ML20032A733
| ML20032A733 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 10/26/1981 |
| From: | Withers B PORTLAND GENERAL ELECTRIC CO. |
| To: | Clark R Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0696, RTR-NUREG-696, TASK-3.A.1.2, TASK-TM TAC-43091, TAC-46131, TAC-46346, NUDOCS 8111020295 | |
| Download: ML20032A733 (8) | |
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9 Lu ;.6 Y October 26, 1981 Trojan Fuclear Plant Docket 50-344 License NPF-1 Director of Nuclear Reactor Regulation ATTN:
Mr. Robert A. Clark, Chief Operating Reactors Branch No. 3 Division of Licensing U. S. Nuclear Regulatory Commission Washington, DC 20555
Dear Mr. Clark:
Your letter of September 2,1981 indicated the need for additional information regarding habitability of the primary Emergency Operations Facility (EOF) located at the Trojan Nuclear Plant. This letter summarizes the information previously submitted to the NRC, provides additional justification to continue using the existing primary EOF and requests an exemption for shielding and ventilation systems for the primary EOF in accordance with Section 4.2 of NUREG-0696.
A description of the Emergency Response Facilities, including the EOF, was originally submitted to the NRC on October 21, 1980. The submittal included an analysis of the radiation doses to personnel within the primary EOF. This analysis was performed in accordance with Standard Review Plan 6.4, using meteorological dispersion coefficients determined in accordance with Regulatory Guide 1.145.
The analysis concluded that the primary EOF meets the radiological habitability criteria in Standard Review Plan 6.4 for the control room and General Design Criteria 19 in 10 CFR 50, Appendix A.
The concept of two EOFs (a primary EOF near the Plant and an alternate EOF at the PGE St. Helens office) was presented in that submittal for NRC review and approval. The PuE alternate EOF which consists of a permanent structure with a permanently installed communications capability comparable to the primary EOF substantially exceeds the requirements for an alternate EOF in NUREG-0696.
A meeting was held on June 15, 1981 between PGE and the NRC to discuss the October 21, 1980 letter. Following this meeting, PGE submitted 3
additional information describing advantages to establishing two EOFs, together with Protective Action Criteria for smooth transition between f
the two EOFs including anticipatory evacuation of the primary EOF.
jIl Based on the NRC letter of September 2, 1981, the practicability and cost-effectiveness of modifying the primari EOF to comply with the shielding and ventilation criteria in NUREG-0696 has been reevaluated.
The result of our evaluation indicates, as shown in Attachment 1, that upgrading the primary EOF is impractical from a structural t,tandpoint.
In addition, the meteorological data presented in our October 21, 1980 l 1 to.encompara 9 years _of 3111 o 2 o#Me ter has been updated in Attachment F m awS~ Sucemaee m a
Portland General ElectricCompa1y e
Mr. Robert A. Clark October 26, 1981 Page two wind-frequency data at the Trojan site.
This expanded data base con-firms the very low (0.5 percent) frequency of winds blowing toward the primary EOF (WSW Sector). Additional wind-frequency data by month is also provided in Attachment 1 to show that there is only a small month-to-month variation in the frequency of winds blowing toward the VIC.
Based on the additional information provided in Attachment I and the data already submitted to the NRC, the primary EOF, when provided with the alternate EOF outside the 10-mile radius from the Plant, meets the intent of NUREG-0696.
Thus, additional shielding and special ventila-tion treatment systems are not required at the primary EOF.
For your information, a copy of the previous PGE submittals of October 21, 1980, April 15 and July 2, 1981 is provided in Attachment 2.
NUREG-0696, Section 4.2, states that licensees who cannot meet the requirements of size and habitability for the EOF must submit to the NRC a request for an exemption to be reviewed by the NRC on a case-by-case basis. PGE hereby requests an exemption from the requirements of additional shielding and special ventilation at the primary EOF.
If the current primary EOF is not accepted by the NRC, PGE will reconsider the option of primary and alternate EOFs, and give serious consideration to desig. ; ting the PGE office in St. Helens, which is approximately 14 miles from the Plant, as the only EOF for the Trojan Plant, in accor-dance with the provisions ir Section 4.2, Table 2, of FUREG-0696. This would clearly Se a less technically desirable option in view of the close proxir -7 of the current primary EOF to Trejan, together with a substantial c.,perience and planning basia for the current facility.
Sincerely, a
W- - J Bart D. Withers Vice President Nuclear Attachments t
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Trojan Nuclsar Plant Dockst 50-344 License NPF-1 ATTACHMENT 1 TROJAN NUCLEAR PLANT SUPPLEMENTAL INFORMATION ON NUREG-0696 Emergency Operations Facilities (EOFs)
This Attachment is submitted to the NRC in-response to the NRC letter of September 2, 1981 to supplement PGE responses transmitted on October 21, 1980, April 15 and July 2, 1981. The information-provided in this Attachment covers 1) feasibility of upgrading the Visitor Tnformation Center (VIC) in accordance with NUREG-0696 criteria, and 2) habitability of the existing VIC.
1.
Feasibility of VIC rodifications The primary EOF is 4ted in the VIC which is approximately
.5 miles WSW of the Plant. The VIC building is designed-in accordance with Uniform Beilding Code and State of Oregon requirements. The building is a substantial structure capable of maintaining required EOF functions for extreme 4
environmental conditions which might reasonably be postu-lated to occur within the life of the' structure.
NUREG-0696 specifies a radiation protection fict4t of 5 and a ventilation system with HEPA Filters for.the EOF when located'within ten miles of the TSC.- To achieve a protection factor of 5, the addition of 8-in. thick concrete walls (or the equivalent) and roof slab to the existing VIC would be required. A structural analysis was performed to evaluate the feasibility of implementing these modifications. The result of.the analysis shows that the existing wall footings would not.be adequate to support the new walls. The existing roof framing and metal decking would also not be structurally adc:;uate to support a new S-in. thick concrete slab roof. Therefore, extensive structural modifications would be required to upgrade the existing VIC building to accomodate new structural elements in order to provide the protection factor of 5.
Modifications would include removal of the existing wall and column footings, metal siding, structural wall, roof framing and concrete floor slab. Because of soil.consolida-tion conditions already experienced, new pile or. tt founda-tions would probably have to be constructed to su)
-t the new walls, columns and roof slab. A preliminary se Lement analysis indicates that a significant weight.additi a due to the modifications for the EOF would affect the VIC struc-ture in terms of additional total and differential settlement, and= tilting. Damage to the existing VIC structure could result.
It is therefore not considered to be structurally
2
' practicable to upgrade the existing VIC building to comply with the NRC requirement for shielding.
Feasibility of upgrading the existing ventilation system in the VIC was also evaluated.
In order to install HEPA filters in the VIC ventilation system per NUREG-0696, extensive modifications will be required to the mechanical rooms which house the supply and return fans, filters, and plenums. There are two supply fans and two return fans in the VIC ventilation system.
Based on the design volumetric flow rate of the ventilation system, apprcrimately ten HEPA filters will be required for each supply fan. The mechanical rooms cannot physically accommodate the required filters. The HEPA filters and prefilters will have to be p)sced on the roof of the VIC Building in a wentherproof housing. Roof penetrations will have to be made ':0 accommodate the revised supply and exhaust ductwork. The revisions to the supply, return, and exhaust ductwork will involve major modifications to the original design. Also, the addition of HEPA filters and prefiltet; w1Al increase the total resistance of the ventilation system, thereby affecting the output of the supply fans. Modifications or replacement of th* supply fans and associated motors is highly probable.
= balancing of the volumetric airflows throughout the buildi,.111 a.lso be necessary.
Tased on the above, it is not prmaticable to modify the existing VIC to comply with the s51elding and ventilation provisions of the NUREG-0696. Further, as discussed in the following sectioc-such provisions are not required for necessary radiologn sl protection.
2.
Habitability of Existing VIC a.
EOF Camma Ray Shielding Factor The gamma shielding factor of 1.33 assumed for the EOF in our October 21, 1980 submittal has been reexamined. The structural composition of the EOF was examined in detail, and the effective shielding factor for 0.7-MeV gamma rays was calculated for the Auditorium of the Trojan VIC (the location of the EOF communications, command and dose assess-ment functions). The results of this analysis confirm the appropriateness of the factor of 1.33.
b.
Meteorology Assumptions The October 21, 1980 submittal used atmospheric dispersion factors (X/Q) calculated using Regulatory Guide 1.145 assump-tions. The X/Q used in the analysis is the 0.5 percentile value over the entire year for the WSW sector, which at Trojan 3
3 is the least frequent wind direction. Regulatory Guide 1.145 includes the wind frequency as part of the X/Q model. Table 1 shows a 30-f t tabular wind rose for the years 1971-74 and 1976-81. The percentage of hours of wind blowing toward the EOF (ENE wind direction) for this period is 0.5 percent.
Table 2 shows the monthly percentage of winds blowing toward the EOF for the period of 9/1/71 to 8/31/74. This table
. hows that the monthly variation of winds blowing toward the EOF is small.
The use of the Regulatory Guide 1.145 models for this analysis is appropriate for the following reasons:
(1) The model refleer.s the fact that atmospheric dis-persion conditicas and wind frequencies are usually directionally dependent; that is, certain air-flow directions can exhibit substantially more or less favorable diffusional conditions than others, and the wind can transport effluents in certain direc-tions more frequently than in others.
(2) The introduction to the Regulatory Guide states that the Guide should be used generically in place of Regulatory Guide 1.4 assumptions for accident analyses until specific NRC guidance is issued.
It further states that until such guidelines are developed, the methodology provided in the Guide is acceptable to the NRC Staff.
