the program provides for examination of all welds at specified design basis break locations and one-third of all the welds at specified locations where a weld failure could result in unacceptable consequences.
During each one-third of the succeeding 10-year intervals, the program provides for examination of one-third of the welds at design basis break locations but continues unchanged with one-third of the welds at locations where a weld failure could result in unacceptable con-sequences.
This program is designed to detect flaws capable of causing pipe failure.
The frequency of reinspections is designed to detect any change in condition in advance of a potential failure.
We have concluded that this augmented inspection program is a prudent measure to ensure a very low probability of any break in the main steam and feedwater lines.
The inspection.requirements fp this program have been incorporated into the Technical Specifications< '.
The Instrumentation Channels that initiate the protective action in the event of a main steam ine or feedwater line break are:
Pressurizer
- Pressure, Steam Line Pressure, Steam Line Flow, Feedwater Flow, Pressurizer Water Level, and Steam Generator Water Level.
The pressurizer
- pressure, steam line flow, pressurizer water level and steam generator water level transmitters are located inside containment
- and, therefore, their operability would not be affected by a high energy line break outside the containment.
Some of the signal cables from these transmitters,
- however, are routed through cable trays in the Intermediate Building.
To ensure that the minimum number of these channels required to produce the protective actions (safety injection, reactor trip, and feedwater and steam line isolation) are not adversely affected by a high energy line break in the Intermediate Building, their signals have been rerouted out of other containment penetrations and do not pass through the Intermediate Building.
The Steam Line Pressure and Feedwater Flow Transmitter Si nal Cables have been relocated to areas with no high energy lines.
The sensing lines for the transmitters are susceptible to damage since they connect to high energy lines.
However if they rupture, the channels fail downscale and since low steam line pressure and low feedwater flow produce the trips for protective action, the channels fail in the safe direction.
In addition, the signal cables for a cold leg reactor coolant temperature channel from each loop have been rerouted outside the Intermediate Building to provide the operators with additional information to follow the course of the accident.
The following instrument channels are isolated from the effects of high energy line break outside containment.
Instrument No. of Protected Channels Ho. Required to Tri Steam Generator Level 2 per loop 2 per loop 1 per loop with Steam Flow-Feed Flow Mismatch Steam Line Flow Feedwater Flow Steam Line Pressurizer Pressurizer Pressure Pressurizer Level Reactor Coolant Temperature 2 per loop 2 per loop 3 per loop 2 per loop 1 per loop 2 per loop 1 per loop 1 per loop 2 per loop 1 per loop 1 per loop The instrument channels or signal cables that remain in the unprotected areas of the Intermediate Building are likely to perform their trip function by providing protective action signals for the steam or feedwater line breaks either in the normal fashion or by the fail-safe trip.
This is because any failure which could occur auld most likely be a separation of the sensing line or signal cable and, except for the steam flow channels, loss of signal trips the channel.
Also the required protective actions can be initiated by the response of a single one of the parameters monitored by the channels
- above, such as low steam pressure on two channels in one loop, or by a number of diverse responses, such as low pressurizer pressure and level on one channel.
Therefore, the protected channels and those remaining in the unprotected area maintain the required diversity and redundancy for reactor protection systems.
In addition, the protected channels will ensure that the operator is provided with information for'the course of the accident.
On this basis, we find these modifi-cations acceptable.
The Auxiliary Feedwater System is also located in the Intermediate Building with all three pumps in the same vicinity.
There are two motor driven pumps and one steam driven pump.
These pumps are only used during start-up and normal or emergency shutdown of the plant.
The pumps are susceptible to damage from the effects of breaks in the main steam and feedwater lines and the auxiliary steam and feedwater lines.
To ensure the heat removal capability for core cooling, the licensee proposed and later installed a Standby Auxiliary Feedwater System adjacent to the Auxiliary Building along the south wall.
The Standby Auxiliary Fee]g~ter Pumphouse is a seismic Class 1 concrete structure supported by caissons The Standb Auxiliar Feedwater S stem consists of two, independent 100 percent capacity subsystems in a new structure remote from high energy lines.
The discharge piping from the pumps was routed through the Auxiliary Building, enters the containment through penetrations remote from the main steam and feedwater lines, and connects to the feedwater lines near each steam generator with check valves near the connection to minimize the amount of line pressurized during normal plant operation.
The pumps take suction from the Service Mater Loops inside the Auxiliary Building are motor driven from the Engineered Safety Features
- busses, and are manually started from the control room in the event that the Auxiliary Feedwater
- pumps, which start automatically, are not operable.
