IR 05000206/1986023

Emergency Response Facilities Appraisal Rept 50-206/86-23 on 860722-25.No Significant Deficiencies or Violations Noted. Major Areas Appraised:Licensee Implementation of Suppl 1 to NUREG-0737 Requirements & Regulations

ML20215N741
Person / Time
Site:	San Onofre
Issue date:	10/30/1986
From:	Fish R, Temple G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:
Shared Package
ML20215N735	List: IR 05000206/1986023 ML20215N734
References
RTR-NUREG-0737, RTR-NUREG-737 50-206-86-23, TAC-46126, NUDOCS 8611070218
Download: ML20215N741 (21)

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U. S. NUCLEAR REGULATORY COMMISSION

REGION V

Report No._50-206/86-23 Docket No. 50-206'

License No. DPR-13 Licensee: Southern California Edison Company P. O. Box 800 2244 Walnut Grove Avenue Rosemead, California 91770 Facility Name: San Onofre' Nuclear Generating Station, Unit 1 Inspection at: San Orofre Site, San Diego County, California Inspection Conducted: July 22-25, 1986 Inspector: hhMt A G. M. Temple, Emergency Preparedness Analyst

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/c/Jo/8L Date S'igned

. Team Members: G.~F. Martin, Research Scientist, Battelle-Pacific Northwest Laboratories (PNL)

W. V. DeMier, Research Scientist, PNL M. K. Lindell, Research Scientist, Battelle-Human Affairs Research Centers D. H. ' Schultz, Reactor Operations Engineer, Comex Corporation

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' Approved by: v F.. Fish, Chief, Emergency Preparedness DatTe Signed Section Summary:

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Inspection on July 22-25, 1986 (Report No. 50-206/86-23)

Areas Inspected: n announced appraisal of the Emergency Response Facilities (ERFs) was conducted using IE Inspection Procedure 82212 to determine,if the licensee has successfully implemented the requirements in Supplement 1 to NUREG-0737 and the regulations. The appraisal covered the Technical Support Center (TSC) and Operations Support Center (OSC), as well as the instrumentation, supplies, and equipment for these facilitie ,

Results: No significant deficiencies or violations of NRC requirements wer .

identifie /

8611070218 861031 PDR ADOCK 05000206 G PDR

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., FOR THE. DETAILED ERF. EVALUATION

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w 1.0 Technical Support Center . ... ... . ...... .... ... 1

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. ' ' Physical Facilities . .. . . .... ..... ....... I g'

1.- l .1 ' Design, Location, and Habitability. . . . . . ... 1

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,1 Size and Layout . . .......... . . .... 2

'1. Equipment and Supplies. . . ............ 2 1. . Communication Systems . ........... ... 3 1. Power Supplies. . . . . . . ...... ...... 3 1.' 1. 6 Conclusio ... . .... ........ . ... 4 1.2 N Information Management. .. . .......... .. .... 5 1. Variables and Parameters. . . . . . . . .... .. 5-1. Foxboro Fox 3-Data Acquisition and Display System. ..... ........ . ... 7 1.2.3_ Health Physics Computer System. . ..... . ...~9 1.2.4- Manual Information Systems and Display Interfaces. .. . ........ . . ...... 9 1. . Reactor Technical Support . ..... . ... . . . - 10 1. Dose Assessment / Source Term . .... .... . . 11 1. Conclusio .... ... ......... .. . 13 1.3 Functional Capability . ... ......... . .... . 14 1. Operations and Control Room Support . . 14

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1. Onshift Dose Assessment . ...... ... ... . 14-1. Conclusion. ... . ... ...... . .. . . . . 15 2.0 Operations Support Cente . .. .... . ........ ... . 15

~ Physical-Facilities . . .. . ... .... ... ... . . 15 2. . Design, Location, and Habitability. . . . . . . . 15 2. Equipment and Supplies. .. ...... ... . . 16 2. Communication . . ... .. .... . .. . . . 16 2. Conclusio . .. . ... . . .... . .. . . . 16

- 2.2 Functional Capability . .. . ... . ..... . .... . 16 2. Staffing and Activation . . . .... ... . . . 16 2. Conclusion. . . . . ... . ..... .... . . 16

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3.0 Exit Interview ~ . . . . .. . .. . ... . . ........ . . 16

Attachment A -~ Persons Contacted i

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DETAILED ERF EVALUATION a

1.0 Technical Support-Center (TSC)

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1.1 Ph'ysical Facilities

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1. Design, Location, and Habitability

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The San Onofre Nuclear Generating Station'(SONGS) Unit 1

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cTSC is-located in the Unit 1 Control and Administration Building, adjacent to the Control Room (CR). .Due to the

, close proximity, the CR can be accessed from the TSC-within a matter of-' seconds. The TSC has been built to the California Uniform Building Code-(Revision 0.67). Gamma shielding for the TSC is provided by only two walls (north and west), both-of which consist of two feet thick concrete. The whole body dose to personnel in the CR and TSC does not meet the requirements of General Design

+ - 5 ' Criterion (GDC)- 19 nor Supplement 1 to NUREG-0737. .Until -

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October 1982. the CR and TSC shared the same heating',

ventilation and air conditioning (HVAC) system. By October 1982, a separate-HVAC system was installed for th '

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TSC in order to accommodate the heat loads associated with

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the number of emergency response personnel ' assigned to the