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TABLE 1 Annual Wind Rose 33 Foot Level Total Observations = 71,044 9/1/71-8/31/74, 11/1/76-6/30/81 Wind Direction Percent N
15.3 NNE 5.8 NE 1.6 ENE*
0.5 E
1.9 WNW 3.5 l
hM 7.5 NNW 12.5 l
Calm 1.2 Location of Trojan EOF (WSW Sector) l l
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TABLE 2 Monthly Wind Rise for ENE Direction
- 33 Foot Level-9/1/71-8/31/74 Month Percent January 0.3 February 0.4 March 0.5 April 0.9 May 0.6 June 1.1 July 0.9 August 0.7 September 0.4 October 0.3 November 0.4 December 0.5 AVERACE 0.6 i
- WSW Sector l
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L ATTAClefENT 2 TROJAN NUCLEAR PLANT DOCKET 50-344 Previous PGE Submittals on NUREG-0696 Emergency Response Facility (October 21, 1980, April 15, 1981 and July 2, 1981)
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E Portland General Electric Company V:
,,~m vce Resent July 2, 1981 Trojan Nuclear Plant Docket 50-344 License NPF-1 Mr. Darrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, DC 20555
Dear Sir:
p LJ Attached please find supplemental information regarding NUREG-0696
" Emergency Response Facilities" which was discussed at the June 15, 1981 meeting between the NRC and PGE. The Attachment covers five
lity, 4) EOF Data System, and 5) SPDS.
As discussed at the June 15, 1981 meeting, design descriptions of the Emergency Response Facilities were submitted to the NRC on October 21, 1980 and April 15, 1981 as well as in the Trojan Radiological Emer-gency Plan whic; was submitted on December 31, 1980. Subsequent to the first submittal. procurement and construction of the TSC and SPDS were initiated and are in progress to comply with the NRC-required'imple-mentation schedule.
Pursuant to your request, five copies of WCAP-9725 (proprietary) reports entitled " Westinghouse Technical Support Complex" are attached with this submittal. These reports were originally transmitted to the NRC by Westinghouse (T. M. Anderson's letter to J. R. Miller dated June 13, 1980).
In conformance with the requirement of 10 CFR 2.790, as amended, of the
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Commission's regulations, we are attaching with this submittal an appli-cation for withholding from public disclosure and an affidavit by Westinghouse.
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o N GenieralBecideW Mr. Darrell G. Eisenhut O
July 2, 1981 Page two Resolution and approval by the NRC staff in the areas discussed in the attachment are required by August 1, 1981, in order for us to comply with the October 1, 1982 completion date specified in NUREG-0696.
Sincerely, Bart D. Withers Vice President d
Nuclear BDW/G 4j 9A8 Attachment c: Robert A. Clark, Chief Operating Reactors Branch No. 3 Division of Licensing U. S. Nuclear Regulatory Commission Lynn Frank, Director State of Oregon Department of Energy
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6 Trojan Nuclear Plaat Docket 50-344
(A License NPF-1 ATTACEHENT 1 Sheet 1 of 6 c)
TROJAN NUCLEAR PLANT SUPPLEMENTAL INFORMATION ON NUREG-0696 EMERGENCY RESPONSE FACILITIES (TSC, EOF and SPDS)
A detailed design description of the Technical Support Center (TSC),
Emergency Operations Facilities (EOF) and Safety Parameter Display System (SPDS) for the Trojan Nuclear Plant was initially submitted to the NRC on October 21, 1980 in response to the NRC letter of September 5, 1980.
Draft NUREGs-0696 and -0660, NUREGs-0578, -0585, and -0654 (Revision 0) were reviewed and considered in the design of the Trojan Emergency Response Facilities for the October 21, 1980 PGE submittal. Utilization of the EOF and TSC was described in the Trojan Radiological Emergency Plan, which was submitted on December 31, 1980.
Due to the required implementation schedule of January 1,1982 at that time, procurement of the TSC computer system and the construction of the TSC building were commenced in May and December of 1980 respectively, which was before the final issuance of NUREG-0696.
NUREG-0696 was issued in final form in February 1981.
Supplemental infor-mation identifying exceptions to the final NUREG-0696 criteria was trans-mitted by PGE to the NRC on April 15, 1981 with appropriate justifications.
O Pursuant to the June 15, 1981 meeting between the NRC and PGE, the follow-ing additional information is provided.
I.
TSC Building The identification of of fice spaces on the first floor of the TSC is described in Figure 1.
The lunch / conference room located on the northwest corner will be assigned as the of fice space for the NRC personnel during emergency conditions.
The remaining three of fice spaces will be used during normal and emergency conditions by PGE personnel. During emergency conditions the of fice/ conference room in the southeast corner can be used for special conferences and meetings.
Since the TSC is planned to be used by the Plant personnel during normal operations, the work area next to the command center is divided by temporary partitions about five feet in height as office spaces.
This area can also be used for of fices during emergency conditions.
The command center, which is approximately 985 sq ft, is located on the southwest corner of the building and is equipped with accordian-type curtains to isolate that space from the remaining office spaces.
The detailed dimensions and layout of the command center is described in WCAP-9725 (June 1980), Figure 2.1.
As shown in Figure 1, the TSC Building is also equipped with a decontamina-tion room to prevent radioactive contamination of the building because of personnel access during emergency conditions.
Decontamination equipment as well as anticontamination clothing will be located in the decontamina-Q tion room.
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4 6
Trojan Nuclear Plant Docket 50-344 License NPF-1 ATTACEMENT 1 Sheet 2 of 6 In case the need arises to travel between the TSC and the control room during emergency conditions, the optimum travel route is out the emergency exit or the entrance door of the north side of the TSC, east through the yard to the Turbine Building doors (Figure 2).
There are three doors on the west side of the Turbine Building at El. 45 ft. and two doors on the south side. A person traveling to the control room is expected to take one of the Turbine Building doors, walk through the Turbine Building El 45 ft. floor to the Control Building and take either an elevator or stairs to the El. 93 ft. level where the control room is located.
In case this travel route is not available, there are also stairways internal to the Turbine Building which connect all three floors to the access door from the yard.
Each floor of the Turbine Building is connected to the Control Building floors by doors.
Therefore, the access through the Turbine Building to the Control Building is assured. Travel through the Turbine Building will also provide shielding from radioactivity releases which may be occurring.
In any case, the travel between the TSC and the control room can be conducted without passing through the security gate.
II.
TSC Computer System A general description of the TSC Computer System was provided in the October 21, 1980 PGE submittal, which was supplemented by WCAP-9725 I)
(June 1980) for detailed design philosophy and descriptions (five copies
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of WCAP-9725 are attached to this submittal for your review). As discussed in the October 21, 1980 submittal, the Westinghouse Computer System located in the TSC will be utilized to accomplish the functions of data collection, data manipulation, and data display for the TSC, the SPDS and the Bypassed and Inoperable Status Indication (BISI) System. A description of the SPDS and BISI System is provided in Section V of this report.
The parameters to be included in the TSC Data System are delineated in Table 1.
The parameters in Table 1 constitute critical parameters consi-dered necessary for recognition and mitigation of accident -conditions.
Other parameters, which are specified in Regulatory Guide 1.97 (Revision 2) but not included in the Trojan TSC Data System, will be made available on demand to the TSC and EOF by telephone communication.
Containment sump water temperature and Containment spray flow,which are not currently mea-sured, will be inferred by the Containment temperature and Containment spray pump discharge pressure.
Qualifications of the components which provide input to the TSC Data System are presently being reviewed as part of the Regulatory Guide 1.97 review.
Information regarding equipment qualifications will not be available until the Regulatory Guide 1.97 review is completed in 1982.
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s Trojan Nuclear Plant Docket 50-344 License NPF-1 ATTACHMENT 1 Sheet 3 of 6 g
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III. EOF HABITABILITY The Trojan EOF is located in the Trojan Visitors Information Center (VIC),
approximately 700 meters WSW of the Containment. A floor plan of the EOF is shown in Figure 3.
The VIC possesses several advantages as an EOF location:
A.
Its closeness to the Plant allows for fast activation, and permits the use of Plant personnel as the initial EOF staff.
B.
It can serve as an Assembly / Decontamination Area for personnel evacuated from the Plant. Any personnel can then quickly travel to the Plant if needed.
C.
The EOF staf f, including the Emergency Response Manager (PGE Vice-President, Nuclear), is located close to the Flant, providing mcre effective control of the emergency.
/
D.
NRC, FEMA, and State and county representatives are located close to the Plant.
E.
The VIC is a good location from which to control site security.
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F.
The VIC has been used as the EOF since 1975. Plant personnel, as well as State and county emergency response personnel, are N'
familiar with the facility and its location.
G.
Using the VIC as an EOF helps to utilize existing structures near the Plant most efficiently.
H.
A low wind frequency (0.06%) toward the WSW provides good assurance of EOF habitability.
The EOF is a well engineered structure that is designed to withstand a 100 yr.
flood.
An analysis has been performed on the ability of the EOF to withstand a 100 yr wind. At this wind level, some damage to the EOF Building may occur, but will likely be limited to broken windows.
However, the command, communi-cations, and dose assessment functions of the EOF take place in the building auditorium, which has no windows.
The auditorium is protected from the building areas that have windows by internal walls. The re fo re, the essential functions of the EOF are not expected to be adversely affected by a 100 year wind.
A description of the operation of the EOF and the EOF staff is contained in Chapter 2, Section 5.2 of the Trojan Radiological Emergency "lan.
The alternate EOF is located at the PGE St. Helens District of fice in St. Helens, r' w Oregon (13 miles from the Plant), and if activated will have the same
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staff and functions as the primary EOF.
Emergency equipment and supplier are stored at the alternate EOF, and all E)F communications and dose assessment capabilities exist at the alter 1 ate EOF.
Trojan Nuclear Plant Docket 50-344 License NPF-1 ATTACHMENT 1 Sheet 4 of 6 f3i )i The alternate EOF will be prepared for activation whenever the EOF is acti-vated by notifying the St. Helens District Manager.
If evacuation of the EOF is required, the St. Helens office staff will be notified to begin activation of the alternate EOF.
During the time required to move the EOF staff to the alternate EOF (approximately 30 minutes), the command, communications, and dose assessment functions of the EOF will be assumed by the TSC.
The TSC staff is trained and qualified to carry on these tasks, and the required communications and dose assessment facilities exist in the TSC.
