ML19250C008

Responds to 790913 Request for Commitments to Meet Requirements of NUREG-0578.Forwards Response Delineating Plant Design & Improvements in Emergency Preparedness

ML19250C008
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	10/17/1979
From:	Goodwin C PORTLAND GENERAL ELECTRIC CO.
To:	Eisenhut D Office of Nuclear Reactor Regulation
References
RTR-NUREG-0578, RTR-NUREG-578 TAC-10558, TAC-30680, TAC-30681, TAC-30682, TAC-30683, TAC-30684, TAC-30685, TAC-30686, TAC-30687, TAC-30688, TAC-30689, TAC-30690, TAC-30691, TAC-30692, TAC-30693, NUDOCS 7911060386
Download: ML19250C008 (64)

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October 17, 1979 Trojan Nuclear Plant Docket 50-344 License h7F-1 Mr. Darrell G. Eisenhut Acting Director Division of Operating Reactors U.S. Nuclear Regulatory Commission Washington, D. C. 20555

Dear Sir:

On September 13, 1979, you issued a letter to all operating nuclear power plants requesting, within 30 days of receipt of the letter, com-mitments to meet the hutEG-0578 requirements and the near-term require-ments for improving e=ergency preparedness in accordance with the i=plementation schedules specified in your letter. We have reviewed NUREG-0578 and the NRC guidance for emergency preparedness on its implications to the Trojan plant. is PCE's response to your request, which delineates current Trojan plant design and constitutes a PCE commitment for implementation in areas of the NUREG-0578 recom=endations. Please note that the NRC positions in the attached response are direct quotations from NUREG-0578, which were referenced in your September 13, 1979 letter. PGE's response on improvements in emergency preparedness is included in Attach =ent 2.

Sincerely, C. Goodwin, Jr.

Assistant Vice President Thermal Plant Operation and Maintenance CG/ /mg/4kk50A5 Attachment c:

Mr. Lynn Frank, Director State of Oregon Department of Energy

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k Mr. A. Schwencer, Chief g

Operating Reactors Branch #1 Division of Operating Reactors 1276 221 Mr. R. H. Engelken, Director Nuclear Regulatory Commission Region V

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ATTACTENT 1 Response to D. C. Eisenhut Letter of Secte=ber 13, 1979 on Nt: REC-0578 Short-Term Recot=endations 1276 222

Para 1 of 49 Section 2.1.1 - Emergency Power Supply Requirements for th', Pressurizer Heaters, Power-Operated Relief Valves and Block Valves, and Pressurizer Level Indicators in PWR:

NRC Position 1 Pressurizer Heater Power Supply 1.

The pressurizer heater power supply design shall provide the capability to supply, from either the offsite power sourca or the emergency power source (when offsite power is not available),

a predetermined number of pressurizer heaters and associated controls necessary to establish and maintain natural circulation at hot standby conditions. The required heaters and their controls shall be connected to the emergency buses in a manner that will provide redundant power supply capability.

2.

Procedut:s and training shall be established to - 2 the opera-tor aware of when and hew the required pressurizer heaters shall be connected to the emergency buses. If required, the proce-dures shall ideatify under what conditions selected emergency loads can be shed from the emergency power source to provide sufficient capacity for tae connection of the pressurizer heaters.

3.

The time required to accomplish the connection of the pre-selected pressurizer heater to the emergsney buses shall be consistent with the timely initiation and maintenance of natural circulation conditions.

4.

Pressurizer heater motive and control power interfaces with the emergency buses shall be accomplished through devices that have been qualified in accordance with safety-grade requirements.

PCE Response 1.

The power and control circuits for the pressurizer heaters are supplied from the nonsafety-related 480 V load centers B09 and B10 (FSAR Figures 8.3-1 and 8.3-15).

However, all pressurizer heaters (backup and control groups) can be manually connected to the emergency diesel generators through 4 kV switchgear. Therefore, the current Trojan design meets the requirements to connect the pressurizer heaters and controls to the emergency buses in a manner that will provide redundant power supply capability.

A transient analysis has been performed by Westinghouse (for the Westinghouse Owner's Group) of the loss of offsite power event to determine the minimum pressurizer heater capacity needed to maintain subcooled conditions in the RCS under natural circulation conditions. The results of,4is conservative analysis indicate that, for a 4-loop plant 3 pressurizer, a heater such as Trojan with an 1800 f t 1276 223

S Page 2 of 49 capacity of 150 kW is more than eAequate. Furthermore, the analysis demonstrated that the Ec3 heat capacity is such that adequate subcooling would be saintained in the RCS for up to 4 hr without heat input ; Mom the pressurizer heaters.

The current Trojan design has 26 heaters with a capacity of 69 kW each, which are evenly divided into each bus. One train of power supply will be more than adequate to supply 150 kW.

2. & 3.

Current Trojan Plant Emergency Instruction EI-4 " Plant Operation After a Loss of Normal and Preferred Power" instructs the operator on manually connecting pressurizer heaters to the emergency buses. Load shedding of emergency loads is not required. This procedure will be revised to require that, if offsite power cannot be restored to the pressurizer heaters within 60 min, at least 150 kW of heater capacity shall be manually connected to an emergency bus immediately. The procedure vill include precautions to ensure that, if one emergency electrical bus is deenergized, the heaters are connected to the redundant bus. The manual connection can be completed well within the necessary time period for providing heat inpot from r.he pressurizer heaters.

4.

The existing electrical distribution system that enables the operator to manually connect the pressurizer heaters to the emergency buses is qualified to safety requirements in accordance with the Trojan FSAR.

Implementation Schedule No equipment changes are needed to meet the~ requirements of this position.

The required changes to the Emergency Instruction and associated operator training will be completed by January 1, 1980.

TJi/CJP/ HEW /gah/4kk66.44A1f

Page 3 of 49 Section 2.1.1 - Emergency Power Supply Requiresents for the Pressurizer Heaters, Power-Operated Relief Valves and Block Valves, and Pressurizer Level Indicators in PWRs NRC Position 2 Power Sucoly for Pressurizer Relief and Block Valves and Pres 4uriser Lovel Indicators 1. Motive and control components of the power-operated relief valves (PORVs) shall be capable of being supplied from either the offsite power source or the emergency power source when the offsite power i; not available. 2. Motive and control components associated with the PCRV block valves shall be capable of being supplied from either the offsite power source or the emergency power source when the offsite power is act available. 3. Motive and control power connections to the emergency buses for the PORVs and their associated block valves shall be through devices that have be.en qualified in accordance with safety-grade requirements. 4. The pressurizer level indication instrument channels shall be powered from the vital instrument buses. These buses shall have the capability of being supplied from either the offsite power source or the emergency power source when offsite power is not available. PCE Response

1. & 2.