TSC.and the pcesence of computer equipment. The new TSC '

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HVAC was designed snd installed to be similar to'the CR

. , HVAC. NUREG-0737 Item III.D.3.4,-Control-Room

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, 'Habitabilit'y, has not yet been resolved for SONGS- ,- -

" 3 Associated with this review, an NRC staff inspection was conducted on May 20-22, 1986 and a number of deviations

from current acceptance criteria'related to the CR ventilation system were identified. Since the TSC and the CR have similar HVACs, the deviations associated with

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the CR apply ~to the TSC. 'In order to meet'the

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requirements'of GDC-19, the licensee has taken credit for shift changes (two for.the'TSC), respirator protection

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blocking agents (potassium iodide (KI)). The licensee assumed a protection factor of 20 for the K Based on this' evaluation, the TSC habitability does not meet the requirements of Supplement I to NUREG-0737. However, on October 10, 1986 the licensee submitted proposed upgrades to the Unit 1 CR HVAC system which in numerous cases deviate from NUREG-0737' Item III.D.3.4 and GDC-19 acceptance criteri If the NRR staff concludes that these CR deviations from the GDC-19 criteria are acceptable, then the TSC need not meet the GDC-19 dose criteria as long as it meets the same dose criteria as the CR, except for seismic qualification and automatic initiation of the HVAC. Resolution of the TSC deviations will depend.upon the level a radiological habitabilit found acceptable for the CR by NRR during the ungoing NUREG-0737 Item III.D.3.4 revie .

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1. Size'and Layout

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The TSC ~ consists of approximately 900 square feet of space

.that is divided into four areas. The Emergency, Assessment areat consists of approximately 450 square' feet.of space for-Technical Assessment.and Radiological Assessment. The Emergency Coordinator (EC) area and the Emergency Communication area'are approximately 125-and 150. square feetr respectively. NRC personnel are assigned to.an area +

approximately 175 square feet. Since the amount.of'

available space is quite limited for the number of personnel in the TSC; a significant amount of congestion

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during. full activation would be expected. The space <

limitations are more significant with respect to circulation than for operational space.- The layout tends

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to_ control the impacts:of noise because-of the location of ,

permanent and moveable _ partitions within the TSC. The appraisal staff judged the size _ and layout to be marginal to support its function and consideration should be given to-increasing.available spac +

,1. ' Equipment and Supplies The TSC has=been supplied with an excellent technical

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determined by the users, and consists of all resources

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conceivably needed by the TSC staff to perform their

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function. Copies of 'the Emergency Plan (EP), offsite emergency response. plans, Emergency Plan Implementing-L

Procedures _(EPIPs)',-Administrative Procedures, Operating

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Procedures, Emergency Procedures, System Descriptions, Technical Specifications, and Final Safety Analysis Report (FSAR) are examples of plant reference materials '

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. av'ailable. ' Source-reference' materials.such as engineering handbooks-for electrical / mechanical engineers, steam '

tables, and other.information, are also available. Each TSC manager is provided with a responsibility-specific loose-leaf binder that includes pertinent forms, procedures, telephone lists, etc. Tabs are included to provide easy location of sections within the binde s Drawings are available in.hardcopy form in a stick file of Piping and Instrument Drawings (P& ids) electrical and instrument one-line diagrams, plant-layout diagrams, and others. .In addition to these drawings, a microfiche file contains all the hardcopy drawings, plus numerous architect / engineer and vendor drawings for selected components. A reader printer is installed in the TSC for the fiche fil In^the event any other source material, 4 for example, component technical manuals, is desired, the onsite Corporate Document Management (CDM) Center is a short distance away (3-5 minute walk, outside the protected area)'. The.CDM is computer _ linked to Southern California Edison's (SCE) corporate headquarters for any .

documents not held onsit , . .

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s The.TSC was'noted to;be adequately equipped with other

. materials for accident ~ assessment and management. Locked emergency. lockers, under strict inventory. control, provide radiological equipment storage, and storage for

, administrative supplies such as paper.and hand-held ,

calculators. Pertinent status boards are posted _or ready-

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for'immediate use. Several maps, including. demographic, radiological, and geological types, are.available.-

Onsite, near-site, and large-scale maps are included.

l-The radiological equipment and supplies provided in the Unit 1 TSC are the same as those:provided in the Units 2/3 TSC. This matter was evaluated during the Units 2/3 Emergency Response Facilities (ERF) /.ppraisal conducted i March, 1984, and the findings were documented in Section 1.1.3 of inspection report numbers 50-361/84-08 and

50-362/84-07. This portion of the previous inspection was

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not duplicate . _-Communication Systems The communication systems utilized in the Unit 1 TSC are the same as those used in the Units-2/3 TSC. This area was evaluated during the Units 2/3.ERF Appraisal conducted in March, 1984, and the findings were documented in Section 1.1.4 of inspection report numbers 50-361/84-08 and 50-362/84-07. This portion.of the previous inspection

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was not duplicate ,

1.1.5 - . Power Supplies

- Continuity of power was considered for the' following' major

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Foxb.oro Fox 3 Computer and' peripherals (data

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acquisition system)

Health Physics computer terminal

HVAC

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Normal lightin *- -Eniergency! Lighting

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Communications - telephone, paging, evacuation siren During emergency conditions that result in a reactor trip,

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station power to the two separate 4 Kilovolt (KV) busses (1C and 2C) that power emergency safeguards equipment is provided from two separate offsite 220 KV source In the event of loss of both of these feeders, the station diesel generators (two) may power busses IC and 2 .