The analysis performed by the licensee assumes that feedwater is not available for 10 minutes following the worst case line break.
This is ample time for the control room operator to take action since alarms and indications are available in the control room to alert the operator to the lack of effective auxiliary feedwater flow and the standby pumps can be put into operation from the control room.
Our concerns for the structural,.mechanjq) and material aspects of the modifications were adequately addressed~ 'y the RG8E letter dated July 28,
- 1978, in response to our request dated June 21, 1978.
In the event of loss of off-site power, the pumps would be powered by the diesel generators.
The diesel generator s have sufficient capacity for this additional 225 Kw load.
However, to prevent an overload of the feedbreakers tying the diesels to the buses, an interlock has been installed to prevent starting a standby pump when its associated auxiliary pump is running on the diesel.
The Standby Auxiliary Feedwater Pump Building and System design satisfied ~9 ]
)
the codes and standards applicable in 1974 when the building was designed Me conclude that these modifications provide an acceptable backup to the Auxiliary Feedwater System for maintaining the plant in a safe shutdown condition.
The scope of the Safety Evaluation of the Standby Auxiliary Feedwater System is presented in the enclosed Appendix 1.
On the baqjq of this evaluation, the Technical Specification changes proposed by RG8E< ', which we revised with RGSE
,'oncurrence, are acceptable.
Also, the same operating procedure requirements for the prevention of water hammer in the Auxiliary Feedwater System should be applied to the Standby Auxilfary Feedwater System.
l
The Ventilation S stems were evaluated to determine whether the steam from high energy line breaks would intrude into an area where personnel or equipment important to safety would be endangered.
It was determined that modifications were necessary to the control room lavatory exhaust, the control building ventilation equipment room relief opening, the relay room cable tray openings and tunnel, the battery room exhaust and cable tray openings, the diesel generator room piping and cable tray openings, and some interconnecting ventilation ducts between the Intermediate Building and the Auxiliary Building.
All of these openings have been sealed and the exhausts have been ducted to areas not subject to intrusion of the steam from a high energy line break.
Based on the above, we conclude that these modifications reduce the probability of adverse con-sequences from the postuated high energy line breaks and are, therefore, acceptable.
Pressure Shieldi teel Dia hra m Walls were. proposed by RG8E's letter dated February 6, 1978
. " The steel diaphragm walls were to have been erected between the Control Building and the Turbine Building and between the Diesel Generator Rooms and the Turbine Building.
The walls would:
Comply with the requirements for physical protection of licensed activities against industrial sabotage (10 CFR Section 73.55)
Provide protection from postulated fires on the operating level of the Turbine Building Provide protection from postulated high energy line breaks in the Turbine Building We met with representatives of the licensee in Bethesda,
- Maryland, oo February 15, 1978, to discuss fire protection and structural aspects of the diaphragm wall and on January 30, 1979, to discuss structural design criteria fear ggq wall.
On the basjq qf information provided by the licensee('
'"'e have concluded<'
that the steel wall and door that have recently been added between the Control Room and the Turbine Building are designed for high power rifle resistance (level IV bullet resistance) and, therefore, meet the requirements of 10 CFR 73.55.
We have reviewed the adequacy of the Steel Diaphragm walls between the Control Building and the Turbine Building and between the Diesel Generator Rooms and the Turbine Building witgl~es~~ctl)o fj~e protection.
Base)
~q the information provided by the licensee<
we have concluded
< 'hat the concept of a steel diaphragm wall between the Turbine Building and the Control Room protected by an automatically actuated water curtain is acceptable, but the details of the water supply and actuation system must be submitted for our review.
Concerning the Pressure Shielding Steel Diaphragm Turbine Building walls adjagyng to the Diesel Generator, Relay and Battery Rooms, the licensee has agreed<'6~
to conduct studies to determine what active and passive systems should be installed to prevent structural failure from fire tha$
>~quid jeopardize safe shutdown of the plant.
We have also identified< 'he requirements for fire doors in the areas where the steel diaphragm wall is being constructed.
f
The NRC Safety Evaluation of the structural adequacy of the Pressure Shielding Steel Diaphragm Walls is presented in the enclosed Appends~12.12Weljwav~4~oncluded on the basis of information presented in ]j~qnsee letters and during a meeting with NRC representatives',
that the structural criteria and design methods for the steel diaphragm walls are adequate to assure safe shut down of the reactor following a high energy pipe break in the Turbine Building.