The existing electrical power and control circuits for the PORVs (PCV455A and PCV456) and the PORV block valves (M08000A and B) are safety-related circuits that are supplied from either the Class IE station batteries or the emergency diesel generators when offsite ' power is not available (FSAR Figure 5.1-1). Hence, electrical power is available for these valves independent of the availability of offsite power. 3. The electrical circuits for the PORVs and block valves are safety-grade circuits with no interfaces with nonsafety-related electrical buses or equipment. The control power for ene ersin (PORV plus block valve) is provided from Train "A" of the safety-related electrical system, while controlling power for the other valve train is provided from the electrical system Train "B". Furthermore, the control power supply to the FORV on each train is diverse from that to the block valve; the PORV is controlled from the Class IE station batteries and the block valve from the emergency diesel generator. The safety grade requirement for power supply is in accordance with the Trojan licensing commitment in the FSAR. 1276 225

Page 4 of 49 The motive power for the PORVs is air and is supplied from the station instrument air supply. Each PORV has Seismic Category I air accumulators for operating the valve in the event that the normal nonseismic air supply is lost or isolated during a design basis event. The accumulator tanks are sized to support approximately 30 PORV operations. The PORVs will fail closed on loss of the motive air supply. 4. The Trojan plant pressurizer has three level indicators (LI459, LI460 and LI461) which are powered frca che emer-gency instrument buses. These instrument buses can be supplied by the offsite power source or the emergency power source when offsite power is unavailable. Implementation Schedule The current Trojan design meer.s the requirements of this position; hence, no equipment changes are needed. m KM/CJP/Hr4/4kk66.44A3 1276 226

Page 5 of 49 Section 2.1.2 - Performance Testing for BWR and PWR Relief and Safety Valves NRC Position Pressurized water reactor and boiling water reactor licensees and appli-cants shall conduct tasting to qualify the Reactor Coolant System relief and safety valves under expected operating conditions for design bcsis transients and accidents. The licensees and applicants shall detecnine the expected valve operating conditions through the use of analyses of accidents and anticipated operational occurrences referenced in Regula-tory Guide 1.70, Revision 2. The single failures applied to these analyses shall be chosen so that the dynamic forces on the safety and relief valves are maximized. Test pressures shall be the highest pre-dicted by conventional safety analysis procedures. Reactor Coolant System relief and safety valve qualification shall include qualification of associated control circuitry, piping and supports as well as the valves themselves. PGE Response PGE is working with the Westinghouss owner's Group and the Electric Power Research Institute (EPRI) to develop a program for qualification testing, on a prototypical basis, of relief and safet; valves. The testing program to be developed'will meet the intent of the NRC concerns by demonstrating valve operability under expected operating conditions, including solid-water and two-phase flow conditions. The test program will include consideration of the effects of both upstream and downstream piping configuration, and the results of the test program will be used to evaluate the specific discharge piping configuration at the Trojan plant. Details of how this test data vill be applied to Trojan will be included in the test program description. It is recognized that the NRC intends to work closely with EPRI to develop an acceptable test program. Implementation Schedule A description of the testing program and the testing schedule will be provided by EPRI prior to January 1,1980. The ' completion date for the test program depends on the availability of test facilties adequate to perform the tests over the parameter range of interest. The NRC will be advised by EPRI, as soon as the 'information is available, of the availability of suitable test facilities, and an acceptable comple-tion schedule can be developed at that time. 1276 227 KM/CJP/gah/4kk66. 2 5

Page 6 of 49 Section 2.1.3.a - Direct Indication of Power-Operated Relief Valve and Safety Valve Position for PWRs and 37Rs NRC Position Reactor system relief and safety valves shall be provided with a positive indication in the control room derived from a reliable valve position detection device or a reliable indication of flow in the discharge pipe. PGE Response The two Trojan prsssurizer power-operated relief valves (PORVs) (PCV-455A and PCV-456 at Location H-7 on FSAR Figure 5.1-1) have NAMCO, D2400X limit switches that are used for direct position indication. There is open and close indication for these valves in the control room. The limit switches are seismically qualified, but not designed for post-LOCA radiation, presaure and temperature conditions. These switches will be upgraded to meet the qualification for post-LOCA environmental conditions in accordance with the Trojan licensing commitments in the FSAR. The cabling and power supply equipment are presently qualified to the safety-grade requirements as specified in the Trojan FSAR, Sections 3.10 and 3.11. The three Trojan pressurizer safety valves were supplied by Crosby Valve and Gauge Company (PSV-8010A, B, and C at Location H-6 on FSAR Figure 5.1-1) and do not have positive valve position indication. Since the safety valves are provided for overpressure mitigation and a LOCA is a depressurization event, we do not believe it is necessary to have associated equipment qualified for extreme, worst-case LOCA pressure and temperature operation. To accomplish positive safety valve indi-cation, the two alternatives which are under consideration at the present time are: (1) Direct position indication by use of Ibnit switches and indicating lights in the control room; or (2) Flow indication by use of acoustic monitoring devices on each safety valve outlet line with associated cabling, amplifiers, power supply, signal processing equipment and indicating lights. Although there currently is no preven, reliable design cevering the installation of a position limit switch on a Crosby safety valve, this approach is being considered. Likewise, qualification of the acoustic sensor equipment for safety-related operation has not been accomplished. However, near-ters investigation will be continued on these technologies to choose the best of these two alternatives. Indication of the safety valves will be accomplished by a safety grade power source and seismic-ally qualified cabling, signal processing equipment, and indication which satisfy design requirements consistent with the Trojan licensing commit-ments in the FSAR. 1276 228

Page 7 of 49 Sufficient ibdirect valve position indication already exists through flow parameters so that redundancy for the above described equipment is not required. The indirect indicators are: (1) The temperature in the pressurizer safety and relief valve discharge lines which is indicated in the control room (TE 463, 464, 465, 466 at Location H-6 on FSAR Figure 5.1-1). (2) The pressurizer relief tank high pressure, high level and high water temperature which are indicated in the control room (PT 469, LT 470 and TE 468 at Location G-2 on FSAR Figure 5.1-1). Implementation Schedule We are attempting to meet a schedule date of January 1, 1980 for instal-lation of additional pressurizer PORV and safety valve indicator devices. However, the Plant would have to be shut down in December 1979 to perform the installation work, and an extremely deficient power supply situation is expected in the Pacific Northwest this winter due to low hydroelectric reservoir levels. Replacement power would likely have to be generated by burning oil in our combustion turbine facilities. Therefore, the instal-lation work will commence during the first cold shutdown after December 1, 1979 and will be completed no later than the startup of Cycle 3 operation in 1980. The currently installed PORV stem position indicators, the discharge manifold temperature monitors in the PORVs and safety valves, and PRT pressure and level indicators, together with increased operator attention to these parameters as an outcome of the DfI accident, provide reasonable assurance of open valve diagnosis in the unlikely event of an overpressure transient during the interim period. Any necessary addi-tional qualification tests or documentation will be completed by June 1, 1980. 1276 229 DIH/ HEW /kk/4mg66.44A6