If the Emergency Response Manager (PGE Vice-President, Nuclear) has already reported to the EOF to assume command of the PGE emergency organi-zation (Emergency Coordinator), he will turn his duties as Emergency Coordinator back to the Plant General Manager in the TSC until the alter-nate EOF is operational.
State and county representatives at the EOF do not have decision-making roles, so relocation to the alternate EOF will not affect the emergency response capabilities of the State and county EOCs.
PGE will include anticipatory protective action criteria for relocation to the alternate EOF in the next amendment to the Trojan Radiological Emer-gency Plan which is expected to be issued in Augusc 1981. These antici-
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patory protective action criteria for the EOF, which are shown below,
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should permit relocation prior to substantial releases of radioactivity from the Containment:
Fuel melting indicated by Containment area monitors or dose rate measurement outside Containment; and 1.
Containment sprays and air coolers not func-tioning; and 2.
Containment pressure greater than 70 psig for 2 min.; or 3.
Other conditions exist which will lead to loss of Containment integrity.
The anticipatory protective action criteria are in addition to the following criteria which are based on radiation levels in the EOF:
1.
One rem /hr whole body; or 2.
100 MPC 1-131; or 3.
Ten times these levels persisting for greater ew than 5 min.
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PGE will test the relocation to and operation of the alternate EOF during the 1982 Trojan Emergency Plan Exercise.
Trojan Nuclear Plant Docket 50-344 License NPF-1 ATTACHMENT 1 Sheet 3 of 6 O
HEPA filters are not required for the EOF ventilation system because of the protective action criteria established for the EOF.
The anticipatory protective action criteria for the EOF (listed above) will require reloca-tion to the alternate EOF before slow-developing accidents (i.e., greater than one hour) reach the point where significant particulate radioactivity will be released. For accidents involving a quick release of radioactivity (i.e., less than one hour) the releases will initially be dominated by volatile nuclides (noble gases and iodine).
Particulate radioactivity will consist of iodine or short-lived noble gas daughters which do not contribute significantly to the dose rate inside the EOF.
Furthermore, the presence of these short-lived noble gas daughters in the EOF would not be prevented by the HEPA filters because of their continuous generation by the decay of the noble gases inside the EOF.
The dose rate and iodine air concentration protective action criteria will ensure relocation to the alternate EOF before significant long-lived particulate radioactivity such as Cesium-137 will be released.
IV.
EOF DATA SYSTE,M The TSC data set discussed in Section II above will be transmitted to the primary EOF in the Visitors Information Center. As a minimum, both meteoro-logical and radiological data will be transmitted to the alternate EOF in r-g St. Helens, Oregon.
U As shown in Table 1 the EOF data set contains the meteorological parameters required to calculate dispersion coefficients in the Evacuation Planning Zone (i.e., wind direction, wind speed, and 200-33 ft delta temperature) and Plant parameters required to calculate offsite doses (e.g., effluent radioactivity concentrations, effluent streau flow rates, area radiation monitor readings and Containment pressure).
These data can be input along with radiation dose and airborne radioactivity measurements from field monitoring teams into an existing computer program which operates on a commercial time-share computer system located in Los Angeles, California.
The computer can be accessed via commercial and PGE telephone circuits.
Since PGE-owned microwave telephone circuits connect both the primary and alternate EOFs directly to major telephone exchanges in Portland, Oregon, access to the time-share computer will not be affected by the potential overload of telephone circuits in the vicinity of the Plant following an accident.
The computer program calculates atmospheric dispersion factors and offsite doses upon which offsite protective action recommendations can be made.
The program incorporates terrain correction factors which compensate for the river valley terrain present in the Evacuation Planning Zone.
By October 1,1982 this program will be modified to calculate relative concentrations and offsite doses consistent with NUREG-0654, Appendix 2.
PGE's course of action regarding the remaining items of NUREG-0654, Appendix 2 (i.e., a meteorological measurements system in accordance with
[)
proposed Revision 1 to Regulatory Guide 1.23 and remote interrogation of meteorological measurements) will be submitted to the NRC before Januarv _1,
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Trojan Nuclear Plant Docket 50-344 License NPF-1 ATTACHMENT 1 Sheet 6 of 6
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V.
Safety Parameter Display System (SPDS)
The SPDS is an integral part of the TSC computer system and constitutes a series of graphic displays to provide the current status of the parameters important to safety in the main control room and the TSC command center.
One of the graphic displays utilized for the SPDS is an iconic display of the critical parameters (Figures 4-3,
-4, and -5 in WCAP-9725) in order to provide a quick recognition of accident conditions. Even with current technology, this display cannot be made available at the EOF without a great deal of difficulty and at a very high cost.
The purpose of the iconic display is to provide an immediate identificaticn and analysis of accidents in the control room and TSC.
These functions are not performed by the EOF.
The re fore, the iconic displays are not required in the EOF.
Instead, the values of the critical parameters displayed in the iconic display along with their corresponding technical limits, will be transmitted to the EOF by the TSC Computer System. This will ensure that EOF personnel have the same data as that in the control room and TSC.
The TSC Computer System which develops the SPDS is not a safety system and thus, is not seismically qualified.
In response to the NRC require-ments in NUREG-0696 for a seismically qualified backup display system, the critical parameters listed in Tabic 2 will be displayed on a seismi-(~'/
cally qualified " post-accident" panel (panel C-09) which has been added h
to install the post-TMI instruments in the control room.
Among the eight parameters listed in Table 2, the neutron flux indicators and steam generator water level recorders are displayed on panels C-02 and C-05, respectively.
Although equipment qualifications for these parameters are still under evaluation for Regulatory Guide 1.97, selection of the parameters are based on ANS Standard 4.5 for accident recognition and mitigation.
The distances for detectability (i.e., trend reading of recorder and relative position of indicator) of recorder pointers and indicator pointers have been reviewed in a preliminary human engineering review.
Detectability of meter pointers was subjectively evaluated and found to be adequate at 15 ft from the panel. However, detectability of recorder pointers was poor beyond 10 ft as was differentiation of recorder trace color.
Since the distance between panels C-05 and C-09 is approximately 24 ft, the detectability distance of 10 ft for recorders from each panel leaves about 4 ft in which an operator needs to travel for adequate detectability. Obviously, detectability of the meters is not a problem.
The distance of 4 f t (or 2 ft for each side) is insignificant compared to normal walking distance for the control room operators from panel to panel.
The only two recorders for which an operator would have to move'approxi-mately 2 ft in either direction are steam generator water level on C-05 and Containment pressure on C-09.
The three panels, C-02, -05 and -09, are thus considered to be adequately located for the purpose of a backup SPDS.
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Trojan Nuclear Plant Docket 50-344 License NPF-1 TABLE 1 Sheet 1 of 4 7-~
V TROJAN NUCLEAR PLANT TSC Data System Parameters (June 12, 1981) 1.
SI Pe'r O!acharge Flow 2.
Charging Flow to BIT 3.
Charging Flow 4.
Letdown Flow 5.
RCP Seal Injection Flow 6.
Aux. Bldg. Vent Flow 7.
Containment Purge Flow 8.
Air Ejector Flow 9.
Effluent Flow (Discharge and Dilution) 10.
Ec' Seal Leakage Flow
( ')s 5
11.
Emergency Boration Flow 12.
RCS Flow 13.
Steam Generator Feedwater Flow 14.
Steam Generator Steam Flow 15.
RHR Flow to Hotlegs 16.
RHR Flow to Coldlegs 17.
AFW Flow 18.
Pressurizer Level 19.
BAT Level 20.
VCT Level 21.
PRT Level 22.
Steam Generator Level (narrow and wide ranges) 23.
Accumulator Level
-s
%J 24.
RWST Level
.I
f.
Trojan Nuclear Plant Docket 50-344 License NPF-1 TABLE I Sheet 2 of 4
%J 25.
NA0H Tank Level 26.
RCDT Level 27.
Clean Waste Tank Level 28.
Waste Monitor Tank Level 29.
Dirty Waste Drain Tank Level 30.
Dirty Waste Monitor Tank Level 31.
CST Level 32.
Containment Sump Level 33.
Reactor Vessel Level 34.
Emergency Diesel Tank Level 35.
Diesel AFW Pump Tank Level 36.
Wind Speed (U) 37.
Wind Direction 38.
Dew Point Temperature 39.
Delta Temperature 40.
Containment Hydrogen 41.
Nuclear Instrumentation System (Source, Intermediate and Power Ranges) 42.
VCT Pressure 43.
Charging Pump Discharge Pressure 44.
Steam Generator Feedwater Flow Pressure 45.
Nonregenerative RX Outlet Pressure 46.
RCS Pressure 47.
Pressurizer Pressure 48.
PRT Pressure
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47.
SG Steam Pressure L J
Trojan Nuclear Plant Docket 50-344 License NPF-1 TABLE I Sheet 3 of 4
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52.
Accumulator Pressure 53.
RHR Suction Pressure 54.
Containment Spray Pump Discharge Pressure
- 53. Waste Gas Decay Tank Pressure 56.
Turbine First Stage Pressure 57.
Containment Pressure 58.
Instrument Air Header Pressure 59.
PRMs 60.
ARMS 61.
In-core Temperatures 62.
Letdown HX Outlet Temperature 63.
RCP Seal Water Temperature O
5 64.
RCP Lower Seal Water Bearing Temperature 65.
RCS Hotleg Temperature 66.
RCS Coldleg Temperature 67.
Pressurizer Liquid Temperature 68.
Pressurizer Steam Temperature 69.
Pressurizer Relief Valve Discharge Temperature 70.
PRT Temperature 71.
Feed Water Temperature 73.
RCS Delta Temperature 74.
RCS T avg.
75.
Tref 76.
Containment Temperature 7_s
's
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Trojan Nuclear Plant Docket 50-344
/-w License NPF-1 TABLE 1 Sheet 4 of 4 0
77.
RCP Motor Bearing Temperature 80.
Containment Isolation Valve Positions (Open/Close) 81.
Safety-Related System Valve Positions 82.