However, our conclusion is based on the premise that the peak Turbine Building pressure and temperatures that the Turbine Building steel diaphragm walls adjacent to the Diesel Generator Rooms and the Control. Building
(Control Room, Relay Room and Battery Room) must withstand, results from a postulatp) rupture of the 20" Feedwater Line.
Since the licensee had previously reported<
that the pressure on the operating level of the Turbine Building as a result of a break in the 24" or 36" steam line peaked at 0.098 psig with steam relief through the building ~~)aust fans in the wall and roof, and later reported 0.70 psig pressure peaks, resulting from a break in the 20" main feedwytp line, we requested RGSE to submit additional analysis.
The licensee's basis' for using the Main Feedwater 20" pipe break to determine peak Turbine Building transient pressure and temperature for the structural design of the new steel diaphragm walls was justified because of the augmented In-Serv'ice Inspection of all welds in the steam lines in the Turbine Building and the resultant low probability of a large break in the steam lines.
Nevertheless, at our request, by letters dated Mqy 1
1979 and July 6, 1979, the licensee provided supplementary infonaation<'
~ which in addition to the Turbine Building pressure transient analyses for postulated feedwater pipe breaks, also included steam line breaks in the Turbine Building.
As expected, these calculations showed that the steog~line break pressure transients were significantly greater than originally reported<
The following additional information provided by RG8E:
The peak pressure transients in the Turbine Building calculated by the licensee are less than the 0.7 psig structural design pressure for the steel diaphragm wall on the mezanine floor along the control room wall and less than the 1.14 psig structural design pressure for the steel diaphragm walls on the operating floors at the relay, battery and diesel generator room walls.
The new steel diaphragm walls are at nearly opposite ends of the Turbine Building from the high energy piping thereby providing adequate separation to preclude wall damage at these locations because of pipe whip or jet impingement that could accompany a high energy pipe break in the Turbine Building.
Based on this information and our detailed Safety Evaluation of the pressure shielding steel diaphragm walls in the Turbine Building which is included as Appendix 2 to this Safety Evaluation, we have concluded that thal~tructural adequacy of the steel diaphragm. walls as described by the licensee~
is acceptable.
A schedule for completion of the installation of the steel diaphragm walls, in a:cordance with the information provided, should be submitted within 60 days.
Environmental Consideration We have determined that the amendment does not authorize a change in effluent types or total amounts nor an-increase in power level and will not result in any significant environmental impact.
Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental
- impact, and pursuant to 10 CFR 551.5(d)(4), that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.
Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents, previously considered and does not involve.a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed
- manner, and (3 ) s'uch activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Enclosures:
1.
Appendix 1, "Detailed Evaluation of the Standby Auxiliary Feedwater System - R.
E. Ginna" 2.
Appendix 2, "Detailed Evaluation of the Pressure Shielding Steel Diaphragm in Turbine Building - R.
E. Ginna" Date:
August 24, 1979
REFERENCES:
NRC letter to RG&E, dated December 18, 1972, "Effects on Essential Auxiliary Systems of a Major Break of the Largest Steam or Feedwater Line".
2.
3.
4.
6.
7.
8.
9 10.
12.
RG&E letter to NRC dated November 1,'973, "An Analysis of the Effects of Postulated Pipe Breaks Outside Containment".
RG&E letter to NRC dated September 4, 1974, "High Energy Line Break Outside Containment - Alternate Possibilities for Protection from.
High Energy Line'reaks".
RG&E letter to NRC dated October 31,
- 1974, "Proposed In-Service Inspection Program for High Energy Piping Outside Containment".
NRC Amendment No.
7 to Provisional Operating License No.
DPR-18 for the R.
E. Ginna Nuclear Power Plant, dated May 14, 1975.
RG&E letter to NRC dated May 24, 1974 - RG&E response to NRC staff request for additional information (April 24, 1974) related to "Effects of Postulated Pipe Breaks Outside Containment".
RG&E letter to NRC dated November 1, 1974, "Updated Schedule for Analysis and Plant Modifications".
RG&E letter to NRC dated February 1, 1977, the third of three RG&E requests "Te'chnical Specification Changes to Include Requirements for the Standby Auxiliary Feedwater System" where it is declared operational.
RG&E letter to NRC dated May 20,
- 1977, "Standby Auxiliary Feedwater System Design Criteria".
RG&E letter to NRC dated July 28, 1978, "Additional Information Related to Strcutural, Mechanical and Material Aspects of the Standby Auxiliary Feedwater System Modification Responsive to NRC Request dated June 21, 1978".
RG&E letter to NRC dated February 6, 1978, "Pressure Shielding Steel Diaphragm in Turbine Building R.