Page 8 of 49 Section 2.1.3.b - Instrumentation for Detection of Inadequate Core Cooling in PWRs and BWRs NRC Position 1. Licensees shall develop procedures to be used by the operator to recognize inadequate core cooling with currently available ins trumentation. The licensee shall provide a description of the existing instrumentation for the operators to use to recog-nize these conditions. A detailed description of the analyses needed to form the basis for operator training and procedure development shall be provided pursuant to another short-term requirement, " Analysis of Off-Normal Conditions, Including Natural Circulation" (see Section 2.1.9 of this appendix). In addition, each PWR shall install a primary coolant saturation meter to provide on-line indication of coolant saturation condition. Operator instruction as to use of this meter shall include consideration that ie not to be used exclusive of other related plant parameters. 2. Licensees shall provide a description of any additional instru-mentation or controls (primary or backup) proposed for the plant to supplement those devices cited in the preceding section Fiving an unambiguous, easy-to-interpret indication of inade-quate core cooling. A description of the functional design requirements for the system shall also be included. A descrip-tion of the procedures to be used with the proposed equipment, the analysis used in developing these procedures, and a schedule for installing the equipment shall be provided. PCE Response 1. The Westinghouse Owner's Group, of which PGE is a member, is per-forming calculations associated with the definition and recogni-tion of inadequate core cooling conditions in accordance with Item 2.1.9. A description of the analysis program and its rela-tionshi~p to this item is found in our response to Section 2.1.9. Appropriate procedures will be developed in accordance with the analysis program. Redundant, safety-grade Reactor Coolant Subcooling Margin Monitors (SMMs) will be installed to provide on-line indication of the saturation condition of each of the RC loops. Design details of the SMMs are currently under development. The SMMs and the signal inputs from temperature and pressure measurement will satisfy the safety-grade requirements in accordance with the Trojan FSAR. Operating procedures will be developed so that information from the SKMs is coordinated with other moni-tored variables by the operator to ensure that proper subcooling margin is maintained. 1276 230'

Page 9 of 49 2. A description of the analysis program and commitment to meet this item can be found in our response to Section 2.1.9. Implementation Schedule Operating procedures and associated operator training vill be Duplemented in accordance with the saatysis and the development of precedure guide-lines described in our response to Section 2.1.9. Equipment for the SMMs can be delivered by approximately January 1,1980. A Plant shutdown could be required for installation of the SKMs in the control reem panels, and a cold shutdown would be required if additional RCS RIDS are included in the system design. Therefore, installation of the SMMs vill be completed by startup of Cycle 3 in 1980. This is the earliest practical Laplementation date based on the time required to develop the final system design, procure the raquired equipment, install and test the system, and develop the required operating procedures. During the interim period, as indicated in PGE response (dated June 25, 1979) to IE Bulletin 79-06A, the Emergency Instruction has been revised to ensure that the operators maintain adequate subcooling margin based on current pressure and temperature instruments and the pressure-temperature curve attached to the Emergency Instruction. 1276 23i KM/CJP/CCT/4kk66.44A8

Page 10 of 49 Section 2.1.4 - Containment Isolation Provisions for PWRs and BWRs NRC Position 1. All Coacainment isolation system designs shall comply vich the recommendations of SRP 6.2.4; i.e., that there be diversity in the parameters sensed for the initiation of Centainment isolation. 2. All plants shall give careful reconsideration to the definition of essential and non-essential systems, shall identify each system determined to be essencial, shall identify each system determined to be non-essential, shall describe the basis for selection of each essential system, shall modify their Con-tainment isolation designs accordingly, and shall report the results of the re-evaluation to the NRC. 3. All non-essential systems shall be automatically isolated by the Contain=ent isolation signal. 4. The design of control systems for automatic Containment isolation valves shall be such that resetting the isolation signal will not result in the automatic reopening of Containment isolation valves. Raopening of Containment isolation valves shall require deliberate operator action. PCE Response 1. The diversity for the initiation of Containment isolation at the Trojan plant is satisfied by the following six parameters: Number of Number of Functional Unit Channels Channels to Trio a. Manual Containment 2 1 Isolation b. Manual Safety 2 1 Injection c. High Containment 3 2 Pressure (5 psig) d. High Differential 12 (3/ Steam 2/ Steam Line and Pressure Between Line) 1/4 Comparison Steam Lines Between Steam Lines e. High Steam Line 8 (2/ Loop) for 1 per Steam Line Flow Coincident with Hi Steam Line and 2/4 loops

1) Low Steam Line Flow; 4 (1/ Loop) for Hi Steam Pressure for Lo-Lo for Low Steam Line flow; 2 for T

and Line Pressure; Lo Steam Line Nctavke,eamLine 4 (1/ Loop) for Pressure; 2 for 2) SI E '~k Lo-Lo Tavg Lo-Lo Tavg f. Low Pressurizer Pres-3 2 sure and No Pressur-izer SI BTock 1276 232

Page 11 of 49 These diverse Containment isolation initiators assure the early isolation of the Concainment Building. 2. The design basis for Containment isolation is given in the Trojan FSAR Section 6.2.4, and each of the Containment isolation barriers is listed in Table 6.2-1 of the FSAR. All fluid lines which do not serve an engineered safety features function subsequent to a design basis event are isolated. The criteria for the number and location of Containment isolation valves in each fluid system depend on its function and whether it is open or closed to the Containment atmo-sphere or Reactor Coclant System. Fluid system penetrations era classified into four types: Type I, Type II, Type III, or Type IV, as described in FSAR Section 5.2.4. The criteria for Containment penetrations ensure that all fluid lines penetrating the Containment have at least one isolation valve near the point of penetration. Those lines that concunicate directly with the Containment atmosphere have, as a minimum, two isolation valves in series. The exception to this criterion is the Ccocainment pressure-sensing lines which are required to function following a design basis accident to actuate Containment high pressure and Containment high-high pressure switches. The attached Trojan FSAR Table 6.2.1 summarizes the different types of isolation barriers provided for each fluid system penetrating the Containment. This table indicates the type of isolation valves pro-vided, the normal and accident position, the mode of operation, the actuation signal, the time required for closure. The time required for valve closure is defined as the time from receipt of the electri-cal signal to the end of stem movement. In accordance with NUREG-0578, the four types of fluid systems as defined in the FSAR can be grouped into " essential" and "nonessencial" systems as follows : a. Essential systems are defined to be those ESF system fluid lines penetrating the Containmenc Building, which must remain operable following a design basis accident. These lines are the Type IV fluid lines for the ESF systems (i.e., Residual Eeat Removal, Component Cooling Eater, Centrifugal Charging Pump, Safety Injection, and Containment Spray Systems), the four Containment pressure sensing lines and the hydrogen vent lines listed on FSAR Table 6.2-1. With the exception of the hydrogen vent line, the Containment boundary isolators on these essential lines do not receive Containment isolation signals. These systems can be selectively isolated if it is determined that they are not needed to re=ove reactor heat. The Containment hyrdogen vent lines are automatically iso-laced by the Containment isolation signal. Hydrogen vent lines may be required at about 8-9 davs after a design basis accident to exhaust Containment air with hydrogen gas which 1276 233

Page 12 of 49 is being mixed and recombined inside the Containment. The hydrogen vent syster does provide a direct path for radio-activity to flow fr.a the Containment to the atmosphere. However, by the time it is required to function, the Containment isolation signal can be reset. b. Nonessential systems are those non-ESF system fluid lines penetrating the Containment Building which are not required to be operable immediately following a design basis event. These lines are the Type I, II and III fluid lines listed on the attached FSAR Table 6.2-1. These lines are either:

1) Used during normal Plant operation and receive a Containment isolation signal to close.