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The data acquisition system (Fox 3)!forzthe Unitf1 TSC.is

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powered from a 120 VAC Uninterruptible Power Supply (UPS).

Severalf peripherals, located in the TSC, are 'not <xt the UPS; restoration'of site power from either station emergency diesel restores power to the peripherals through wall plugs that are clearly marked and keyed for such

'se rvice. The Health Physics Computer System (HPCS)

!- central processor and peripherals', located in the Units-2/3 TSC, are powered from an UPS. The Unit 1 HPCS TSC-

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terminal and: printer are powered in a similar fashion to the Fox 3 peripherals,; distinctive wall plugs capable of

being powered from station diesels. The.TSC INAC is powered from Motor Control' Center (MCC) 3A. On a loss of power, MCC 3A islde-energized, but may be restored from offsite or the station diesel generators, as appropriate.

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. Normal' lighting to the TSC is provided.through a transformer from the 4 KV 2C bus, thus the station diesel L

, generators may provide power in'the event of a reactor trip and/or a loss of offsite power. The TSC flxed emergency lighting consists of one wall-mounted-

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eight-hour, battery. powered emergency lighting pack (two j; '

lights aimed at the two ingress / egress routes). .It is

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recommended that appropriate relay activated, battery

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powered emergency lighting be installed in the TSC to 4_ permit:the TSC to maintain its function in the event of a

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. , ; loss of all onsite power. Allzcommunications systems are

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capable of being powered from the diesel generators; many

, systems, such'as the Emergency Notification System (ENS)

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.and Health Physics Network (HPN), are also provided with an UP ,

1.1.6- Conclusion

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, . Based on the' findings in.Section 1.1, this portion of'the U- '

_ licensee's program meets the requirements of Supplement 1

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to-NUREG-0737. However, the Region intends to track the

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-following open item:

w Pending resolution of.the CR habitability issue,

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evaluate the adequacy of the TSC habitabilit (86-23-01)-(Section 1.1.1)

The following items are suggested for improving your

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program:

, (1) Provide the.TSC with additional emergency lightin (Section 1.1.5)

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(2) Enlarge the size and impr~ove the layout of the'TSC to

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accommodate assigned personnel (Section 1.l.2)

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C Information Management -

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-1. Variables and~ Parameter '~

The' licensee h'asJinstalled two computer _ systems to provid l , , the TSC staff ~with an emergency data acquisition system to determine the. safety status of,the plant and the potential

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for radiation' exposures to the_public during plant-

. emergencies. ;The Foxboro Fox 3 makes use of approximately

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. 57 variables'and plant, status. indicators to provide a

. continuous .on-line sub-set of the Regulatory Guide (RG)

. 1.97,, Revision 2, parameters to the TSC. The HPCS provides input of meteorological, area, and process radiation monitors .for accident assessment- and dose projectio ~

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As indicated'above, the parameters provided to the TSC are

a sub-set of RG 1.97, Revision 2, that SONGS 1.has selected as the critical plant parameters. Correspondence dated December 16, 1985, has.been submitted from SCE to NRR concerning the RG 1.97 review conducted in accordance-

- with Supplement 1 to NUREG-0737. A resolution is expected j- from NRR by approximately June, 1987. The appraisal team noted numerous deviations between RG 1.97 guidance and

, the submittal of SCE. Complicating the evaluation of the

submittal is the ~ fact- that SCE performed the evaluation by l first_ determining the SONGS 1 variables, specifically.

I required for post-accident operations, in accordance with Emergency Operating Instructions (E0Is) or EPIPs. Then, SONGS ~1 instruments were identified which fulfilled.the E0I/EPIP step. These identified instruments were then compared to RG 1.97, Revision 2, for design and qualification requirements. Discrepancies were

- identified, justified, and. recommendations formulate Thus, the submittal lists deviations between SONGS 1 instruments available, and the SONGS'l list'of required instruments derived from a task analysis of procedure As an example of this problem, Quench Tank Level',

Temperature, and Pressure are listed in-RG 1.97, Revision 2,-Table ?, as Type D variables. Table 5-2, " Comparison-of SONGS 1 Site-Specific Variables and RG 1.97 Generic Variables" of the SCE submittal lists these variables as

"Not Included" and "Not required in response to accident

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condition," meaning, presumably, that observation of these parameters is not required by an E01 or an EPI In fact, all three parameters are available in the C Notwithstanding the above situation, the appraisal team evaluated parameters available to the Fox 3 data acquisition system,.using RG 1.97, Revision 2, as a basia for a minimum acceptable data set in the TSC. The team noted several missing parameters that would be necessary for prompt and adequate TSC staff evaluation of various hypothetical accident scenarios. For example:

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Type B Variable - Reactivity Control: ' Neutron Flux

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oCR - Source, Intermediate and. Power Range: available TSC - Source, Power Rangei available

/ SCE. technical personnel stated that. intermediate

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power range' instrumentation was not necessary to adequately monitor core reactivity conditions in the

LTSC under, e.g., anticipated transient without scram p c(ATWS) conditions. A callLto the CR could be made to.

j determine power level.