E. Ginna Nuclear Power Plant, Unit No. 1, Docket No. 50-244 - Design Criteria, Rev.
2 July 18, 1977".
RG&E letter to NRC dated August 25, 1978, "Request for NRC Approval of the Proposed Turbine Building Modifications in the Vicinity of the Control Building and the Diesel Generator Annex.
REFERENCES:
13.
RG&E letter to NRC dated.October ll, 1978, "Responses to NRC Letter dated September 13, 1978, Concerning the Pressure Shielding Steel Diaphragms.
14.
RG&E letter to NRC dated October 18,
- 1978, "Request for NRC Approval of the Fire Protection Aspects of the Control Room and Relay Room Doors that Provide Pressurization and Fire Protection.
15.
NRC Amendment No.
25 to Provisional Operating License No.
DPR-18 for the R.
E. Ginna Nuclear Power Plant dated February 22, 1979 - Physical Security Plan.
16.
NRC Amendment No.
24 to Provisional Operating License No.
DPR-18 for the R.
E. Ginna Nuclear Power Plant dated February 14, 1979 - Fire Protection.
17.
NRC "Su@nary of Meeting Held on January 30, 1979" dated February 9, 1979.
18.
RG&E letter to NRC dated May 17, 1978, "Calculated Pressure Transients in Turbine Building" responsive to NRC commitment (Ref. 17).
19.
RG&E letter to NRC dated June 27,-1979, "Plant Modifications Resulting from High Energy Line Break Outside Containment Evaluation".
20.
RG&E letter to NRC dated July 6,
- 1979, "Pressure Transients in the Turbine Building - Pressure Shielding Steel Diaphragms".
APPENDIX 1
DETAILED EVALUATION OF THE STANDBY AUXILIARYFEEDWATER SYSTEM R.
E.
GINNA
~Back round Rochester Gas
& Electric'RG8E, the licensee) in its report "Effects of Postulated Pipe Breaks Outside the Containment Building for the Robert E.
Ginna Nuclear Power Plant Unit No. 1", dated October 29, 1973 (Reference 1), determined that the Auxiliary Feedwater System (AFS) in the intermediate building could be damaged by a high energy line break (HELB) in that build-ing.
Therefore, RG8E has installed a standby Auxiliary Feedwater System (SAFS),
independent of and remotely located from the AFS.
The SAFS, described in References 1 through 4, was installed solely as a backup to the AFS in the event of a HELB and is not intended to be used to mitigate the effects of other plant accidents or transients.
The SAFS was designed and constructred for the purpose of providing feedwater to the steam generator in the event that a large main steam or feedwater line break in the Intermediate Building were to disable both trains of the AFS.
The licensee has designed the SAFS as a safety related system and the system and its enclsoing structure were built to the most current criteria at the time.
Under the Systematic Evaluation Program, all safety related equipment and structures at Ginna are being re-evaluated (Reference 5).
Because of this ongoing re-evaluation, the SAFS and its enclosing structure were examined (1) to determine if SAFs operation would acceptably mitigate the consequences of AFS damage which could result from a postul'ated HELB, (2) to determine that the addition of the SAFS and its enclosing structure would not adversely impact other, previously approved, safety related
- systems, structures or components, and (3) to determine if the. technical specification requirements for SAFS operability and surveillance were acceptable.
Descri tion and Evaluation of Standb Auxiliar Feedwater S stem The SAFS provides two independent feedwater flow paths from separate seismic Class I Service Mater System loops, via two motor driven pumps to the Ginna plant's two steam generators.
Each SAFS pump can, within a few minutes of reactor trip, provide the required flow for removing reactor core decay heat.
In addition, a comparison of the SAFS with the AFS, which has almost ten years of operating experience, shows that each of the SAFS pumps has sufficient capacity to cool the plant adequately.
Each pump supplies flow through its normally open discharge
- valve, a containment isolation stop check valve, and a check valve to the main feed line for its respective steam generator.
Also, two motor operated cross-connect valves in parallel can be used to direct flow from one pump to either steam generator.
Two manually operated valves in series can be opened to cross-connect SAFS pump suction lines.
The system layout and capacity provide adequate re-dundancy to accommodate a single active component failure without loss of system function.
A condensate storage tank-i s'used to store 10,000 gal lons of condensate quality water for periodic SAFS testing.
The tank is located-in the Auxiliary Building Addition.
Piping is provided to supply 125 psig condensate
.o the SAFS to pressurize the pump suction lines arh to fill the conden" ate storage tank.