2) Not used during normal Plant operation with their Containment isolators normally closed and, therefore, not requiring a Containment isolation signal.

3) Part of a closed system inside Containment and, therefore, no* a highly probable pathway for a radioactivity release to the atmosphere.

4) Used during noinal Plant operation and receive a safety injecti-,n signal to close which also causes a Contaiczent isolation.

5) Part of a closed system outside Containment and, therefore, not a highly probable pathway for radio-activity release from the Containment to the atmosphere.

No Containment isolation design modifications are required for either the essential or nonessential systems defined above. 3. As described above, all nonessential system fluid lines are either: a. Used during normal Plant operations and isolated by a Containment isolation signal or a safety inj2ccion signal. b. Not used during normal Plant operation and normally closed. Part of a closed system either inside or outside Containment c. and not a highly probable pathway-for radioactivity release from the Containment to the atmosphere. Normally closed valves during Plant operations are assumed to remain closed following a design basis accident and, therefore, provision of an automatic closing signal is not required. Those systems which form a cicsed loop either inside or outside Containment are not connected to the primary system where the design basis acci-dent would be occurring and, therefore, unless some other event occurs to violate their integrity, a pathway from the Contain=ent atmosphere to the autside atmosphere cannot be established through them. As such, no automatic isolation is required. }27b

Page 13 of 49 Since current Trojan design requires all nonessential systems, which could be ope.n during normal Plant operation to provide a pathway between the Containment atmoschere and the outside atmosphere, to shut on a Containment isolation signal, no additional Plant modifi-cations are required. 4. A complete review of the control circuits for each of the Contain-ment isolation valves has been made. The review has identified 23 Containment isolation valves that will return to a previously open position after resetting the Ccntaicsent isolation signal. This design review as well as implementation of the design changes of these valves were discussed in previous PGE response (dated July 7, 1979) to IE Bulletin 79-069A. The control circuitry for these 23 valves will be changed to prevent the autenatic opening upon resetting of the Containment isolazion signal and to require a manual operator action to open the valves after the Containment isolation reset. All other valves which receive a Containment isolation signal cannot be reopened following a Concainment isolation reset without requiring a deliberate operator action. Implementation Schedule Since material lead times for this modification are from 10 to 14 weekr and some of the circuit modifications identified in the Response to Question 4 above must be performed inside Containment, installation of these modifications mast occur during a cold shutdown. Therotere, tais verk will be completed prior to the startup of Cycle 3 in Ifd0 fer the reasons discussed in the implementation schedule for Section 213.a. In che interim period, administrative controls have been implemencea ths.c require operators to ensure that the 23 affected valves are in the closed position prior to resetting a Containment isolation signal. KM/ RAY /kk/4sa66.44A10

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9 fe I g 3 3 3 3 I. 3, 3 1 t t e e sp 1 1 I I 3 3 r ey h TT E es e) umc si e 0 0 0 0 o 0 e lot s ( 1 C 1 1 1 t 1 l t a ne d J d d J d d J J d d J e8 e e e e v e e e e e e e s s s s s s n s s s s d4 o o o o o .s il o l l l C C 4 C l l l l o o o o o l ce l l 3 l C C C C c e C C C C AP A A a A. A e A 8 A 3 l l l l l l l leS e l 3 eS e 1 e l 5 S e3 e 3 n l nI n n lseb e I nC n C nC n a s t n n;n l nI n ig CnC n $ n C nC n C a a a a a e a a a S h h h h h C C C h h h C C C C C C d J d d J d J J e e e e e e e e n t t t t d s t d t J t d e a a e a e a e f o a a a a t r s t s t r t 4 oi r r r r t e e e e a e o a e a e a ea p p p p r p p r p r p r J u o o o o e o o e o e o e ot p p p p Mc r r r r o r r o r o r o A o o e o o o o o t t t t r t t r t r t t M 1 l M A M H i N i N t o a u o i o u 6 e A A A ev e e e e e e e e e c v v v v v v v ) v v v J l l l i l l l l l l l l a a a a t v a a a a a a a g s s v v v a g v v v v v v v no o m ( d m o, C a n n n n nen 8 n m n a m a o e e o u o o e uso ( t o e e u t e t u t ) l t l i i t l loi o o t e pi e t e t i t v s v se t t v 1 a, s e t t t t t t rl l y l y e e l t y v .e d a a a a a acat d a ev ndl d dlJ dld l 2, c a a s a e lenon v loe ta, e 6, so v v l a n d, s l s i. eg a ae ue loeoe loeoe usne a m s e e ms s aees eF d d s eee ep , d v sss l laI l cC cC cacC s D k e k e $ pg p k t O ev c s c s O O c r s h

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cC cC lvt o uio l eoa o u l l l l l i r a e o e o c c e c2 e, o t e v scc it a t t t t C s em h h l e A ir C l C C i i h i p, l h a t i i i vi J.e C t t C t og,C t T r o a a a C ni a a a a a a) a rN m m m r, ioe k m m m m m m4 m o u o o( o a o o o og r o u s t t t t B t t t t, C( u t t A A A pe g, h u u e u u u u u u u C A A A A A A A e l l t l V V p I I I I I I I 8 8 I y T n e e e e o e e a e e e d e d e J e d e e d e J e J e J t Jts i s Jt l J l Jl l J6 Jt i Jt e t e t l e e t e o o e t J l J t d e s t s t s t e t o t o t e a a a s e s t. t. a a n u u u n u s u u n h e n h n w n r.. I t I t i 0 t o I t B o l G i O l O i o L 1, 4 4 ee, 4 2 3 / / ipat I 2 2 4 3 1 1 3 3 n pS( s L en 9 t n n i l e s r p p t e ae e a a i l 8 e 3 m ) m 3 mn2 vl% J 8 vtS 8 a 8 3 M t el p M eH le2 u 0 u 0 u 0 oi0 ot 0 ot u 2 r 1 ,0 s 2 d r2 M py2 py2 pn2 Cl2 m n2 maM ) e p i t lp tlp t u es r e i .t e M M rM M e c s u m t s m t o a np np nt mn r u ny8 s6 tn tl s t a 8 4 lal r s 0 e l u lau ae u a, S aa6 w a a. a1 2 le s s3 s3 r4 l u l. e.- - t on< r - ot em-v. h a 2, n 2 oa2 e2 o o u c e c e. a e. V a2 h s 2, or2 s2 s o r 2, t 6 a6 e4 l a o, r s 6 t 6 s6 l h. l M6 vI s I. 6, e n, c sa6 r r r a l. o a r a r a a a, . s . a. r s e ege . e ewe ewe ewe c me e f ee uee tel f e gs '.t r t s ms er J ps u l u a u sk u u u t s s r t s tet u tewg u sl a el u cl u i .r 9 a s t g snm s ml M mE M aD l , s M n r J ig eng s g aag nag a as e sg l eai ci f Weoi