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Type B Variable - Reactivity control: Control Rod :

Position

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CR.- Individual Rod Position: available TSC' " Reactor Tripped / Normal": 'available (not rod position).

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Rod position'could be determined by calling the CR.'

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Type D Variable - Residual Heat-Removal'(RHR) or

, Decay Heat-Removal System: RHR System Flow

~RHR' Heat Exchanger Outlet

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. Temperature CR - Availability unknown TSC - Not available

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SCE stated that these parameters are " Required for' cold shutdown. Not. required in. response to accident conditions.", thus, the' availability of-the parameters in the CR was not' addressed in the RG11.97 review. The appraisal team noted, however,: that 'the system description

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procedure SD-S01-320, " Residual Heat Removal' System",

listed as the third function under " Functions / Design Basis", "To provide 1on term emergency core cooling capability...following a loss of coolant accident".

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Type D Variable - Safety Injection System: Flow in High Pressure Injection (HPI) System

CR - Available

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TSC; ,Available SCE stated in th'eir RG 1.97 review (submittal) that the

1 variables,are "Not Included" and "NA to SONGS". A review'

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- - of the feedwater system description indicates that the 2 a feedwater system functions as the equivalent of a HPI

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. system !and 'is ". . .used for safety. injection operations". This latter example particularly illustrates the confusing

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. Numerous other examples were noted to be absent from the

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. data set'of tFe TSC. In general, no alternative h parameters were made available to substitute for the 4 . missing parameters of RG.I.97, except the isolated case m 'such as " rod position" noted above. The one other simila example noted-by the appraisal team concerned the Type B Variable for " Maintaining Containment Integrity:

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Containment Isolation Valve Position". Instead of the

" Closed not-closed" parameter of RG 1.97, the Fox 3 displays " Containment Isolation Initiation" as a bistable state; thus, the TSC.would be unable to determine the status of containment isolation valves without consulting the C Although the Emergency Operations Facility (EOF) was not a portion of this appraisal, the appraisal team reviewed the real-time. data acquisition system available to the EOF staff from Unit 1. It was noted that the E0F "Emergtacy Support Organization Procedures Manual" lists responsibilities similar to the TSC staff; a Technical Group is stationed in.the EOF with various engineering disciplines to " Assist in assessing the operations related mitigation efforts being performed by the CR and TSC personnel". The appraisal team determined that fourteen (14) preselected parameters of the data set available to the Fox 3 acquisition system are transmitted to the EOF, at one minute intervals, on a printe The fourteen parameters represent the maximum number of variables which can be transmitted at any one time. Re-structuring the data set to include different variables for a different accident sequence requires approximately ten (10) minutes; the appraisal team noted that only one individual on site is capable of performing the configuration change. Based on the information provided, the appraisal team concluded that upon completion of NRR's evaluation of the SCE submittal concerning RG 1.97 variables, the licensee should assure that the appropriate variables essential for the performance of TSC and EOF functions are made available on a real-time basis, in accordance with NUREG-0737, Supplement .2.2 Foxboro Fox 3 - Data Acquisition and Display System The Foxboro Fox 3 data acquisition system collects and stores safety-related data for the plant and provides a mechanism for retrieving and displaying the data in any of several concise formats. The system consists of: a 16-bit microprocessor (CPU) with 64 kilobytes (K) of memory, an operator's console (color monitor with special keyboard), two hard-disk drives (one of which serves as a spare), two floppy disk drives, two alarm typers, a logging typer, a slave color monitor, a black and white cathode ray tube (CRT) " status" terminal, a video copier (black and white), a Foxboro universal field multiplexer (UFM), and a Foxboro SPEC 200 isolation substation. All of the Fox 3 equipment is located in the TSC except for one alarm typer located in the EOF. An acoustic coupler and dial-up (300 baud) phnne line provide the data link between the Fox 3 and the EOF type . . . ._ _ __ ._ _ . . .. .- _ . . _ _ . _ . . . .

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The CPU,hard-disk drives, alarm printer, slave color monitor, and UFM are on an UPS to assure that data'

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collection,- data ' storage, and alarm printing functions -

will not be interrupted if normal power to the Fox 3 is z

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lost. The operability of system peripheral. devices, such

, g as types, disk drives and operator's console, is automatically. checked at_one minute intervals. . Detection of an " inoperable" state for a device causes a reconnect ,

atteapt'to'be initiated. tor that device. -If the reconnect

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attemptiis-not' successful, then the device state is signalled by. an audible " SOS alarm. Also, a display on -

the status.CRT-shows what device is inoperable and provides instructions for obtaining help'in restarting the devic '

Fifty-seven analog and digital' inputs are currently being

. monitored at scan rates of once per second for critical ,

, :F parameters and once every five seconds for non-critical ones. Transmission rate between UFM and CPU is 50,000

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baud. Although the UFM has. capacity for 192 inputs, the actual. number that can be processed by the Fox 3 is much-smaller.due to the limitations imposed by the 64 K memory

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for the CPU. Current operation with 57 inputs leaves less

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ithan 2 K of free memory. If.it turns out that more:

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variables need to be brought'into the Fox 3 system, the existing memory may not be adequate. Adequate disk

storage for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of data from 57 inputs at one minute:

intervals is provided by the Foxboro SPEC 200 isolation

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subsystem. .All SPEC 200 hardware is qualified for class i y

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1-E nuclear service, and conforms to IEEE 323-74 and IEEE