The SAFS piping is of ASllE Code, Section
". II, Saf'ety Class 2 and 3
design exceot fo'.
t;:.'o interfaces wi th non-nuclear safety class piping.
Tl!ese interfaces are used for systc-".
pressu;ization fry~ tihe condensate system ar0 for pump flow recirculation testing, r!hich are no;!-essential system functions.
These nor-essential port'ions of the system can be isolated to pemit continued system function in th event of their failure.
The SAFS instrum.ntation is capable of de.ecting sigvi=icant le-kag from the system, and system leakage is directed to
";l!e waste holdup tank in the Auxiliary Building bascm nt.
Tiie "A"S system satisfies the requirements for quality group and seis!!ic si ication Qs ide!ltified in R'gulatory Cuides 1.26 and
'>.29 ~
Tl o s'i.".,ic classi'i'icat.on is based on the defini.ions provided in the "ac'l i "<'s Final Oesign and Safety Analysis Report (Reference 5).
The
~A,"S
.".as designed for installation wi th no degradation in the i'."si
~.
or function of existing systems and, for operation within code al 1 o:,;:ble stress s over the full range of expected op rating tempera-U~res,
>.'el ding pvo edur es used to fabricate the system were in a cordavce wi th Section III of the AR'lE Code.
Pre-opevational hydvo-st~I,"ic testing or" the AB]E III Class 2 and 3 portions o, the system was in c".ccord'nce Mi th tne A31E Code.
The SAFS di scharge lines were vou"ed ti1!"'i".'h th<<primary containmpnt Do!!ndary via cxi sting spare penetrations, Local leak testing of these penetrations is conducted in accovdance wiih !0 CFR 50, Appendix J.
Viaterial selection for the Class 2 and 3
c
",',pon nts of i;i!e SAFS was i!! accordance v i th Section
!II of the A"lE Co;c.,
The 1>c.:nsoe ha" not evaluated t'!e SAFS i 'sel f wi th regard to the <<if'cts of pi';"=
h-;p a!!d jet iiipingem<<nt except for tl'.e piping inside contairaer.";
bc "'een 'he mair' edwatev 1'! ne ard th'losest check valve in the SAt S in,i;.".".ion lines.
This is in accot'dance wi tl'l the provisions of cur vent rlRC ci'iteria (Reference
?) for systems which do not operate during r,or",::: i plant con!i'ions.
The section of S.FS piping inside contairment whic') is a h'igh energy line du! ing normal plant conditions has been eval ua ted by the l.icensee and found acceptab'ie fvom the standpoint 0!".
I ~
I hLL3s because oi i ts
'. oca.io>>
a",ay from safe shutdown equipment..
Tl!e SAFS i ts>'I f is protected from the effects of )lELSs outside contain;.!ent bec,us'her" ar'.
no high <<nevgy lines in the vicinity of SAFS cQ!Ipone'lt.'.
I J
!'.".e desi!.": of the S"."S does not pr cclude fe"d system watevhamm;:r.
The occur! o>>ce of w-'..rrham:';ev is cur; ently prevented by AFS op rating pro-cedures
~
similar pi ocedures (requiring an u!pper
! imit on feedwatev addition rate whenever steam generator level is below the feed ring) would be used for the SAFS.
The licensee has completed (February 1979 outage) steam generator modifications to further limit. the potential for occurrence of feed system waterhamner (Reference 9), in accordance with current requirements for operating reactor plants.
The electrical power for each SAFS train is suppli ed by one of the two redundant 480 VAC emergency power systems at Gi nna.
SAFS pu'mps C and 0
are po>>ered frcm existing spare feeders on emergency busses 14 and 16 respectively.
These busses can receive power from offsite ar onsite (diesel) sources.
-To prevent exceeding electrical load limits on the feeder breakers tying the diesel generators to the buses electrical interlocks are provided to prevent an AFS and a
SAFS pump from being connected to the same diesel generator simultaneously.
The require-ment for power supply diversity (Reference
- 8) is not applicable to the SAFS because the requirement is based on an assumed loss of all AC power (offsite and onsite).
If this is assumed,,the SAFS is rendered inoperable, and the turbine driven pump of the AFS provides the required power supply diversity.
The staff considers the simultaneous occurrence of a HELB which renders the AFS turbine driven pump inoperable and the total loss of both onsite and offsite AC power to be sufficiently unlikely as not to be credible.
The SAFS is manually started frcm the control room or frca a local station in the Auxiliary Buildinq Addition pump room.
A switch for transfer of control from ',ocal to the cont. ol roc. is provided at th local control station.