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== .9 .e_ e 9 W m W _.d _- N, g 3-g

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m _.- l-w -m am 6.m g e. t 4 N o N e .= a ~, N, N, N, N, N, a .] 6 m m 6 6 6 6 6 y J 6 (C;) tit) A nd at 24 (N.ve.o r 1975) 276 238 ~.

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a. ce.

.=. ee .. ce g =. m. 1 g . ce ce ce 1 ee e ce se .s. c. m e = .. c.e c,e a e e =e = ee

e es e s

= p e e ce ce ce a ce a es a ce a es e ce e.s. ce_ s. s.= d w e ) _ t g 4 ; a. s es en - w ; w _ .e 4 =ee; e4; ae a4 e=d e4 es as as e as ee ee .e 4 a e. e e. = e ee e ee e e e w e en w s-. 8' e e. I-3 3s e e. e e. e e e s-a i, 3,. .s. m ,3 ,.m -a3 .3 3 3 3 a es ea e. .e .e a .e .e e g 3 g =. .=. s-e e ee e e g-w e. 2 r r. 2 -e- -e- _e._ =. r e 5 c.e c,o =y e -a d .e. ee ce a e e 4 e= .o e. e. e, m, w s = s. a. a, s. e, a. a. a. a. a. a. 3 5 b4 e a. IIII NI UII N dU Amendment 24 1276 239 Otovember 1973)

o 9 9

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s. I. s 6 r-

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• t T

- 3 - 3 4 3-j T T f 4 6 0 5 - 5 - d -LI: e e e e =

n. e -

_o e = ~ ~ ~ ~ u - ew z r - r ce me - ~ o e ~ o o o o az . x v N w s ae ee ce ce ce x .x l ,.z z = 3 ,e g .a g 1 .S 3 a.n L l. 4 i2 5 a, z 8 ,= a .m 3 m .m 6 I .a-J } 4 3% 2. J. g .,, N .= ** N S 1 m? 4 3 w e . r. 8 . &M .m A me 3 ** a. J e. N SS 3 3 - =,. - _1 * ; -d E J .3 ) 4 4 ,N ,Me g.M J g34 .< d a4 g d 2 3* e e ,.3 ,.m a, 38 ,.m . c.. , a,. s a.. ,s a.. -s 3 s. -,3 =. .s a .i. n. as. s s = s s s a.=. .a. . = .s w

• am

].u. . ).=. 1 1 =.- n.- m-i u gi.. e 4 m e- = o 4 a 4, 4 4 4, e, _e. o.* A n% e 4 . a e. a. a. a. a. a. a. u n. a. b A a. utJ LEJJ ot; s7c) OU LG) sta Am.nd nt '4 vv 3 4$76 241 (Noveener 1973)o** o .A.tv o oo o

9 fo g 3 1 3 1 1 1 3 2 ) 1 I g 3 3 I 1 ?, I 3 t e 1 8 1 3 1 3 1 t s p e e ey h TT S e r ) eec s me n u 5 5 l T( i l S i 3 0 0 ot 1 1 0 ois C t n no ei d d d d J d J J d d J J d t e e e e e e e e e e e e. e J i s s a o s m s s s s s s l AP l l C l C l l l l l l o o cs o o e e o e o o o o o l l. c o l C l C C C C C C t C C C A B A n A B A a l ? l l l l l l l an C eS e S eS e 3 eS e S eS e nI n lg B nB n B nI n l nI n I nC n C nC n CnC n CnC n C a a a a a a a a h h h h h h h h C C C C C C C C d d d d d d d d e e e e e e e e n a a s d t d d t t t d t t t a e a e e a a a e ar f e r r r 't r t t r r r t e ol e e e a e a a e e e a l p p p r p r r p p p r p ee o o o e o e e o o o e o d u p p l p p ot r r r o r o a o r r r o r o o Mc o o o o u o o t l M r n r t t t t A t t t r t i ll A o o o o i a a i N M M A N M N t A S A e s v d - o l d - o d - o e esn e an v een a r s r s e e e e s l r e e J e e e e e e ir e ) v iupe l l et i n a v v upe v v v v v v v upe n q r q r t n l l la a l l et et a a l q r n o 's i n v v v v v v r nui i r nui v v r nu a a o a a a a ( t e d eq e eq eg l e ta t pt n n n n n n ee s me v sme n n sme 1 it o i i i i i v o inr l l i ler o o i e s a v o o o o o e vs i 7 c s t t t t t t l lal nae d vd nae t t nae l l m i i l o ri i a a a S a ot r sd vd a a a a a a wc ot r e e ot r a lod t of s n. 6, so a ted l l l t v i no & sns e e i no l i n o. I eP l t of uJ D u o o o e ld t of ) oool r r i sis r r sl i i a ee s s s e k ae ac e 2li ace n cs e aca & noo r l l nS s i nf n c uk r r (Ct C i n i n. r a e e e e nc t ee s a t ee o o t ee eCt t i em e o s a s e erh cl cl eri v A i r c pcl cpcp cpcp h a o e ri e l nut iCiC nut l l Oio C Ml nut v o oiC ioio i T r lo it t t t t t es l I pa& e a a pa6 s v rl s es t t es a al' a a a a a a a pa& s a e v m m e e e n n m m e e v B ms t t t t t ismev ismev o o nmev k i p e u u o rl rl t t i rl r s b oia u u h oua e e u u u u u u h oua h r Tt ev A A T t sv k A A A A A A Tt sv e r h 2 't e V 8 I V I p I 1 I B I y T n e e e e e e e e e e J e J e J e d e d e J o e J e J e J e d e t d t t l i e l l Jl t J J d l J l s t t Jt l dl l Jl t s l t s s n e a t n e s J l t e a a e a s a a s ts s s n u a u n t n u a n u a e s a s t s t a o s. I au h h u 3 O l o I O I O l 0 I O l s i ht l t I t s. 4 ) ec. ipr In 4 4 1 4 5 S 1 4 f Ps( 8 7 6 7 ) 7 iw ) 8 e 2 n 0 0 4 0 e 4 4 8 e M n 2 a H l 2 2 2* 2 f 2 2 6 ts 2 l 9 P i r M w M M s ' t