Safety-Related Pump Status 83.
ESF BUS / Breaker Status (>480 volt level)
O KM/4mm7B10 L j
I Trojen Nuclect Plant Docket 50-344 License NPF-1 TABLE 2 f%
Q TROJAN NUCLEAR PLANT CRITICAL PARAMETERS FOR BACKUP SPDS ON POST-ACCIDENT PANEL (C-09)
ANS Standard 4.5 Type B Function Panel Parameter (Reg. Guide 1.97 Type B)
Number Installed?
1.
Neutron Flux (a) Reactivity Control Operator Yes Console C-02 2.
Reactor Vessel Reactor Core Cooling C-09 Operational Water Level upon NRC Approval in 1982.
3.
RCS Hotleg Reacto: Jore Cooling C-09(b) ye, and Coldleg Temperatures (a) 4.
RCS Pressure (a) Reactor. Core Cooling C-09(b)
Yes and RCS Integrity 5.
Containment RCS and Containment C-09 1/1/82 O'
Water Level Integrity 6.
Containment RCS and Containment C-09 1/1/82 Pressure Integrity (Recorder)(a) 7.
Containment RCS Integrity C-09 1/1/82 High-Range Area Radiation (a) 8.
Steam Generator Not Included C-05 Yes i
Water Leve (Recorder) a)
(a) Seismic qualification is still under review and will be included in Regulatory Guide 1.97 evaluation.
(b) Information is available on demand through the subcooling margin monitor located on C-09.
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.t bc Messrs. With2re, Brochl, Yundt, Steele, Orser, Sullivan, Galdos CEN: GEN EMGR 7:SA,TMI TNP: POW ST OP 2-1:RERP 1.0 - all w/ attach rm
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W eM :,? m &&csm r April 15, 1981 Trojan Nuclear Plant Docket 50-344 License NPF-1 Mr. Darrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, DC 20555
Dear Mr.. Eisenhut:
Forwarded hereby is PGE's response to h7C generic letters 81-10 (February 18, 1981) and 81-17 (March 5, 1981), requesting a commitment to the imple=entation schedules for TMI Action Plan Item III. A.1.2 (l'pgrade Ecergency Support Facilities) and NUREG-0696.
A.
Emergency Response Facilities PCE's basic commitments to the e=ergency response facility requirement:.
of Action Plan Item III.A.1.2 and NUREG-0696 were submitted on October 17, 1979 and October 21, 1980. At this time, PGE commits to the require-ments and schedule set forth in generic letter 81-10 with the following exceptions:
1.
Technical Support Center (TSC) Computer System:
The TSC computer system will provide the necessary informa-tion for the Plant management, Engineering, and technical personnel assigned to the TSC to aid the control room operators in handling accident conditions. The data set available for display at the TSC includes the Safety Para-meter Display System (SPDS), the Bypassed and Inoperable Status Indication (BISI), pertinent radiological and meteoro-The logical data, and the critical Plant system parameters.
critical system parameters available in the TSC computer system do not include the entire set of variables listed in Regulatory Guide 1.97 (Revision 2), since the Regulatory Guide is still undergoing evaluation for its implementation in 1983. However, the set of paramenters included in the TSC data system covers the majority of the variables in oh Regulatory Guide 1.97 (Revision 2), which are pertinent to determine the Plant systems dynamic behavior through the course of the accident and appropriate mitigating actions, yg (yf ?[0hA Dbl5 m c. w. w n. m,, m J
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Mr. Darrell G. Eisenhut April 15, 1981 Page two Selection of the parameters is based on an evaluation of Plant systems / functions which are critical in mitigating
. dent conditions. The critical Plant systems / functions include: reactor core integrity, primary coolant system integrity, reactor core cooling, Containment integrity, and radioactive effluents.
The TSC data set will also be made available on demand to the EOF and alternate EOF.
2.
Emergency Operations Facility (EOF) Location and Habitability:
PGE's October 21, 1980 letter described the location and habitability of the Trojan EOF, which is located at the Trojan Visitors Information Center (VIC), approximately 1/2 mile from the Plant. A fully-equipped alternate EOF is' located at the PGE office in St. Helens, Oregon, 13 miles from the Plant. The alternate EOF will have the same data displays and communications as the primary EOF. The primary EOF does not meet the NUREG-0696 requirements for shielding
,-~s and ventilation system filtration, although the building
(#
does meet the 100 year flood criteria. The co=pliance of the EOF with the 100-year wind criteria is currently being evaluated.
In PGE's October 21, 1980 response, an evaluation of the habitability of the EOF was made versus the Standard Review Plan (SRP) 6.4 dose criteria. Due to the location of the EOF (in a very infrequent wind direction), the results of the analysis indicated that the EOF met the SRP 6.4 habita-bility criteria for a Design Basis Accident at the Trojan Nuclear Plant.
In view of this, and the desirability to keep overall control of the e=ergency as close to the Plant as possible, PGE hereby requests that the Trojan EOF be exempted from the shielding and ventilation requirements of NUREG-0696.
3.
NUREG-0654, Appendix 2 Implementation:
NUREG-0654 Appendix 2 requirements, as they relate to descriptions of the TSC and EOF instrumentation and data systems, will be addressed according to the implementation schedule for NUREG-0654 Appendix 2 (system description submitted by January 1, 1982). PGE conpliance with 1
to be
/ 'l the scheduled dates for equipment installation (July 1,
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1982) and full operation (October 1, 1982) will be depen-dent upon hardware and software availability and equipment delivery schedules.
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r Portland Gcasm' Eb2Co:rgany
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Mr. Darrell G. Eisenhut April 15, 1981 Page three 4.
Safety Parameter Display System:
The SPDS is an integral part of the TSC computer system which provides the data collection, manipulation and the CRT graphics displays of the Plant system safety parameters.
The SPDS displays will be located in the control room and the TSC.
The SPDS is not designed to meet seismic qualifications since it is a non-safety system and is not required to mitigate accidents. In response to the requirements in NUREG-0696 for seismically-qualified backup displays, a minimum set of the critical parameters will be displayed on a seismically-qualified panel (post-secident panel) which is installed to incorporate certain post-TMI instru-ments in the control room. The critical parameters located in the post-accident panel include:
tha reactor vessel water level, Containment water level, and Containment high-range pressure. In addition, the RCS hot leg and cold leg temperatures and the RCS pressure are also avail-able on this panel although the circuitry from isolators O
located elsewhere in the Control Room to the display devices is not seismically qualified. Other pertinent parameters, such as core flux, Containment isolation valves, and radio-active effluent monitors are already displayed in locations which are easily accessible end highly visible from the operating station of the control room operator. These parameters constitute a minimum set of critical parameters which meet the intent of ANSI Standard 4.5, 1980, for monitoring Plant safety functions.
(
B.
Emergency Staffing Requirements to this letter shows PGE's proposed staffing levels for emergency situations. The proposed on-shif t staffing levels are con-sistent with Table B-1 of NUREG-0654, Revision 1, provided the following assumptions are made:
1.
The requirement for two maintenance technicians on-shif t can be met by utilizing a single on-shif t main-tenance technician to perform both mechanical and electrical /1&C maintenance, and by utilizing the Plant Auxiliary Operators to perform Radwaste Operator functions.
2.
The Security Watch Supervisor and other security Os personnel can be_ utilized to perform a portion of the
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Hr. Darrell C. Eisenhut April 15, 1981 Page four communications / notification function until the TSC is activated.
The on-shif t staf fing levels will be fully implemented by September -1, 1981 with the exception of the licensed operators and maintenance techni-cian. The licensed operators requirement will be implemented according to the schedule specified in Task Action Item 1.A.1.3.
The on-shift maintenance technician position will be filled on an interim basis by on-call personnel who will be able to augment the Plant staff within 60 minutes; the on-shif t maintenance technician requirement will be implemented fully by July 1, 1982.
The PCE proposed augmentation capability complies with Table B-1 augmentation requirements with the exception of the following areas:
1.
Table B-1 requires a certain number of personnel to report to Trojan within 30 minutes. PCE proposes that the minimum augmentation time be 60 minutes.
A 60-minute augmentation time is appropriate for the f-~3
(,)
following reasons:
Compliance with a 30-minute time requirement a.
is not feasible given the location of the Trojan Nuclear Plant relative to the areas where Plant personnel live. Sixty minutes is a more reasonable amount of time for Plant personnel living in the vicinity of the Trojan Nuclear Plant to report onsite after notification. In reality, some per-sonnel may be able to report onsite in less than 60 minutes.
b.
It is unlikely that requiring Plant per-sonnel to report onsite within 30 minutes would add significantly to the Plant staff's j
ability to cope with a quick-occurring accident, such as a ste2, generator tube i
rupture, main steam line break, or WASH-1400 PWR-8 or PWR-9 accident. Accidents that involve core degradation and significant radioactivity releases to the environment occur over much longer periods of time (ie, several hours); for these accidents, a 60-minute augmentation time would be sufficient.
(}
2.
The total number of Chemistry and Radiation Protection (C&RP) Technicians reporting onsite within 60 minutes l
i
-9 Po;Wm! Gaanra' El:2 Co:n):ny O
Mr. Darrell G. Eisenhut
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April 15, 1981 Page five is 3 versus 13 required by Table B-1.
As shown in, these three personnel, in addition to the two C&RP technicians on shift, will provide suf ficient chemistry and radiation protection exper-tise to meet all anticipated ecergency requirements in the short te rm. It should be noted that licensed Plant operators are qualified in radiation protection access control and conitoring procedures. The Table B-1 requirement is not feasible in that it would require a total of 15 of the 18 available C&RP technicians to be onsite within 60 minutes.
3.
The total number of maintenance technicians reporting onsite within 60 minutes is 2 versus 5 in Table B-1.
These two personnel, along with the on-shif t Maintenance Technician and Auxiliary Operator (Radwaste Operator),
will provide sufficient maintenance expertise for emergency situations in the short term.
In addition, the following interpretations of the Table B-1 augmentation
()
requirements were made:
1.