Control rom indication shows the status of thi s trans fer switch.
A TEST/NGRliAL mode sw. icl, is also provided local to the pumps.
The switch will be in the t!OR!NL posiii'on at all times exce-t
>~hen system operational tests are be',nq con<lucted.
In the TEST mode, interlocks will prevent pump operation unle s ihe corresponding manual suction valve in the tank ou.lei lire is in the fully open position, and will trip the pumps>>hen
.ark LO level is reached.
In the NOR!1AL mode, interlocks wi!1 prevent pu;;p operation unless the motor operated valve in ihe correspondirg service water lin-:
iss
- open, and will also prevent two auxiliary
$eedwater!)umps fron b in",
connected tc the same diesel genera".or simuli;aneously.
Control room instrumentation will alert the operator thai the AFS is ineffective and the SAFS should be started.
4-Pump discharge flow and pressure indication is provided locally and in the main control roon.
1n the event of damage to a steam generator or associated piping the operator will use this flow indication.og.t<er with steam generator level
<nd steam pressure to determine w!erich steam genera"or system is damaged and isolate the ieedwatcr flow to t>>at stea".
gene! a".or, using the motor operated valves if necessary.
There is no provision for either the AFS or the SAFS to automatically terminate flo to a depressurized steam generator and automatically provid~ ilaw to the iritact stem generato-.
This is acccnplisi:ed by
~he control room oper';or.
The effect of the lack of autcnatic switchirg o, flow to the intact steam generator will be assessed in the SEP main steam lin l>reak evalua.ion for Ginna.
Thc inst.umentation and controls for the SAFS conform to Gereral Oesic-.~
Criteria (GDC) -19, "Control Room".
The SAFS is a manually in.itiat d system intended to be used in the even'. that a postulated i1ELB i" the Intemediate Building were to disable the AFS.
An analy~is of a worst case feedwater line break using conservative assumptions was nrovided by th'icensee in Reference 2.
This analysis has shown that a 10 minu:e delay in the ini tltition of auxiliar<< feed flov from one motor driven pump ensue t s i n acceptable conseq<<ences.
The 1 os ". o feed accident anal ysii s pr esented i n Chapter 14 of Reference 6 assn<!es auto:,!atic in1 t a-tion of oi>e motor driven auxiliary feed pu'.<;p one,minute after acciident initiation and results in no )oss of decay heat rer.oval capability of the steam g nerator receiving auxiliary feed and no loss of coolant from the primary system pressur.zer relief valves.
The latter analysis covers a spectrura of loss of feed events from those of hlgil ">robablllty of occurrencee to low probai>ility events including pump failures, val'.i=
!ideal functions, loss of o. Fsite power and pipe br'a!;s; anted, fo; th: se, the auto'.Iatical 1 v i ni tiated AFS prov'os adequate nrctcction.
The S,"FS has been installed to pl otect against the low probability event of postulated main f cd or steam 1
ne break in the intermediate buil "ir<g
- that cc-:<pletely disables the AFS.
L(e have reviewed the sequence of actions that the control roc'perator must take to ir<itiate the SAFS and concluded that tie oper< tor woulC h<"ve sufficient time fo~ manual initiation of
".he systemi wi thin the conscr!atively eel cula"'"d 10 minu<te period.
Thereforo, manual initiation of the SAFS is <iccepta>'la.
Since the acceptance criteria for fcedwater line break analyses ilave changed sirce the 1974 analvsis submitted t>y the licensee n suppo.t of the 10 minute period for SAFS initiation, we na!e also reevaluated the cors-quences oi thiis accident assuming core d<.mage.
Using the ass(r!ptions of instantaneous release of Ri; oi the iodines and noble gases in t.e core, a reactor coolant volume of 5750 cubic feet, primary to secondal j lea!-:
rate at the maximu!n technical specif1cation limit, and a
elativ concentration X/9 du. to atmos';etheric diffusion of 1
x 10
/
v>e 1
calculate r,he resulting doses at the nearest site boundary to be sec/m within 10 CFR Part 100 criteria for offsite radiation doses.
Even though two actions are required to initiate the SAFS (open suction valve, star pump), the system confor1"..s to Regulatory Guide 1.62.
This is because systemi initiation depends on operat',on of a reasonabl~
q 'l
. t consideri:1g the desire to avoid introduction into the stea'
.nerator of chemica!
i!~purities from the Service liater System through a sinole operator error or a single electrical malfunction.
Pioxil iarv Building Addition Tne SAFS pumips are located in tile Auxiliary Building Add t'>j
)e south s1de of the Auxil iary Building.