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Page 14 of 49 Section 2.1.5.a - Dedicated Penetrations for External Recombiners or Post-Accident Purge Systems NRC Position Plants using external recombiners or purge systems for post-accident combustible gas control of the Containment atmosphere should provide Containment isolation systems for external recombiner or purge systems that are dedicated to that service only, that meets the redundancy and single failure requirements of General Design Criteria 54 and 56 of Appendix A to 10 CFR Part 50, and that are sized to satisfy the flow requirements of the recombiner or purge system. PGE Response The Trojan plant design currently includes redundant Seismic Category I hydrogen recombiners inside Containment as the Hydrogen Control System. The hydrogen recombiners are the primary hydrogen removal system and, because they are located inside Containment, no Containment isolation devices are required. As a backup system, there is a redundant Seismic Category I Hydrogen Vent System which would be operated in the event the hydrogen recombiners are inoperable. The penetrations for the Hydrogen Vent System are dedicated for this operation (FSAR Figure 6.2-48) and are equipped with two Seismic Category I dampers in series in each line penetrating Containment. These dampers are normally closed (FSAR Fig-ure 6.2-45). Each line is sized to 8 in. in diameter to accommodate & maximum flow of 140 cfm which is the rated capacity of the blower in the Hydrogen Vent System (FSAR Section 6.2.5). The Containment penetration lines and isolation components for the Hydrogen Vent System are designed in compliance with the redundancy and single failure requirements of General Design Criteria 54 and 56 of Appendix A to 10 CFR Part 50 as discussed in FSAR Section 3.1. The Seismic Category I requirements of the hydrogen recombiners and Hydrogen Vent System are describ6d in FSAR Section 3.7. Based on the above, the Trojan plant design currently satisfi 's the NRC position and no action is required on this item. Implementation Schedule No action is required on this item. 1276 245 icf/ Soc /kk/4sa66.44A14

Page 15 of 41 Section 2.1.5.c - Hydrogen Recombiners NRC Position 1. All licensees of light water reactor plants shall have the capanility to obtain and install recombiners in their plants within a few days following an accident if Containment access is impaired and if ruch a system is needed for long-tern post-accident cembustible gas control. 2. The procedures and bases upon which the recombiners would be used on all plants should be the subject of a review by the licensees in considering shielding reiuirements and personnel exposure limitations as demonstrated to be necessary in the case of TMI-2. PCE Response 1. The Trojan plant design currently includes redundant Seismic Category I hydrogen recombiners located inside Containment. Therefore, the Trojan plant design satisfies the NRC position and no action is required on this item. 2. Since the hydrogen recombiners located inside Containment can be operated remotely from the control room, no shielding analysis or procedural review for personnel exposure is required. Imolementation Schedule None required. KM/ SGG/gah/4kk66.44A15 1276 246

Page 16 of 49 Section 2.1.6.a - Integrity of Systems Outside Containment Likely to Contain Radioactivity (Engineered Safety Systems and Auxiliary Systems) for PWRs and BWRs NRC Position Applicants and licensees shall immediately implement a program to reduce leakage from systems outside Containment that would or could contain highly radioactive fluida during a serious transient or accident to as-low-as practical levels. This program shall include the following: 1. Immediate Leak Reduction a. Implement all practical leak reduction measures for all systems that could carry radioactive fluid outside of Containment. b. Measure actual leakage rates with system in opera-tion and report them to the NRC. 2. Continuing Leak Reduction Establish and implement a program of preventive maintenance to reduce leakage to as-low-as practical levels. This program shall include periodic integrated leak tests at a frequency not to exceed refueling cycle intervals. PGE Response _ 1.a. Trojan currently has a preventive maintenance program for the purpose of keeping leakage from radioactive systems to as low as practicable levels. Leakage control was included in the preventive maintenance program to minimize in-Plant airborne radioactivity and contamication levels and to reduce vaste inputs into the radwaste systems. 1.b & 2. A supplemental program will be developed for periodically testing systems outside Containment which may contain highly radioactive fluids or gases following an accident. The following systems will be included in this test program: a. Chemical Volume and Control System b. Residual Heat Removal System c. Containment Spray System in the recirculation mode d. Safety Injection System in the recirculation mode e. Radioactive Waste Gas Treatment System. 1276 247

Page 17 of 49 Where possible, the testing vill be performed by inspecting potential leakage areas such as flanges, pump seals and valve packing while the system is maintained at operating pressure. An alternate means of testing will be developed for the Radioactive Waste Gas Treatment System or systems that do not contain water except during actual operation following an accident. The acceptance criteria for leakage will be che design leak race for individual components within the system. A schedule for repair of the source of excess leakage will be established to ensure excessive leakage is repaired in a timely manner consistent with system operation and the magnitude of the leak. Testing vill be done at refueling cycle intervals. The results of the initial leak rate tests will be reported to the NRC. Weld and piping integrity will continue to be tested in accordance with the Inservice Inspection Program. Imolementation Schedule The supplemental program will be developed and ready for implementation January 1, 1980. The initial tests will be performed at the first opportunity after this date but, in any event, no later than the startup of Cycle 3 in 1980. For some systems, the tests may have to be conducted during Plent shutdown or during specific system conditions which are not possiale to schedule precisely. K.M/TDW/GC/gah/4kk66.44A16 1276 248

Page 18 of 49 Section 2.1.6.b - Design Review of Plant Shielding and Environmental Qualification of Equipment for Spaces / Systems Which May be Used in Post-Accident Operations NRC Position With the assumption of a post-accident release of radioactivity equiva-lent to that described in Regulatory Guides 1.3 and 1.4 (i.e., the equivalent of 50 percent of the core radioicdine and 100 percent of the core noble gas inventory are contained in the primary coolant), each licensee shall perform a radiation and shielding design review of the spaces around systems that may, as a result of an accident, contain highly radioactive materials. The design revies should identify the location of vital areas and equipment, such as the control room, radwaste control stations, emergency power supplies, motor contrek centers, and instrument areas, in which personnel occupancy may be unduly limited or safety equipment may be unduly degraded by the radiation fields during post-accident operations of these systems. Each licensee shtll provide for adequate access to vital areas and protection of safety equipment by design changes, increased permanent or temporryy shielding, or post-accident procedural controls. The design review shall determine which types of corrective actions are needed for vital areas throughout the facility. PGE Response A design review of the radiation shielding at Trojan plant will be conducted for vital areas and safety equipment locations to ensure that personnel occupancy will not be unduly limited by post-accident radiation fields. The following source terms referenced in NUREG-0578 vill be used in the design review: 1. For liquid systems - 100 percent of the core inventory of noble gases, 50 percent of the halogens, and 1 percent of others. (Certain liquid systems may be analyzed with reduced noble gas concentrations when " upstream" liquid-to-gas partitioning can be demonstrated.) 2. For the Containment atmosphere - 100 percent of the core inven-tory of noble gases and 25 percent of the halogens. The following systems / functions will be considered in the design review: 1. Residual Heat Remcval System 2. Safety Injection System (recirculation mode) 3. Containment Spray System (recirculation mode) 1276 249