The senior health physics expertise identified in Table T,1 is being interpreted to be the Trojan Radia-tion Protection Supervisor or one of his alternates.
2.
The senior manager identified in Table B-1 as the EOF Director is interpreted as the duty Trojan Manager, Plant Services.
The augmentation levels shown in Attachment I will be fully implemented by September 1, 1981.
PCE believes that the minimum staffing and augmentation requirements given in Attachment 1 provide a reasonable effort to comply with the intent of Table B-1 to have an effective and responsive emergency organization during off-hours.
It should also be noted that the Trojan Radiological Emergency Plan Implementing Frocedures contain procedures for notifying the Plant
~
e PO:Uwd Gm::C ElnbCompn; Hr. Darrell G. Eisenhut (Q/
April 15, 1981
_Page six staff during off-hours. Therefore, the actu 1 nu=ber of personnel augment-ing the Plant off-hours staff would likely be considerably greater than the minimum number specified in Attachment 1.
Sincerely, i
Bart D. Withers Vice President Nuclear Y'
BDW/IDW/SGG/41m8B15 c:
Mr. Robert A. Clark, Chief
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J. W. Let.tich ~~
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Operating Reactors Branch No. 3 j
Division of Licensing U. S. Nuclear Regulatory Cormission
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Mr. Lynn Frank, Director State of Oregon
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Department of Energy
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W. S. Orser l
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4-ATTACllMENT 1 TABLE 2:5.2-1 Sheet 1 of 2 STAFFING REQUIREMENTS FOR TROJAN NUCLEAR PLANT EMERCENCIES Capability l
Position Title for Additions Major Functional Areas Major Tasks or Expertise on Shift 60 Min }
Plant Operations and Shif t Supervisor 1
Assessment of Operational Assistant Shif t Supervisor 1
Aspects Control Operator 1
Assistant Control Operator 1
Auxiliary Operator 3
Emergency Direction and Shif t Supervisor 1
Control (Emergency Duty Plant General Manager 1
Coordinator)
Notification /
Notify licensee, State, Security Watch Supervisor /
2 Communication local, and Federal per-Security personnel sonnel and maintain Shift Supervisor [a]
1 communication Duty Manager, Plant 1
Services (EOF Director)
Radiological Accident Emergency Operations Duty Manager, Plant 1
Assessment and Support
' Facility (EOF) Director Services of Operetional Accident Assessment Offsite dose assessment Assistant Control 1
Operator [b]
2 Engineering Emergency Team (TSC)[bj 1
Duty Radiation Protection
! Supervisor (EOF)[b]
2 Offsite surveys Field Team (C&RP Technicians)
Onsite (out-of-Plant)/
C&RP Technicians 2
u 1
in-Plant surveys /
j chemistry / radiochemistry l
l
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ATTACp
_1, TABLE 2:5.2-1 Shrct 2 cf 2
. /.
Capability Position Title for Additions Major Functicnal /.reas Major Tasks or Expertise On Shift 60 Min f
Plant System Engineering, Technical Suppori Duty Manager, Technical 1
Services Repair and Corrective Shift Technical Advisor 1
1 Actions Engineering Emergency Team 2
i Duty Maintenance Supervisor 1
2 Repair and corrective Maintenance Technician 1
actions Protective Actions Radiation protection:
C&RP Technicians 2[C]
3(Cl a.
Access control (In-Plant) b.
IIP coverage fo r repair, corrective actions, search and rescue, first aid, 1
i and firefighting c.
Personnel monitoring d.
Dosimetry Firefighting Fire Brigade Sld) l Rescue Operations C&RP Technicians 2[el llel 1
and First Aid Site Access Control and Security, firefighting Security personnel All per Personnel Accountability communications, personnel Security Plan
}
accountability Totat:
13(fl
{
13(f) {
[a] Security Watch Supervisor and Security personnel perform initial notifications of State and county Supervisor notifies NRC and maintains communications until Duty Manager, Plant Services agencies; Shift arrives at EOF.
[b] Assistant Control Operator performs dose assessments until the TSC is activated; Engineering Emergency Team performa dose assessments at TSC until EOF is activated; Duty Radiation Protection Supervisor per-forms dose assessments at the EOF.
[c] Duty performed by C&RP Technicians assigaed to onsite surveys (Sheet 1).
Licensed operators are also trained in radiation protection.
[d] Fire Brigade consists of three operators and twa Security personnel.
le] Duty performed by C&RP Technicians assigned to onsite surveys (Sheet 1).
[f] Does not include Security personnel.
TDW/SGG/4 caw 7B18 O'1
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He2cro. Broehl, Durh =, H:idst, Str:2t, Yundt, Lent cb Zimmerman, Chricten An, Grid 3, Sullivan, P. Chang-Lo, ff T. M. Shem, L. W2iclogal, D. Axtoll, R. Nyltnd, LIC, Reading File, GEN: GEN ENGR 7:SA,TMI, TNP: GEN ENGR 7:SA,F-15.8
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b Portland General ElectricCorrpany at
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Docket 50-344 License NPF-1."
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Mr. Darrell G. Eisenhut, Director Sl.
Division of LicensinS
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U. S. Nuclear Regulatory Commission K.
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Washington, D. C.
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Dear Mr. Eisenhut:
kE h W-Attached is a report on the Trojan Nuclear Plant Emergency Response Facilities which provides a Jescription of the Technical Support Center (TSC), Emergency Ope v: ions Facility (EOF) and Safety Parameter Display System (SPDS) for the Trojan Nuclear Plant. This report is submitted in response to the NRC letter of September 5, 1980 " Preliminary Clarification of IMI Action Plan Requirements", which requires licensee's submittal of the information and description of the Emergency Response Facilities by January 1, 1981.
Draft report NUREG-0696 issued in August 1980 was considered in the design of Trojan Emergency Response Facilities where possible.
In this regard, the SPDS and TSC/ EOF instrumentation systems had already been pcocured and were in advanced stages of design and fabrication at the time draft NUREG-0696 was ic wed in August 1980.
It is currently planned that the TSC and the EOF will be operational by April 1, 1982, and the SPDS by Jatuary 1, 1982.
Sincerely, Bart D. Withers Vice President Nuclear
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BDU/L m/41m2A7
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Attachment p
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VA NM,L w c:
Mr. Robert A. Clark, Chief Operating Reactors Branch No. 3
[fg Division of Licensing C g T. Yundt U. S. Nuclear Regulatory Commission Mr. Lynn Frank, Director
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p State of Oregon
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W. Lentsch g,jj e C t-Department of Energy Vl 10/02DMf
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I TROJAN NUCLEAR PLANT D1ERGENCY RESPONSE FACILITIES l
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October 20, 1980 i
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PORTLAND GENERAL ELEC'"RIC COMPANY
.121 S. W. Salmon Street Portland, Oregon 97204 h ttp c o f $0lblSO $ A
1.0 INTRODUCTION
This report describes the preliminary design, design criteria, and safety analysis aspects of the Emergency Response Facilities for the Trojan Nuclear Plant.
Included is information on the Technical Support Centet (TSC), Emer2ency Operations Facility (EOF), Safety Parameter Display System (SPDS), and Nuclear Data Link (NDL).
2.0 TECHNICAL SUPPORT CENTER Fortlend Ceneral Electric Company (PGE) f S in the process of designing and constructing a Technical Support Center (TSC). This TSC is scheduled to be functional by April 1,1982. An interim TSC as described in PGE's January 2,1980 Less-ns Learned submittal will be maintained until such time as the new TSC is completed.
In order to provide assurance that highly reliable and available systems are installed in the Plant, a verification and validation process will be established for the design, development, qualification, and installation of the TSC.
2.1 TSC FUNCTION The TSC is an onsite facility that provides a workplace for Plant manage-ment and technical support personnel during emergency conditions. Its primary function is to provide support to the control room operating staff while reducing congestion in the control room. Active readouts of Plant conditions and current Plant records, procedures and drawings are provided in the TSC.
TSC personnel will monitor reactor systems status and evaluate Plant syst :m abnormalities. These monitoring and evaluation functions are supported by the Technical Support Center Computer System. The Technical Support Center Computer System, separate from the Plant process computer, is located in the TSC and will drive the Safety Parameter Display System and the instrumentation displays in the TSC, and the radiological and 1
O' s
meteorological data in the EOF. This Computer System provides signal collection, signal manipulation, data display and information storage /
recall capabilities. Sufficient data is available to determine the Plant steady-state and dynamic behavior prior to and throughout the' course of an accident.
The TSC will be activated during Alert, Site Area Emergency and General Emergency levels of emergency action as defined in the Trojan RERP. The TSC personnel will provide guidance to the control room operating person-nel in the management of abnormal conditions and in accident mitigation.
TSC personnel will also provide Plant systems support for the PGE manage-ment located in the EOF.
During early stages of an emergency, the TSC will also perform the functions of the EOF until those functions have been shifted to the EOF.
The TSC is not an engineered safety feature and is not required for either safe operation or accident control and mitigation. The availability of the TSC does not affect capability to safely operate the Trojan plant and the Plant operations can continue if the TSC is not available. However, since it does provide the capabilities that allow technical support personnel to assist the control room operating personnel in assessing the state of the Plant under abnormal conditions, it is intended that the TSC have as high an availability as reasonably possible.
During normal Plant operating conditions, the TSC is planned to be used by Plant Technical Services Department personnel. This normal usa of the facility has the advantages of providing for the assurance of continuing operability of building support systems and for the maintenance of up-to-date drawings and records for emergency use. The facility will also be used for training and emergency drills. Arrangement is such that
,these normal uses do not interfere with immediate activation of TSC operations in the event of an accident or degrade TSC preparedness or systems reliability..
md
i 2.2 TSC LOCATION i
The TSC is a new structure that will be constructed adjacent to the west side of the existing Condensate Domineralizer Building (Figur 1). There will be a seismic gap between the TSC and the Condensate Domineralizer i
Building.