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344-75. The. isolation ~ subsystem converts thermocouple, resistance temperature detector, milliampere and millivolt

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L signals into'0-10 volt signals through a

.," transformer-coupled (E/E) isolation device before passing

,,them~on to.the UFM as non-lE signals. Dry contact inputs

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are isolated through a Foxboro electronic contact output U ,. C isolatok. ~Both isolation devices are designed to

, withstand.up'to 600 VAC at.the device output. Cyclic

' redundancy checking and software checks, on range, are-

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CR'personnelicompare the values displayed in the CR to

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, :those displayed by the Fox 3 on the slave CRT which is

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mounted for viewing from the CR. The Fox 3 system was .

s'bjected u to pre-operational testing (TP-TMI-2.2.2b-5) and post-operational testing (TP-TMI-2.2.2b-10). Loop inputs to Fox 3 were tested in accordance with " Surveillance

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Procedure for Reactor Protection Systems, Loop inputs to

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" Fox 3"., S01-II- System maintenance requirements over the past five years have been minimal, indicating good

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system reliability.

! In addition to the. collection and display of measured 4 variables, the Fox 3 system supports a program for

{ predicting Xenon (Xe) reactivity. This program is always i ,

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executed ,at a priority lower than that assigned to any of -

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the safety-function programs, thereby assuring that the Xe reactivity prediction task will not adversely affect the

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primary function of.the syste . Health Physics Computer System The'Hewlett-Packard model HP-9845 computer, located:in the

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TSC,- is identical .in configuration to the HP-9845s in the EOF which were covered in the ERF Appraisal of SONGS 2/3 in. March, < 1984.-

1. Manual Information Systems and Display Interfaces Plant systems, meteorological and radiological data can be collected in the CR by a dedicated _ communicator who can record the data on a preformatted form and transmit these

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. data'to the;TSC. Data received in'the TSC'can.be recorded and posted on the' principal status board. The data are reviewed by trained enginee's r who screen variable values-for consistency with previous readings of that variable

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and for consistency with the values of other. variable Each of-the principal areas in.the TSC contain large scale di~ splay boards. The Technical Assessment area contains a plant status board that has predetermined variable names, each of.which is labeled in terms of units-of measurement and time of observation. The Radiological-Assessment _ area has wall mounted Emergency Planning Zone.(EPZ) maps that are mounted on rollers. The EC area ~has' a blank marker board and a Notification Status Board used to monitor the required and actual times for nctifications transmitted to local agencies verbally (within 15 minutes), by teletype (within 30 minutes) and to the NRC (within 60 minutes).

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The Emergency Communications area has'a wall; mounted display that indicates the alternative circuits that can bc used to transmit a meccage'to_a decired dentinatio ~ ~

The TSC also contains a Foxboro Fox 3 computer that is J '

used to display analog and digital plant status inputs.

The Fox 3 data displays are' hierarchically organized and l_ can be accessed from a main menu. However, there are

!- displays that contain' incomplete or potentially misleading information. User prompts should more completely describe

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, the input format for.any data entry inputs, such as

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_ starting. hours for trends, and indicate the default value-for, that input parameter. Potentially misleading displays

'should be fixed or eliminated. Historical trends, for

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example,'should automatically stop at.the present' hour;

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'e.g. , a 12:00 -request for a 10-minute increment plot starting at 1:00 should only contain seven points. The

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core exit thermocouple. data that are curiently unavailable j 0 -

should'be blanke'd out or otherwise converted to nonnumeric i , characters.

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- User documentationLfor the Fox 3-is not available at the terminal:and what documentation exists is insufficient for the novice or infrequent. user. 'There should be-documentation available~in the TSC at all-times. The

= primary copy'should be located in a drawer at the terminal and a-backup located in the emergency kit or other secure

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location. The documentation should describe the' control keys and principal displays. -The description of each display should list the exact keyfentry sequence' required

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to bring up.the displa .2.5' Reactor Technical Support

. Most parameters available to the TSC electronic data

. acquisition system-(Fox 3) may be trended in multipl ways, with on-screen or on printer presentation In addition, a screen print function permits a representation on a video hard-copier. 'Most inadequacies in' trending and forecasting capability are a function of; missing. variables as an input'to the. Fox 3. Further? enhancements to the Fox 3 may be limited by memory size, however, the appraisal team noted that'little, if any, correlation exists between" abnormal thermal / hydraulic data and Emergency Action-levelsi(EALs). Thus, limited assistance is provided for the -TSC staff in recognizing changing emergency

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classifications due to plant variables exceeding EALs. It is recommended-that alerting. functions,'such as a flashing

value, for plant parameters correlated to EALs be provided

,

where possible. Only one pre-calculated correlation of a

.

plant parameter to accidentEconditions was noted by the t . -,

-

appraisal team. _ Chemistry procedure S0123-III-8.8,

'!-

" Alternate Methods of Post-Accident Parameter Sampling",

includes a graph depicting the' extent of fuel damage as a function of containment Hi-Range Rad Monitor dose rat ,

.

It was noted that a GAP activity release or 10% fuel damage condition at' shutdown results in a containment dose

^

,

3 -

,- -rate of 10 E+2 R/hr.- Various accident studies auch an

, >

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WASH-1400 indicate an-approximate containment dose rate-3 '

'

, _ for that magnitude of fuel' damage as 10 E+5 R/hr. The

?' ' licensee'should ensure that the plant specific fuel

.