That portion of the Auxtliary 'Building Addi tion which encloses the pumps, the;)r>r!!phou is a reinforced concrete structure designed to meet thr
~ei..
Cl. ~
I cr1 tel' or !,efererce 6 and to protect the essential portions of the SAFS fco no ef fec s of torrlados, 1!icludi ng tom<,-io mi ssil es a il d
I r i
> a
'la err<.'ro,.a~nial con i <ions.
Air cooling ainrl ieat1ng n
~
'i:-
g un1'i~s 1n
< ne p><<'i','>
i)ouse
<".ro designed
'to keep he roo"l tern>)e ature su1 'bl~ f ccess
<o tne pumphouse is via a st el fc<'me temnorM'l e
Tl,e Auxiliar; Building Addition and temporary storage structure v>ere eviewed to assure hat their design or constri.'ction>>1>o ld not hav a',1 adverse im!)act on previously approved safety related stru tures or systems.
Based on our'eview,
'r>e have deter".)it)ed tha+ t<": 't 11- ':
hese s.r<<ctures imposes no adverse impact on 'exi stirg
~ f t.
rela,ed s;ruciures or ystems.
In fact, tl.e pumphouse was built to more recent seismic design cri eria than those identified ir, Reference G.
Techni cal S~eci ficati ons The licensee has proposed operability and surveillance technical speci-fication requirements for the SAFS (Reference 10).
The staff reviewed need to:
the proposed specifications and concluded that some modificat o i ica ions were l.
Better define the plant conditions when the SAFS (and tl;e AFS) are required to be oper'ab>e.
2.
Provide acceptable tin!e periods to rep<,ir an inoperable syste<<)
Ori'i <<
in ~
3.
Assure that the redundancy of the SAFS (and AFS) is maintained when feedwater is being obtained from the Service Hater System.
Based on our review, we conclude that the proposed technical specifications, as modified, are acceptable COVCLUSION Based on our revi ew of the information provided by the licensee, we conclude that
( 1) the SAFS is in conformance wi th the Commission's regulations as set forth in General Design Criteria (GDC) 19, "Control Roan",
GDC 44, "Cooling Water",
GDC 45, "Inspection of Cooling Hater
. Syste!'.!s",
GDC 46, "Testing of Cooling llater Systems",
GDC 54, "P'ipinig Systems Penetrating Contairment",
GDC,57, "Closed System Isolation Valves",
and (2) the SAFS meets the guidelines of Regulatorv Guides 1.26, "guality Group Classification",
and 1.62, "l(anual Initiation oF Protective Actions".
In addi tion; although several aspects of the SAFS and the Auxiliary Building Addition are being re-evaluated in the Systematic Evaluation
- Program, we have concluded that (1) the SAFS would acceptably mitigate the accident for which it was designed, (2) the installation of the SAF."-
arid the Auxili rv Building Addition does not reduce existing safety margins for other safety related structures, systems and components, and (3) the proposed technical specifications for the SAFS, "s modified by the staff, are acceptable.
Therefore, the SAFS should Le placed in an operable status to provide the addi tional plant protection foi. which i t was designed.
Date:
August 24, 1979 I.
Report entitled "Effects of Pos';~!';a"ed P
-.;: "reeks Outside the Containment Buidlinn, Robert E. Gin;>a,'u:".ear PoÃer Plant Unit 1",
dated October 29, 1973.
2.
RG5E letter, K. Amish to J. O'Lea!", da'.'-.'".
i'i:.:J 24,
'1974.
3.
'RG8E letter, L. White to A. Sch!:e;;c. r, d. e.,ay 20,
- 1977, for%>>ar: i,",g reports "De"=ign Crit ria Standby h.;>:i <--,-y F"-:d Syst'-a", '"Pr eliminary System Desigr. Description or thu Stand/". '.:;.'liary Feed<!aber Svsti.',!",
and "Desirn.Criteria for tl'e Addition
> ':;;!!.'!xiliaryBu"',lding".
4.
RGGE letter, L. White to D. Zio;;-:.':.:"
da'"":!;!:.,y 28, 1978.
5.
Systematic Evalulation Progra>>,
Stains S".!::ary Re.ort, tiUREG-0458, September 22, 1978.
6.
Robert E, Ginna Huclea!
Power Plant Unit!~, 1, Final Facility Description and Safety Analysis.",:.";o't.
7.
Branch Technical Position ASO 3-1 appen.-.;:d
.l.;, 3 andard Revio;>> Plan 3.6.1.