It will be within the Plant security perimeter; the west and south walls of the TSC Building will form a portion of the Plant security boundary. At this location, the TSC is in close proximity to the control room (within 2 min. travelling time) to provide ready access to informa-tion and personnel in the control room, if necessary.
2.3 TSC STAFFING The TSC is designed to permit occupancy by a minimum of 25 individuals with adequate work space to allow effective use of the facility.
The TSC staff assignments and the description of their responsibilities will be described in detail in the Trojan Radiological Emergency Response Plan (RERP).
In the event of an emergency, those using the TSC upon activation -include senior plant management and technical personnel (i.e., the Plant General i
Manager, Operations and Maintenance Manager, Technical Services Manager, Supervisors for Maintenance, Chemistry and Engineering, Engineering Emergency Team), the Nuclear Regulatory Commission (NRC), the Oregon Department of Energy, Westinghouse, and Bechtel. In the absence of key j
personnel, an alternate or subordinate will assume the responsibilities f
of the unavailable person.
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2.4 TSC SIZE AND ARRANGEMENT
(
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,The TSC is a separate structure with nominal size of 60 by 100 ft. with a first floor and a basement providing a total area of approximately i
12,000 ft2 The normal entrance is through the northeast door; a second door is provided as an emergency exit. These doors provide a vestibule-type arrangement. Knockouts in the walls, which provide
~
shielding equivalent to the walls, are provided on both floors for the t
movement of major equipment. The basement of the'TSC includes a battery p
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room, electrical and mechanical equipment rooms, computer room, a storage area, and a records storage vault. The first floor is the personnel work area. A conceptual floor plan for the TSC is presented in Figures 2 and 3.
The major portion of the first floor, approximately 3,000 ft2 is devoted to the command center and work area, and can easily accommodate occupancy by a minimum of 25 people. Other areas directly related to the TSC func-tion include the records room and offices. A portion of the TSC work area will be provided with temporary partitions as a space for five NRC personnel. The remainder of the floor consists of personnel support areas, including restrooms, sleeping areas, a kitchen and the cafeteria /
conference room.
2.5 TSC STRUCTURAL DESIGN The TSC is designed as a Seismic Category II structure. The building is a two-story windowless structure constructed of reinforced concrete on a drilled caisson foundation. Wall and roof thicknesses are determined from radiological considerations.
The design loading on the structure is as follows:
1.
Dead Loads - Actual weight.
2.
Live Loads - UBC values.
3.
Wind Loads - Trojan FSAR Section 3.3.1.
4.
Seismic Loads - 0.15g with spectral analysis.
5.
Flood Loads - The structure is above the 100 year flood elevation, so hydrostatic loads are not considered.
Load combinations and factors are in accordance with Section 3.8.1.3.4 of the Trojan FSAR.
The TSC design reflects the following design codes and standards:
a.
Uniform Building Code with State of Oregon inserts.
b.
FSAR for Seismic Category II structures (FSAR Section 3.8).
c.
Building Code Requirements for Reinforced Concrete, ACI 318-77, American Concrete Institute.
d.
Manual of Steel Construction, 8th Edition, American Institute of Steel Construction.
e.
Building Code Requirements for Minimum Design Loads in Buildings and other Structures, ANSI N58.1-1972, American National Standard.
2.6 TSC RABITABILITY 2.6.1 General Description The TSC is designed to provide a safe working environment for personnel and equipment during normal and emergency conditions.
It incorporates measures to protect against fire and radiation.
Radiological hazards considered include direct radiation and airborne contaminants. TSC per-sonnel will be protected to the same degree as those in a centrol room designed to General Design Criterion 19 and Standard Review Plzn 6.4.
The design includes particulate / charcoal filtration of inlet air and radiation monitoring equipment, including area radiation monitors aed air-borne radioactivity monitors.
2.6.2 Radiological Protection The TSC shielding and ventilation are designed to limit exposures received by TSC personnel due to occupancy to less than 5 rem whole body, 30 rem skin and 30 rem thyroid (Standard Review Plan 6.4) for the duration of the Design Basis Loss-of-Coolant Accident (DBA-LOCA).
Radiation doses to TSC personnel result from several sources.. While in the TSC, personnel are exposed to beta and gamma radiation from gaseous fission products which enter via the ventilation system after an accident.
i In addition, personnel are subject to gamma exposure due to shine from fission products in the Containment and from fission products present in the atmosphere outside the TSC.
Personnel also are exposed to an inhala-tion thyroid dose from radioiodines which enter via the ventilation system.
Doses to TSC personnel from each major source were calculated as a func-tion of time and were integrated for the total dose.
The calculational methods used in this analysis are similar to those presented in the Trojan FSAR Section 15.5.
The following assumptions were made in this analysis :
Source Terms and Dose Model Assumptions 1.
Fissien Product core inventory is based on core thermal j
power of 3558 MW(t) and time at constant power of 620 days.
6 3
2.
Containment free volume is 2.0 x 10 ft.
Contain-ment leak rate is 0.1 percent / day for the first 24 hr and 0.05 percent / day for the next 30 days.
3.
A total of 100 percent of the noble gas core inventory and 25 percent of the core iodine inventory is assumed to be immediately available for leakage from the Containment. Of 1
the 25 percent of core iodine available to release, 91 percent is in the form of elemental iodine and 9 percent is in the fore, uf methyl iodine.
Elemental iodine spray removal constant is 10.0 hr-I.
Airborne elemental iodine concentrations were assumed to be reduced by the spray until a reduction factor of 100 has been achieved at which time spray removal was terminated.
4.
Containment concrete wall thickness is 3.5 ft for cylin-drical part and 2.5 ft for hemispherical part.
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5.
Dispersion coefficients were calculated in accordance with K. !!urphy and K. Campe method (SRP 6.4):
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2 g (sec/m )
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Time Period 0-8 hr 1.11E-3 8-24 hr 1.11E-3 1-4 day 4.66E-4 4-30 day 3.87E-4 6.
The QAD-F point kernal code was used to calculate the doses resulting from the fission products in the Containment. A semi-infinite cloud dose model per NRC Regulatory Guide 1.4 was used to evaluate the dose from the fission products in the atmosphere outside the TSC and the beta doses inside the TSC to personnel. A finite cloud dose model uns utilized to determine whole body exposure from fission products inside the TSC.
TSC Building Characteristics 1.
Wall thickness is 1 ft concrete.
5 3
2.
Building volume is 1.44 x 10 fg s
3.
Total ventilation filter flow is approximately 2000 cfm; air intake flow rate and air recirculation rate are approxi-mately 1100 cfm and 900 cfm, respectively. Air infiltration rate is considered negligible since the TSC building is maintained at a slight positive pressure and the entrance door to the building is equipped with a vestibule arrangement.
4.
Charcoal bed depth is 4 in. with a filter efficiency of 99%
for elemental and methyl iodines. 1 J
The resulting total 30-day integrated LOCA whole body gamma dose from the above three radiation sources, thyroid and beta / skin doses are presented below.
1 Total 30-Day Integrated Dose (res)
Whole Body i
Ground Level Release Gamma Thyroid Beta Skin 1.
Direct Radiation 0.55 j
2.
Cloud Shine 1.35 3.
Air Intake 0.95
-15.3 23.4 f
TOTAL 2.85 15.3 23.4 These results are within the limits of Standard Review Plan 6.4 and General Design Criterion 19. This analysis is based on approximately 1-f t.-thick concrete which is provided in the TSC structure to ensure adequate shielding ~for the TSC personnel from the first two sources.
An emergency air filter system is also provided to reduce the thyroid dose from the third source of radiation.
In order to ensure adequate radiological protection of TSC personnel, a radiation monitoring _ system is - provided in the TSC for monitoring of both direct radiation and airborne radioactive contaminants. The moni-tors alarm in the TSC upon detection of high radiation conditions. An area radiation monitor will be provided on the first floor of the TSC on which most of the personnel occupancy sill occur.
In addition, a noble f-gas and particulate airborne radioactivity monitor will be provided as well as a means to distinguish the presence or absence of iodines.
Action levels to define when protective measures should be taken i
(including evacuation) will be designated.
2.6.3 TSC Ventilation System Air-conditioning, heating and ventilation are provided in the TSC to ventilate, maintain personnel comfort, prevent entry of airborne
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radioactive contaminants and assure an environment suitable for the operation of the TSC components. Outdoor air is supplied through an emergency air filter, unit during accident conditions.
The design of the heating, ventilation and air-conditioning (HVAC) is in accordance with General Design Criterion 19, as related to providing adequate protection to permit occupancy of the TSC under accident condi-tions. The TSC ventilation system, however, is neither redundant, Seismic Category I, instrumented in the control room nor automatically activated.
The design basis conditions of interior / exterior temperature and relative humidity are:
Interior Exterior Summer maximum temperature 75'F 91'F (dry bulb)
Winter minimum temperature 68'F 14*F (dry bulb)
Relative humidity
<50%
<100%
The HVAC system consists of an air handling unit, a roof mounted air-cooled condensing unit, a kitchen and toilet exhaust fan, an emergency air filter unit with fan, internal recirculation line, related sheet metal ductwork with dampers, diffusers and registers, and electric baseboard heaters. A sketch of the conceptual design of the TSC HVAC is I
presented in Figure 4.
The air handling unit supplies a conditioned mixture of outside and return air to the TSC during normal conditions.
Air from the kitchen and toilet areas is exhausted outdoors by an exhaust fan. The air handling unit consists of a fan, a direct expansion cooling coil, dampers and filters rated for 80 percent efficiency as determined by the National
. Bureau of Standards (NBS) Dust Spot Method on Atmospheric Dust.
The emergency air handling system consists of two portions, an outside air makeup system that operates at approximately 1,100 cfm and an inter-nal recirculation loop for additional removal of airborne iodine that operates at about 900 cfm..