+

damage / dose rate correlation actually differs from i - '-

theoretical studies by three orders of magnitude, due to

'- '

plant specific' conditions, such as sensor configuration,

?> etc., and is~ not, in fact, an erro It is also recommended'that the-licensee ensures to the maximum

, extent 1possible, that other pre-calculated graphs and charts are prepared to minimize calculational time during

, accident transients. For example, core damage as a l_

-

function of coolant chemistry'(specific isotopec), or dose

'

rates outside containment as a function of dose rates in containment may be pre-calculated.

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. 1.2.6' Dose A'ssessment/ Source Term

;Th.ough discussions with licensee personnel and review of

appropriate documentation, it was determined that many of-the dose.. assessment. capabilities of SONGS UnitL1 are

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identical to those'in use at Units 2/3. Since these

'

. ..

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~ capabilities were evaluated during the SONGS-Units 2/3 ERF

,

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,-appraisal conducted in March,~ 1984, where.it was edetermined-that no significant changes.had occurred, thos . portions of the previous inspection were.not duplicate '

'

The following areas of the. licensee's dose assessment

, ' program have been determined to be essentially the same'as ,

,.

'

.previously evaluated during the-Units 2/3 inspection; computerized dose assessment using the Emergency Assessment and Response System (EARS) computer program, "

'

' meteorology,.and the calculational methodology used in manual dose assessment procedure EPIP S01-VIII-40.100,

" Source Term and Dose Assessment". For furthe ;; ,

information regarding these areas, please refer to Section j_ 1.2.5 of inspection' report numbers 50-361/84-08 and

'

50-362/84-0 ^

Following an accident, the primary release point through which releases of airborne' radioactivity could occur is

,

the plant ~ vent. Air is. exhausted by the-Sphere Purge and Exhaust System fans;from the-reactor containment, the

' Reactor Auxiliary Building (RAB), .and the Fuel. Storage .

-

' Building.through high efficiency filter banks. Also,

' feeding into the plant vent is the condenser air ejector L . exhaust. ..The.following 10 gaseous effluent monitors are.

[ available for source term evaluation:

l

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Detector- 'Name Type ' Range Detected R-l211' Containment Sphere Scintillator' 10El-10E4 Iodine-131

"

and Stack Particulate

'

(SC)' Counts per: (I-131)

. Monitor' ' Minute (cpm)

,

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10El-10E6 cpm- Cobalt-60 (Co-60)

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Contai at Sphere Geiger-

~

. R-1212 10El-10E4 cpm Xe-133-

, and' Stack Particulate Mueller (GM) 10El-10E6 cpm Krypton-85^

.

Monitor ,

(Kr-85)

R-1214l , Stack Gas Monitor GM 10El-10E4 cpm Xe-133 <

10El-10E4 cpm Kr-85 i

R-1215 Condenser Air Ejector GM 10El-10E4 cpm Xe-133-Gas Monitor- '10El-10E6 cpm Kr-85

,

R-1219 Stack Gas Monitor SC 10El-10E7 cpm Beta l R-1220 Stack Gas Monitor SC 10El-10E7 cpm Beta i

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R-1221- : Stack Iodine Monitor SC 10El-10E7 cpm I-131

' R-1254 ' Wide Range Gas Monitor SC 10E-7-10E-l' Xe-131 Curies / Cubic Centimeter (Ci/cc)

, Cadmium- 10E-4-10E2 Ci/cc Telerium 10E-1-10E5 Ci/cc (Cd-Te)

R-1256A&B East Steam Dump Header GM 10E-1-10E4 mr/hr Gamma Monitor Ion Chamber 10E-1-10E4 R/hr Gamma R-1258A&B . West Steam Dump Header GM 10E-1-10E4 mr/hr Gamma Monitor Ion Chamber 10E-1-10E4.R/hr Gamma Other key post-accident monitoring capabilities are provided by the containment high range monitors and the

,

Post-accident Sampling System (PASS). The containment

- high range monitors (R-1255 and R-1257) are two physically separate, redundant channels. Each channel consists of an ion chamber which measures gamma radiation inside containment from 1 to 10E8 R/hr. The PASS provides a-means to remotely collect reactor coolant and containment atmosphere samples following an accident, remotely indicates results'of chemical analyses of these samples, and dilutes the samples for further analysis. Samples are

~

remotely analyzed via an intrinsic Germanium (Ge)

detecto EPIP S01-VIII-40.100 provides a procedure for calculating source terms from available information sources. This procedure. provides for source term determination from PASS

~

results, Stack Monitors R-1254 and R-1219, Main Steam Line Monitors R-1256A&B'and R-1258A&B, Containment High Range-Monitors R-1255 and R-1257, and from field monitoring results. The appraisal staff reviewed this procedure and noted that: (1) there were no references to other related procedures or helpful sources of information; (2) the table of contents did not list Attachment 4,

" Determination of Source Term Based on Main Steam Line Monitor Reading 24 Hours After Shutdown";-(3) the body of the procedure does not mention the use of Attachment 4; (4) Section C.2 incorrectly lists the order of accuracy for monitor / methods to be used for source term calculation ,

and~is inconsistent with the order listed in the note to

' ,

Attachment 1 and Section C.2 of S023-VII-40.100; and (5)