8.
Branch Technical Po. ition ASB 10-"~:.p~=-.~l'.d l;o Standard Revie>> Pla>
10.4.9.
9.
RGEE letter, L. White to D. Zie.'.ann, dated J."ne 15, 1978.
10.
Letter, LeBoeuf, Lamb, Loiby c~ ii,!.'Rae to B. Rusche, dated Feb rua ry 1, 1977.
APPENDIX 2 DETAILED EYALUATION OF THE PRESSURE SHIELDING STEEL DIAPHRAGMS IN TURBINE BUILDING R.
E.
GINNA SCOPE The scope of this evaluation involves (a) assessment of the adequacy of the postulated design basis, (b) review of the ability of the "Pressure Shielding Steel Diaphragm" DESCRIPTION OF STRUCTURE The proposed pressure shielding steel diaphragm walls are being installed between the control building and the turbine building (i.e., adjacent to the control room, relay room and battery room) and between the diesel generator annex and the turbine building at the R.
E. Ginna Nuclear Power Plant (Ginna).
The new structures consist of horizontal steel beams (connected between existing steel columns) and vertical corrugated steel panels.
The new steel beams provide support for the steel panel diaphragms.
A detailed description of the modification can be found in the design criteria and engineering drawings provided by the -Rochester Gas and Electric Corporation (the licensee)
(References 1 and 5).
SUMMARY
. OF DESIGN.,LOADS.:
Seismic Load References 1 and 2:
A.
Peak Ground Acceleration (a review of the definition of seismic input at the Ginna site currently is being conducted by the staff):
O.lg for Operating Basis Earthquake (OBE) 0.2g for Safe Shutdown Earthquake (SSE)
B.
Regulatory Guide 1.60 Design
Response
Spectra were used.
C.
Peak spectral acceleration used for design:
0.28g for OBE 0.55g for SSE D.
Horizontal and vertical seismic loads were applied simultaneously.
Damping ratio:
Equivalent static approach including "1.5" safety factor was used.
Pressure and tern erature loads due to i
e break (reference 4):
Pa
= 0.7 psi - Control Room
- a
= 1.14 psi - Diesel generator room, relay room and battery room The temperature load "Ta" was converted to equivalent pressure load and combined with "Pa".
The combinations of the applied loads used for design are based on Standard Review Plan 3.8.4 (References 1
and 3).
Jet impingement effects were excluded since the high energy lines are located at a large distance from the diaphragm walls.
EYALUATION A.
The criteria used in the analysis and design of the new steel diaphragm walls to withstand the postulated loading conditions are in accordance with NRC Regulatory Requirements (Standard Review Plan 3.8.4, Regulatory Guides 1.29, 1.60, l.61) and AISC, "Specification for the Design, Fabrication and Erection of Structural Steel for Bui'ldings".
B.
The postulated loading conditions (including dead
- loads, live loads, horizontal pressure
- loads, temperature load and seismic loads) and load combinations that may be imposed on the new walls during the service life-tine of the plant conform to NRC - Standard Review Plan Section 3.8.4.
C.
The equivalent static approach applied for the seismic analysis of structures is in accordance with HRC - Standard Review Plan Section 3.7.2, III-lb.
D.
On the basis of the information provided by the licensee, we have concluded that the structural design provides reasonable assurance that the new walls will withstand the specified desi.n conditions without impairment of structural integrity or performance of required safety functions and is therefore acceptable.
This sa ety evaluation is based on the promise that the applied pressure an='e.-perature loads on the steel diaphragm wall are caused by the pos tulated full diameter breaks in the 20" feedwater piping and in the 12" main steam pipe or postulated crack breaks in the 30" main steam line.
The licensee provided the basis for these loads in their reports "High Energy Line Break Inside the Turbine Building," dated Hay 17 and July 6, 1979.
Date:
August 24, 1979
References:
l.
Design Criteria for Pressure Shielding Steel Diaphragm in Turbine Suilding, Revision 2, July 18, 1977.
2.
3.
Letter to D. L. Ziemann, NRC, from L. D. Mhite, Jr.,
RGAE Corporation, August 25, 1978.
Letter to D. L. Ziemann, NRC from L. D. White, Jr.,
RGEE Corporation, October ll, 1978.
4.
Summary of lleeting held on 'anuary 30, 1979.
5.
Engineering Drawings - Turbine 8uilding Pressurization l,'a) ls and Control Room <<(alls dated October 14,
- 1977, Drawing lJo.
04 4594 D-581-020,
-021,
-022,
-023,
-025,
-026.