The emergency air filter system consists of a centrifugal fan, dampers, pre-filters, an electric heating coil, a charcoal adsorber and two high efficiency particulate air (HEPA) filters, one upstream snd one downstream of the charcoal adsorber. The pre-filters are rated for 80 percent efficiency as determined by the NBS dust spot test and are of water and fire resistant design in accordance with Underwriters Laboratory (UL) Class 1 requirements. The charcoal adsorber with a bed depth of 4 in. is designed for a flow velocity of 40 fpm to give suffi-cient residence time. The impregnated charcoal used in the adsorber is capable of removing in excess of 99.5 percent of methyl iodine (CH 1) 3 and 99.9 percent of elemental iodine under inlet conditions of 70 percent relative humidity. The HEPA filters are capable of removing at least 99.97 percent of the 0.30 micron or larger particles which impinge on the filter and are of water and fire resistant design (UL Class 1).
The recirculation loop air mixes with the inlet air upstream of the charcoal filter unit.
In the event of an emergency, the emergency air filtration system is manually activated and the unfiltered outside air duct and the kitchen and toilet exhaust ducts are manually isolated. The building will then be maintained at a slight positive pressure to prevent the infiltration of contaminated air.
Appropriate test connections will be provided in the emergency air fil-tration system.
In order to achieve high system perforuance in an abnormal condition, consideration for volcanic ash fallout will also be incorporated in the design of emergency air filtration systems.
2.6.4 TSC Fire Protection System The Fire Protection Systems include equipment for fire detection and suppression. The TSC contains only nonsafety equipment and is not required for safe Plant shutdowr.. A 3-hr. rated fire barrier separates the TSC from the adjacent Demineralizer Building in order to prevent any fire in the TSC from spreading to other areas. The Fire Protection System provided for the TSC conforms to industry-accepted practice.,
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Design criteria for fire prevention, detection and suppresalon. include:
1.
Building construction materials are noncombustible /
limited combustible in accordance with the National Ffre Protection Association (NFPA) Std. 220, except as noted under Item 2.
2.
The flame spread, smoke development and fuel contribution ratings for construction materials including interior finishes are as low as practicable, not exceeding ratings in UL Std. 723, and the American Society for Testing Materials (ASTM) E84.
3.
Fire barriers with a minimum 3-hr. rating are provided to separate the TSC from the remainder of the Plant.
4.
Fire stops are provided for piping and cables where a fire barrier is penetrated. The fire stops have at least the same fire rating as that required for the barrier.
5.
Air filters are designed for UL Class 1.
High efficiency particulate air filters are fire-resistant and conform to UL S td. 586.
6.
Metal furniture is used in the building as much as i
possible.
The fire detection system in the TSC is designed to provide a positive means of determining which area has indicated a fire condition.
Fire detection systems are provided for charcoal filters, computer cabinets,
, communications cabinets and combustible supplies storage areas. The detector selected is the most suitable for the postulated fire at a particular location. Where smoldering electrical insulation is postu-lated, a smoke detector.ansitive to combustion products is used. For charcoal filters, e detection system using rate-compensated thermal detectors are provided.
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The fire detection system is Ussisted in compliance with the requirements of NFPA Std. 72-D.
Fires are alarmed both locally, in the TSC Command f
Center, and in the Plant control roca. > Unique alarms are used for the
~
fire detection system in order' to avoid unnecessary confusion with other alarm 2 i
The TSC fire suppression system inclnde use of water spray, Halon and portable fire extinguishers.
A water spray system with open nozzles is provided to protect the char-coal filter in the emergency ' air filter unit. The water supply will be taken from the existing Plant fire protection water supply system. The system includes a flow control valve, itxed piping, open spray nozzles
, and conforms to the requirements of NFPA Std. 15.
$A'Ha' ion system is provided for the computer cabinets and the communica-I tion equipment. The system,was selected f ar its suppression capabili-ties,minimalcleanuprdkuirements,abilitytoavoidelectricalderating and ninimaletoxic effects on personnel. The system consists of Halon storage containers, fixed piping and, discharge nozzles and conforms to the requirements of NFPA Std.12A.
,i Portable fire extinguishers are loca'ted at strategic locations throughout the building to permit the immediate' ex*leguishment of limited fires.
Halon extinguishers are provided in areas of potential electrical or instrument fire and dry chemicals or water hose cabinets are provided as required for other areas.
Coverage is based on NFPA Std. 10.
The Plant fire brigade is hatified in the event of a fire and provides support for the fire suppression systems described above.
2.7 TSC Communications The TSC is er, nipped with the in-Plant communication system, commercial telephone lines, dedicated telephone communication to State and local EOFs, and a dedicated NRC telephone. This equipment will provide the _
a i
TSC with communication links to the control room, the Operations Support Center (OSC), the Emergency Operations Facility (EOF), the NRC Emergency Opetations Center and Westinghouse.
2.8 TSC Instrumentation and Power Supplies A reliable electric power supply. system'is provided for the normal and emergency operation of the TSC. This supply is not required to be safety
. grade; however, the power systems achieve a functional unavailability j
of 0.01.
In addition, power supplies for the TSC do not degrade the 1
capability or reliability of any safety-related equipment.
T I
Two independent 12.47 kV power supply feeders are dedicated for TSC l
operation. One of the 12.47 kV -transformers is connected to the 12.47 kV j
yard oop (normal supply) and the other to the 12.47 kV Plant bus, H1 (alternate supply). The 12.47 kV yard loop may be aug:1.cnted by an onsite 1500 kVA diesel generator unit which would provide power under emergency conditions or a loss of offsite power.
i A 480-V load center provides primary distribution and is supplied by two identical 12.47 kV/480-V, 500 kVA dry. type transformers, each with an independent connection to the resr :tive 12.47 kV power sources.
No automatic breaker circuits are employed for these connections. Suffi-cient battery capacity is available to provide enough time to manually connect either transformer to the load center.
Subsequent power distri-bution is accomplished through a 480-V/227-V motor control center, 227-V power panels Land 120-V distribution panels.
I A complete uninterruptible power supply (UPS) system with 45-min. bat-tery backup supports the TSC computer. Power conditioning and regulation are accomplished in the inverter portion of the UPS. A static switch provides bypass capability for maintenance and emergency service.
e A separate d-c system is provided and dedicated for Plant and TSC com-i munications along with independent power sources for emergency lighting.
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s 2.9 TSC Technical Data and Data System The data display systems to be provided in the TSC are described in Section 4.
2.10 TSC Records A dedicated room is provided on the first floor of the TSC for records storage. Critical documents to be maintained in the record storage room include Emergency Procedures, FSAR, Technical Specifications, System Descriptions, and General Arrangement and Elementary Schematic Drawings.
3.0 EMERGENCY OPERATIONS FACILITY INTRODUCTION Existing PGE facilities will be utilized for the PGE near-site Emergency Operations Facility (EOF). A primary PGE EOF is located in the Trojan Nuclear Plant Visitors Information Center (VIC) which is approximately 0.5 miles WSW of the Plant. An alterrate PGE EOF is located in St. Helens, Oregon which is approximately 14 miles by road from the Plant and is outside the plume exposure emergency planning zone.
Both primary and alternate EOFs may be supplemented with temporary facilities, such as trailers, if the accident warrants.
Utilization of a primary and alternate EOF ensures the presence of an operational EOF at all times.
If post-accident dose rates permit. travel to and residence at the primary EOF, it can be used providing convenient access to the Plant.
If accident conditions are substantially worse than expected, the alternate EOF can be utilized.
The following sections provide specific design concepts of the PGE EOFs.
In the following description, the term EOF refers to both primary and alternate facilities unless a specific distinction is made.
l 1 -.
O 3.1 EOF FUdCTION States of Oregon and Washington, Columbia County, Oregon and Cowlitz County Washington have individual Emergency Operations Centeri (EOCs).
These EOCs are responsible for directing public protective actions, such as evacuation, and for coordinating overall State and local response to the accident. The State ard county emergency plans will specify in detail the r,ctions of these EOCs.
The PGE EOF provides the following functions in response to an accident:
1.
Serve as a primary location for tha VGE post-accident recovery management center. This function may be replaced by temporary facilities later in the accident recovery.
2.
Advise State and local agencies directly affected by the emergency of recommended protective actions to be taken until the State and Courity EOCs can assume this advisory responsibility for their respective States. This function is expected to be relatively short-term. It is expected that the State and county EOCs will be activated in a time frame comparable to the EOF.
3.
Receive and interprete Plant and offsite monitoring data and transmit data summaries and subsequent protective action recommendations to the State and county EOCs.
Coordinate offsite monitoring and offsite dose assessment.
4.
Coordinate exclusion area evacuation and access control.
Evacuation of the general public off site will be directed from the State and county EOCs.
5.
Account for a',1 Plant and contractor personnel e"acuated to the EOF. _
6.
Serve as an initial assembly point for essential Plant personnel not assigned to the Operational Support Center or the Technical Support Center.
2 The primary EOF is normally the Trojan VIC. The alternate EOF is normally the PGE office in St. Helens, Oregon. The VIC and PGE St. Helens office will be evacuated in a prompt manner when the EOF is activated so that the normal use of the VIC and PGE St. Helens office will not degrade prepared-ness of the EOF to react to an accident or reduce EOF systems reliability.
EOF Functions 2 and 3 above are the only functions directly affecting offsite agencies that require continuity if the primary EOF is evacuated to the alternate EOF. These two functions initially will be performed in the TSC prior to EOF activation. Therefore, these functions can be transferred easily back to the TSC for the short time period (approxi-mately 20 min.) required to travel to the alternate EOF. The alternate EOF will be readied for activation when the primary EOF is activated to facilitate relocation if nacessary.
1 3.2 EOF LOCATION The primary EOF is located approximately 0.5 miles from the Trojan plant in the WSW direction (see Figure 5).
The EOF is outside the security boundary but is close enough to the Plant to ensure that timely face-to-face communications are possible between EOF management personnel and onsite personnel. Maximum travel time from the Plant to the EOF is less than 5 min.
The alternate EOF is located in the PGE office in St. Helens, approxi-mately 14 miles from the Plant. Travel time from the Plant to this EOF is approximately 20 min.
3.3 EOF STAFFING The EOF staff will be described in the Trojan Radiological Emergency i
Response Plan and will include personnel to perform radiological evalations, to interface with offsite officials and to manage offsite PGE resources.