'Section C.2 and Attachment I reference and use the Emergency Radiation Monitoring System,(ERMS) which consists of monitors R-1250, R-1251, and R-1252 which are no longer in servic Procedure S0123-III-8.8, " Alternate Methods of Post-Accident Parameter Sampling", provides alternate means of

~

- determining parameters normally obtained through PASS, if 4 ,

,

s e

.

the system is inoperable. Contained in this procedure is a graph for estimating percent failed fuel from containment high range monitor reading TSC dose assessment personnel have access to effluent and area radiation monitor readings from the CR or from the HPCS. A review of the monitors polled by the EARS program revealed that data is still being obtained from the ERMS monitors even through they are no longer being maintaine This error could create serious problems with dose assessment if these data were used during an emergency since this system is still referenced in the EPIPs.

1.2.7 Conclusion Based on the findings in Section 1.2, this portion of the licensee's program meets the requirements of Supplement 1 to NUREG-0737. However, the Region intends to track the following open items: Pending completion of NRR's evaluation of SCE's RG 1.97 submittal, assure that the appropriate variables, essential for the performance of TSC and E0F functions, are made available on a real-time basi (86-23-02) (Section 1.2.1) Ensure that the plant specific fuel damage / dose rate correlation actually differs from generic accident studies by three orders of magnitude, as opposed to being an erro (86-23-03) (Section 1.2.5) Remove the ERMS monitors as input to the EARS computer and all other mention or use of the ERMS monitors (R-1250, R-1251 and R-1252) from the EPIP (86-23-04) (Section 1.2.6)

The following items are suggested for improving your program: j (1) Determine the appropriate course of action if the existing memory in the Fox 3 is not adequate to support additional variable (Section 1.2.2)

(2) Eliminate the Fox 3 displays that contain incomplete or potentially misleading information (e.g.,

historical trends, core exit thermocouple data, etc.). (Section 1.2.4)

(3) Provide user documentation for the Fox 3 that is sufficient for the novice or infrequent user and make this documentation available in the TSC at all time (Section 1.2.4)

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'(4) Where possible, provide alerting functions for plant f parameters. correlated to.EAL (Section 1.2.5) ~

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(5).'To the maximum extent possible, prepare c . pre-calculated graphs. and charts to minimize calculational time during-accident transient (Section 1.2.5)

-(6) EPIP S01-VIII-40.100.could be improved in_the

.following areas:

'

-(i) Include references to related procedures or helpful sources.of information (e.g.,

S0123-III-8.8, " Alternate Methods of Post-Accident Sampling, and-SONGS 1 Emergency

'

Calculation Manual").

(ii) List' Attachment 4, " Determination of Source Term Based on Main Steam Line Monitor Reading-24 Hours After Shutdown", in the table of content (iii) Include a description of Attachment 4, and instruction on when it should be used, in the-body of the procedur (iv) Correct Section C.2.which lists-the order of

accuracy for monitors / methods to be used to calculate-source terms and make it consistent:

with the note on Attachment I and Section of S023-VIII-40.10 .

1.3- Functional Capability

'1. Operations and Control Room Support ,

The functional: capability of the TSC was' evaluated by presenting an.NRC' developed accident scenario to key

'

members of the licensee's staff normally assigned to the TSC during an emergency.: Licensee personnel responded to

the~ postulated circumstances by describing their actions, and demonstrating how the equipment and supplies available in the.TSC would be used. The evaluation demonstrate that the TSC would be adequately staffed and capable of performing its as' signed function "

.

1. Onshift Dose Assessment During the appraisal, a_ walk-through demonstration of th TSC functions and capabilities was conducted. The walk-

.through scenario was based on a loss of all offsite power and the accident sequence was allowed to progress to the point of significant fuel damage (1-10%). As,a part of this walk-through, source term generation, manual dose

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'as essment',.and computerized dose assessment were

' demonstrated. Source terms.and dose assessments were calculated for containment leakage conditions and fo releases through the. plant stack using real-time

. meteorology. TSC personnel demonstrated the ability ~to perform these calculations' in a timely and accurate-manne During the walk-through, members of the TSC technical team performing source ~ term calculations were observed using calculational worksheets and information which did not appear .to originate from a station procedure or documen Discussions with licensee personnel determined that the binder containing these aids, called the SONGS 1 Emergency Calculational Manual, was compiled by station chemistry-personnel as a tool for the TSC technical team. The inspector was informed that the manual was an uncontrolled document, not subject to periodic review and update, and not covered during trainin . Conclusion Based on the findings in Section 1.3, this portion of the licensee's program meets the ' requirements of Supplement I to NUREG-0737. However, the following items are suggested for improving your. program: Make-the SONGS 1 Emergency Calculational Manual part of the controlled TSCLdocumentation and ensure that it is periodically reviewed and update . Include instruction on the contents and use of the SONGS 1 Emergency Calculational Manual in trainin . Ensure that other appropriate documents reference the SONGS-1 Emergency Calculational Manual (e.g.,.

'

~

Chemistry Leader's Manual, Health Physics (HP).

Leader's Manual, 501-VIII-40.100).