Page 19 ed 49 4. Chemical and Volume Control System 5. Radioactive Waste Gas Treatment System. Areas will be shielded as required to limit exposures during operational evolutions necessary for accident mitigation /da= age control below the limits of 10 CFR Part 20, CFR Part 50 and General Design Criteria 19. The design criteria for radiation exposure to safety-related equipment in - the Fuel and Auxiliary Building is in accordance with FSAR Section 3.11. Additional details pertaining to environmental qualification of safety-related equipment in the Fuel and Auxiliary Building are provided in PGE's response to lE Bulletin 79-01. Implementation Schedule The shielding design review requires an extentive amount of analysis effort and, based on our best es timate to date, is expected to be com-plete by January 1,1980. We will attempt to complete any Plant modi-fications or changes in post-accident pro:ed cres resulting from the design review by January 1, 1981. However, tue extent of structural modifications or changes in equipment arrangement will not be known until the shielding analyses are completed; some of the work could conceivably not be completed prior to the startup of Cycle 4 in Spring 1981. Addi-tional schedules for any required work will be provided by January 1, 1980. KM/ RNS/4kk66.44A18 l 27 b

2. 50

Page 20 of 49 Section 2.1.7.a - Automatic Initiation of the Auxiliary Feedwater System for PWRs NRC Position Consistent with satisfying the requi'*ments of General Design Cri-terion 20 of Appendix A to 10 CFR 50 with respect to the timely inicia-tion of the Auxiliary Feedwater System, the following requirements shall be implemented in the short term: 1. The design shall provide for the automatic initiation of the Auxiliary Feedwater System. 2. The automatic initiation signals and circuits shall be designed so that a single failure vill not result in the loss of Auxiliary Feedvater System function. 3. Testability of the initiating signals and circuits shall be a feature of the design. 4. The initiating signals and circuits shall be powered from the emergency buses. 5. Manual capability to initiate the Auxiliary Feedwater System from the control room shall be retained and shall be implemented so that a single failure in the manual circuits will not result in the loss of system function. 6. The a-c motor-driven pumps and valves in the Auxiliary Feedwater System shall be included in the automatic actu-ation (simultaneous and/or sequential) of the loads to the emergency buses. 7. The automatic initiating signals and circuits shell be designed so that their failure will not result in the loss of manual capability to initiate the Auxiliary Feedwater System from the concrel room. In the long term, the autoracic initiation signals and circuits shall be upgraded in accordance with safety-grade requirements. PGE Response All of the short-term and long-term requirements listed in the above position are currently satisfied by the Trojan Auxiliary Feedvater System, as discussed in Trojan FSAR Section 6.6. The following is the description of current Trojan plant design in response to each NRC position: 1 The existing Auxiliary Feedwater System (AFW) is designed to automatically provide feedwater to the steam generators during an emergency. Redundant full capacity trains of AFW are supplied with safety-grade (in accordance with FSAR requirements for instrumentation, control and power supplies) i276 251

Page 21 of 49 signals which are generated by (1) a valid safety injec. cion signal (SIS), (2) low steam generator level in any steam generator, (3) loss of normal onsite or preferred a-c electric power, or (4) trip of both main feedwater pumpa. 2. The AFW automatic start signals meet the requirements for engineered safety features actuation system as described in FSAR Section 7.3 and meet single failure design requirements. 3. The automatic start circuits are designed with provisions for periodic testing and calibration during normal Plant operations. 4. All circuits used for initiating AFW automatic start are powered from an ESF vital instrument bus, as defined in FSAR Section 8.3. 5. Manual operation of AFW systems is provided in the control room and at the local station. Each control circuit is independent so that a single failure in one train will not affect the redundant train. 6. All the power necessary for automatic and manual opera-tion of the AFW system is supplied by ESF vital instrument buses. Since one AFW pump is etrbine driven and the other is diesel driven, there are no large a-c loads in the AFW system and, therefore, the AFW system is not sequenced for loading on the emergency buses. 7. The automatic initiating circuits are designed to be electrictlly independent from the control room nanual start circuit so that the failure of the automatic initiating signal does not affect the control room manual capability of AFW pumps. A complete discussion of electrical separation for electrical posar supplies and instrumentation for auxiliary feedwater pumps is contained in FSAR 6.6.3. The seismic and environmental design of safety-related equipment in the Auxiliary Feedwater System meet the requirements described in FSAR Sections 3.10 and 3.11. Subsection 6.6.4 also lists the standards applicable for performance testing. Imolementation Schedule The automatic initiation signals and circuits presently meet all safety grade requirements. Thus, no changes are required. 1276 252 KM/DIH/CCT/gah/4kk66.44A20

Page 22 of 49 Section 2.1.7.b - Auxiliary Feedwater Ilow Indication to Steam Generators for PWRs NRC Position Consistent with satisfying the requirements set forth in GDC 13 to provide the capability in the control room to ascertain the actual performance of the Auxiliary Feedwater System when it is called to perform its intended function, the following requirements shall be implemented: 1. Safety grade indication of auxiliary feedwater flow to each steam genercor shall be provided in the control room. 2. The auxiliary feedwater flow instrument channels shall be powered from the emergency buses consistent with satis-fying the emergency power diversity requirements of the Auxiliary Feedwater System set forth in Auxiliary Systems Branch Technical Position 10-1 of the Standard Review Plan, Section 10.4.9. PCE Resoonse Four flow elements (FE 3043A, B, C and D at Locations H-6 and H-7 on FSAR Figure 10.4-2) are provided at Trojan to measure the auxiliary feedwater flow to each steam generator. Remote flow indication is provided in the control room and locally at the remote shutdown prael C160 in the Turbine Building. The flow elements, as well as the F&P Model 10B249 flow trans-mitters (FT 3043A, B, C and D), meet structural integrity requirements during and after a seismic event, and will function normally before and af ter the event. During a seismic evant, the transmitters may function erratically. However, since the seismic event will be of short duracica and the signals are not used for any equipment actuation, this behavior is acceptable for post-accident flow monitoring purposes. The flow elements and associated cabling are located outside Containment in the main steam support structure adjacent to the Turbine Building. This structure is compartmentalized vertically and open to the atmosphere at the top, which prevents adverse environmental effects (temperature, pressure and radiation) from a pipe break in a high energy line of one loop from being propagaced to the components associated with the other thrr,e loops. The vertical walls n trating compartments provide a complete barrier for steam or water pau ;e between compartments in the event of a high energy line break. The flow transmitters are located inside the Turbine Building. The existing cable runs from the flow transmitters to the panel indicators in the control and remote shutdown ?oom are not completely safety grade, in that there is no cable separa-tion and the power supplies are from an existing non-vital instrument bus. Redundancy in determining indirect auxiliary feedwater flow exists by using the two safety-grade channels of measurement and indication of the water level in each steam generator. The cable runs for each of these 1276 253