3.4 EOF SIZE Figures 6 and 7 show the conceptual floor plan of the primary and alter-nate EOFs. The EOFs will accommodate at least 35 people, along with a sufficient size of records and data displays. Office space for NRC personnel as well as a sufficient area for at least 20 people for meetings are available in the EOFs as shown in Figures 6 and 7.
3.5 EOF STRUCTURE The Trojan VIC and the PGE office in St. Helens are substantial struc-tures built to withstand adverse conditions reasonably expected in the area. Habitability of the EOF is discussed in Sectica 2.1.6 above.
3.6 EOF HABITABILITY Both primary and alternate EOFs are habitable to the criteria established in NRC Standard Review Plan 6.4.
Although the EOF is relatively close to the Plant, calculated doses to EOF personnel are low because the airection of WSW is the sector of least frequent wind direction and lowest wind persistence. During the period 1972-74 the wind blew toward the WSW only 95 hr. out of 15,973 hr., a frequency of about 0.6 percent.
In addition, for the years 1976 to 1980, the maximum wind persistence in the WSW sector was 3 hr. and the average (50 percent) persistence was much less than I hr.
An analysis was performed to calculate the radiation doses to personnel within the primary EOF as the result of an accident at Trojan. The following assumptions were made in the analysis:
s f
1.
Source term and effluent release assumptions consistent with Standard Review Plan (SRP) 6.4 2.
0.5 percentile atmospheric dispersion factors in the WSW direction at 0.5 miles (700 meters) calculated as described in Regulatory Guide 1.145 (Rev. O. August 1979). The,disper-sion coefficients utilized in the calculation are shown below:
xlQ ***
Time Period m3 0-2 hr 6.0E-5 2-8 hr 3.5E-5 8-24 hr 2.7E-5 1-3 days 1.5E-5 3-30 days 6.5E-6 3.
EOF shielding factor of 0.75 for gamma whole body doseIII.
This f6ctor is typical of a western dwelling with no basement.
4.
EOF ventilation isolation factor of 0.65 for IodineI' This assumption is consistent with the study by Sandia Laboratories for normal buildings.
It is assumed that the building ventilation will be isolated when the radioactive plume is located in the direction of the EOF. This assump-tion is reasonable because of the presence of the meteoro-logical and radiation monitoring equipment.
5.
Exposure duration of 30 days. Occupancy of the EOF was assumed to be that descrihed by K. Murphy and K. Campe (SRP 6.4).
_The results of this analysis are as follows:
0-30 day gamma whole body dose
= 0.53 rem 0-30 day beta skin dose
= 0.56 rem 0-30 day inhalation thyroid dose - 28.77 rem,
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These calculated doses are less than the SRP 6.4 guidelines. Addi-tionally, sufficient potassium iodide (KI) tablets are stored at the EOF for all EOF personnel. KI has been shown to be approximately j
99 percent effective in preventing thyroid uptake of radioiodineI33 The alternate EOF is located outside the evacuation planning zone, and therefore is habitable throughout the accident.
l Permanent radiation monitoring systems will be installed in the primary l
EOF by April 1982. These systems continuously measure radiation dose rates and airborne radioactivity concentrations inside the EOF while the facility is in use.
The monitoring systems include local alarms.
The capability is also provided to distinguish the presence or absence of iodines.
3.7 EOF COMMUNICATIONS Both the primary and alternate EOFs are provided with the following communications equipment:
1.
A dedicated telephone system linking the EOF with the TSC, PGE headquarters and the State and county EOCs.
2.
Commercial telephone lines to State, Federal and county emer-gency centers with a minimum susceptibility to overloading.
3.
PGE telephone system for communications to TSC, Plant areas and PGE headquarters.
4.
Trojan plant Executone system for backup communications to Plant areas and TSC (primary EOF only).
5.
Radio systems (ground and airborne) for communications with ground and airborne field monitoring teams.
6.
PGE Company radio system for communications with security personnel and within the PGE system...
1 7.
Hardcopy data transmission system for data transfer to the NRC and State and county EOCs (later).
8.
NRCEmergencyNotificationsystem(ENS)andHealthPbysics Network (HPN) telephones (later for primary EOF).
9.
Dedicated commercial telephone line for time-sharing computer system used for dose calculations.
3.8 EOF POWER SUPPLIES All instrumentation at the primary EOF are normally powered from the building power supply which in turn is powered from the Plant distribu-tion system. The alternate EOF is powered from the city of St. Helens' power distribution system which is approximately 15 miles from the Plant.
3.9 EOF TECHNICAL DATA AND DATA SYSTEM The TSC Computer System will be utilized to transmit the followiag data to both EOFs froc the Plant:
1.
Effluent monitor readouts.
2.
Area radiation monitor readouts.
3.
Meteorological parameters.
This data will be sufficient to allow the EOF to perform its primary func-tion of providing dose assessments and protective action recommendations to State and county EOCs.
3.10 EOF RECORDS Both EOFs will have ready access to up-to-date Plant records, procedures and emergency plans needed to exercise overall utility resources manage-ment and for recovery management. Additional up-to-date records related -
1 to Trojan, State and local emergency response plar,, radiological records,
~
onsite personnel control, offsite population distribution, and evacuation planning will be readily available in both EOFs under emergency operating conditions.
REFERENCES 1.
Aldrich, D. C.,
et al., Public Protection Strategies for Potential Nuclear Reactor Accidents:
Sheltering Concepts with Existing Public and Private Structures, SAND 77-1725, Sandia Laboratories, Albuquerque, New Mexico (1977).
2.
Aldrich, D. C. and Ericson, D. M., Jr., Public Protection Strategies in the Event of a Nuclear Reactor Accident: Multicompartment Ventilation Model for Shelters, SAND 77-1555, Sandia Laboratories, Albuquerque, New Mexico (1978).
3.
Aldrich, D. C. and Blond, R.
M., Examination of the Use of Potassium Iodide (KI) as an Emergency Protective Measure for Nuclear Reactor Accidents, SAND 80-0981, NUREG/CR-1433 (Draf t), Sandia Laboratories, Albuquerque, New Mexico (1980).
e _
$0 DATA DISPLAYS AND DATA TRANSMISSION SYSTEMS A Westinghouse Technical Support Center Computer System will be utilized to accomplish the functions of data collection, data manipulation and data display for the TSC, the Safety Parameter Display System (SPDS) and the Bypassed and Inoperable Status Indication (BISI) System, and radio-logical effluent data and meteorological data transmission to the Emer-gency Operations Facilitity (EOF).
The system may be used to provide input to the Nuclear Data Link in the future. The design philosophy of the computer system is as described in WCAP-9725 (June 1980) with the exception of the computer hardware configuration. WCAP-9725 has been previously submitted to the NRC by Westinghouse.
4.1 General The Westinghousa TSC Computer System is not a safety system and, there-fore, is not Class IE.
However, the TSC Computer System is designed to provide a high degree of reliability.
Proper isolation devices are provided to maintain physical independence and isolation with safety systems, and to ensure that a failure in the TSC will not degrade the Plant's protection system.
The isolation devices are Class IE, meet the requirements of established criteria consistent with the existing Trojan designs.
The TSC design is such that a failure of any subsystem within the TSC will not significantly degrade the overall TSC functions. The TSC Computer System need not be seismically qualified since it is a non-safety system ar.d is not required to mitigate accidents.
The central processing units for the TSC Computer System are located in the TSC BuiJdlag.
The System utilizes remote I/O racks, located in the Control Building, to transmit input signals to the TSC computer.
Approx-
. imately 20 percent of the input signals are connected via a card extender unit to the existing Plant process computer input racks.
All Class IE sensors are electrically isolated..
4.2. Technical Support Center (TSC)
The TSC Command Center layout is in accordance with WCAP-9725, Figure 2-1.
There are five CRTs, two video hardcopy units and one high speed printer located in the Command Center. Each CRT is mounted on a sliding housing on a table to allow movement of the CRT along the table as well as 360-degree rotation. This allows maximum utilization of each CRT by TSC personnel.
Plant parameter signals are collected, processed and displayed by the TSC Computer System. Plant parameter status displayed in the TSC Command Center allows TSC personnel real-time information regarding the current status of Plant systems. The TSC Computer System stores parameter data for historical analysis and provides the capability to assess Plant parameters independent of Plant personnel actions. A means of logging and trending parameters is available to TSC personnel.
4.3 Safety Parameter Display System (SPDS)
The SPDS monitors Plant parameters important to safety and presents the current stacus of these parameters via CRT graphic displays in the main control room and the TSC Command Center. Two CRTs are located in the main control room to assist the operator in the detection of Plant safety abnormalities. The SPDS is an integral part of the TSC Computer System.
In addition to CRT graphic displays, the status of all Plant parameters collected by the TSC computer system are available to personnel for analysis on a real-time basis.
4.4 Bypassed and Inoperable Status Indication (BISI)
The BISI monitors the current status of the availability of safety sys-tems. The BISI system provides component-level status as well as system status for safety systems.
Two CRTs are located in the main control room 1
and the TSC Command Center. The BISI system is an integral part of the TSC Computer System and allows for manual component-level status by l -
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control room personnel. BISI system CRTs located in the TSC Command Center are monitors only without any manual input capabilities.
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4.5 Emergency Operations Facility (EOF)
The primary function of the EOF is to evaluate the magnitude and effects of radiological dose assessments and recommend protective actions to State and County EOFs. To accomplish these functions, EOF will be provided pertinent radiological and meteorological data through the TSC Computer System. Data available in the EOF are described in Section 3.9.
4.6 Nuclear Data Link (NDL)
The NDL System will condition, process and transmit certain reactor process, radiological and site meteorological data from the Plant to the NRC's Operations Conter located in Bethesda, Maryland. The TSC Computer System will be utilized to provide the Plant data to the NDL System.
Detailed design criteria and functional requirements have not been completed by the NRC yet.
PGE is waiting for additional guidance from the NRC.
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