L 2.0 Operations Support Center-(OSC)

2.1 Physical Facilities

>

2. Design, Location, and Habitability The SONGS Unit 1 OSC is located on the second floor (35 foot. level) of the new Unit 1 Control and Administration Building. It should be noted that this is~a new location for the Unit 1 OSC. The new OSC was addressed in Details Section 3 of Inspection Report Numbers 50-206/86-01, 50-361/86-01, and 50-362/86-01. The OSC is adequate in size and layout to support its intended function. The OSC does not have any special shielding or ventilation system;

'

'

however, provisions, including procedures and

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instrumentation, have been made to monitor habitability.in

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'the'0SC. If radiological or other conditions necessitate

"

the abandonment of>the Unit 1 OSC, the Units 2/3~0SC.has

<

been designated as the; alternate'0S , .-

_

,

._2; ^ Equipment and Supplies The' equipment and supplies provided in the Unit 1 OSC are

-

'the same as those provided in the Units 2/3 OSC. This matter was evaluated during the Units 2/3 ERF Appraisal _

conducted in March, 1984, and the findings were documented in Section 3.1.2 of inspection report numbers 50-361/84-08 and 50-362/84-07. This portion of the previous inspection was not duplicated, however, the appraisal team verified that the appropriate items were in plac . Communications The communication systems utilized in the Unit 1 OSC are

~

the same as those used in the Units 2/3 OSC. This area was evaluated during the Units 2/3 ERF Appraisal and the findings were documented in Section 3.1.3 of inspection report numbers 50-361/84-08 and 50-362/84-07. ~This portion of the previous inspection was not duplicate . conclusion Based on'the findings in Section 2.1, this portion of'the licensee's program meets the requirements of Supplement I to KUREG-073 .2 Functional Capability 2. Staffing and Activation

.

The Unit 1 OSC functions in the same manner as the Units 2/3 OSC. Personnel staffing, organization and activation are the same. This a.rea was evaluated during the Units

-

2/3 ERF appraisal and the findings were documented in

, Sections 3.2.1 and 3.2.2 of inspection report numbers-50-361/84-08 and 50-362/84-07. This portion of the

previous inspection was not duplicate .

2. Conclusion

'

, Based on the findings in Section 2.2, this portion of the (

<

licensee's program meets the requirements of Supplement 1 to NUREG-073 .0 - Exit Interview

~'

'An exit interview was held with the licensee on July 25, 1986, for the

'

purpose of discussing the preliminary findings of the appraisal. Those

~

licensee personnel who attended the meeting have been identified in

?

.

'

17 -

,> l Attachment'A to this report. The licensee was informed that no

- significant deficiencies or violations of NRC requirements were identified during the appraisal. - The NRC Team Leader addressed each of the open items and improvement items described in the body of this repor ,

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ATTACHMENT A-

'

. . Persons Contacted

'*L~. Bennett; Nuclear Safety and Licensing Enginee'r

• D. Bennette, Supervisor, Station Emergency Preparedness

- Bray, Health Physics Enginee *J. Brooks, Health Physics Engineer K.. Brooks,' Foreman, Health. Physics,-Unit 1 R. Bush, Senior Training' Consultant-(Non-SCE)

G.'Buxton, Station Engineer C. Chiu, Assistant Manager, Station Technical D. Duran, Health Physics Engineer ,

D. Eckhart, Station Engineer J. Firoved, Emergency Preparedness Engineer

,

.K..Flynn, Station Engineer-

'T.~ Head, Computer Engineer-

- Helm,. Effluent Engineer M.'Herschthal, Station Engineer S. Hetrick, Station Engineer

• S. Hunn,- Supervisor, . Corporate Document Management (Operations)

~

.

• J. Ibarra, Nuclect Safety and Licensing Engineer T. Jackson, Instrument and Contro1' Engineer P. Knapp, Manager, Health Physics J. Kroeger, Nuclear Training Instructor

• A.'Llorens, Nuclear. Safety and Licensing Engineer J. Madigan,: Supervisor, Health Physics, Unit 1 A. Melville, Supervisor, Corporate Document Management

.R. Messeder, Health Physics Engineer S. Olofsson, Associate Emergency Preparedness Specialist

• D. Peacor, Manager, Station Emergency Preparedness N. Pillutla, Controls Engineer-

• J. Reilly, Manager, Station Technical

• R. Rice, Project Engineer-

' S. Schifield, Health Physics Ecgineer M. Tomlinsen, Computer Operato .J. Vandenbroek, Station Engineer J. Walderhaug, Computer Engineer R. Waldo, Supervisor, Computers -

• H. Wharton,' Deputy Station' Manager

• Denotes those present at the exit interview on July 25, 198 Perscas Present at July 25, 1986, Exit Interview Only R. Beatty, Liaison,' Station Security C. Bostrom, Administrator,. Health Physics / Chemistry Training C. Couser, Compliance. Engineer

- -

D. Dack, Quality Assurance Engineer P. Dooley, Supervisor, Emergency Planning H. Morgan, Station Manager J. Rainsberry, Supervising Engineer R.-Reed, Associate Emergency Planning Specialist

_

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_ _ _ _ _- _ _ _ ___._. __

. . .. . . . .. . - -. . . ~ . . - . . - . -

I

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R. Reiss,-Quality' Assurance Engineer G. Robinson, Supervisor, Security Operations

~ R. Santosuosso, Assistant Manager, Maintenance'

D.-Schone, Manager, Site Quality Assurance

.

B. Smith, Supervisor, Facility Planning J. Yann, Manager, Nuclear Project Engineering i

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