Page 23 of 49 channels are separate from the flow measurement discussed above. This indirect indication is a diverse and separate means of ascertaining from the control room auxiliary feedwater performance. Safety grade III Barton local flow indicating switches are also provided on each of the eight auxiliary feedwater lines (FIS 3004A1, A2, B1, B2, C1, C2, Dl, and D2 at Locations G-5 through G-7 on FSAR Figure 10.4-2). The ITI Barton flow indicating switches are located at the 59-ft-6-in. elevation in the main steam support structure adjacent to the Turbine Building. We believe that the diversity and redundancy that already exists in flow indication obviates the need to upgrade the existing cable runs previously described. However, we do plan to provide safety-grade power supplies from the vital instrument bus. Implementation Schedule The design change to provide the safety grade power supplies from vital instrument buses through appropriate isolation devices will be completed by January 1,1981. They will be consistent with the design requirements for Trojan licensing commitments in the FSAR. L4/DIH/gah/4kk66.44 A22 p76 254

Page 24 of 49 Section 2.1.8.4 - Improved Post-Accident Sampling Capability NRC Position A design and operational review of the reactor coolant and Containment atmosphere sampling systems shall be performed to determine the capa-bility of personnel to promptly obtain (less than 1 hr) a sample under accident conditions without incurring a radiation exposure to any indi-vidual in excess of 3 and 18-3/4 rem to the whole body or extremities, respectively. Accident conditions should assume a Regulatory Guide 1.3 or 1.4 release of fission products. If the review indicates that per-sonnel could not premptly and safety obtain the samples, additional design features or shielding should be provided to meet the criteria. A design and operational review of the radiological spectrum analysis facilities shall be performed to determine the capability to promptly (less than 2 hr) quantify certain radioisotopes that are indicators of the degree of core damage. Such radionuclides are noble gases (which indicate cladding failure), iodines and cesiums (which indicate high fuel temperatures), and non-volatile isotopes (which indicate fuel melting). The initial reactor coolant spectrum should correspond to a Regulatory Guide 1.3 or 1.4 release. The review should also conaider the effects of direct radiation from piping and components in the Auxiliary Building and possible contamination and direct radiation from airborne effluents. If the review indicates that the analyses required cannot be performed in a prompt manner with existing equipment, then design modifications or equipment procurement shall be undertaken to meet the criteria. In addition to the radiological analyses, certain chemical analyses are necessary for monitoring reactor conditions. Procedures shall be pro-vided to perform boron ana chloride chemical analyses assuming a highly radioactive initial sample (Regulatory Guide 1.3 or 1.4 source term). Both analyses shall be capable of being completed promptly; i.e., the boron staple analysis within an hour and the chloride sample analysis within a shift. PCE Response Based on preliminary review of the Trojan sampling system and analysis capability following an acciaent, the following actions will be taken: 1. PGE will install shielded sampling facilities to obtain samples of reactor coolant and Containment atmorphere within I hr under accident conditions without incurring a radiation exposure to any individual in excess of 3 and 18-3/4 rem to the whole body or extremities, respec-tively. 2. A mobile low-background counting facility will be estab-lished onsite but outside Plant structures. The facility will contain a Ce-Li gamma spectroscopy system with provi-sions for counting high-activity samples. Procedures will be writt_ a to require the use of this facility if the r in-Plant counting facility is unusable. 1276 253 =.

Page 25 of 49 3. PGE will implement procedures and/or design changes to ensure that borou and chloride analyses of highly radio-active samples can be obtained within 1 hr and one shift, respectively. The 'Jestinghouse Owner's Group is also currently investigating this issue and plans to complete its review by March 1980. PGE will follow the progress of this review through its membership in the Owner's Group and will evaluate the Group's recommendations against PGE's proposed design and procedure changes. If necessary, PGE will revise its design and/or procedures to incorporate the Owner's Group recommendations. Implementation Schedule 1. a. A design review of currantly installed reactor coolant and Containment atmosphere sampling system and a description of the preliminary design of the augmented shielded sampling facilities will be provided by January 1,1980. The facil-ities will be installed by January 1, 1981. b. Current operating procedures for radioactivity, boron and chloride sampling and analysis.will be reviewed and augmented, if necessary, by January 1,1980 to upgrade existing sampling capabilities as much as possible. Oper-ating procedures that reflect the augmented sampling facil-ities will be implemented by January 1,1981. 2. The mobile counting facility will be operational by January 1,1981. 3. Procedures and/or design changes to ensure prompt boron and chloride analysis of highly radioactive samples will be Dnplemented by January 1, 1981. KM/SGG/4kk66.44A24 1276 256

Page 26 of 49 Section 2.1.8.b - Increrded Range of Radiation Monitors NRC Position 1 The requirements associated with this recommendation should be considered as advanced implementation of certain requirements to be included in a revision to Regulatory Guide 1.97, " Instrumentation to Follow the Course of an Accident", which has already been initiated, and in other Regulatory Guides, which will be promulgated in the near term. 1. Noble gas effluent monitors shall be installed with an extended range designed to function during accident conditions as well as during normal operating conditions; multiple monitors are considered to be necessary to cover the ranges of interest. Noble gas effluent monitors with an upper range capacity a. of 105 uCi/cc (Xe-133) are considered to be practical and should be installed in all operating plants. b. Noble gas effluent monitoring shall be provided for the total range of concentration extending from a normal condi-tion (ALARA) concentrations to a maximum of 105 uCi/cc (Ie-133). Multiple monitors are considered to be necessary to cover the ranges of interest. The range capacity of individual c uitors should overlap by a factor of 10. 2. Since iodine gaseous effluent monitors for the accident condition are not considered to be practical at this time, capability for effluent monitoring of radioiodines for the accident condition shall be pro-vided with sampling conductad by adsorption on charcoal or other media, followed by onsite laboratory analysis. 3. In-Containment radiation level monitors with a maximum range of 108 rad /hr shall be installed. A minimum of two such monitors that are physically separated shall be provided. Monitors shall be designed and qualified to function in an accident environment. PCE Response 1. FSAR Section 11.4 describes the noble gas radiation monitoring systems. The Containment Radiation Monitoring System has an existing Ci/cc (Ie-133). The Auxiliary Building upper acble gas range of 300 3 and Condenser Air Discharge Monitoring Systems are being modified to Ci/cc (Xe-133). also have an upper noble gas range of 300 3 The new 300 pCi/cc range is considered to be acceptable for the Condenser Air Discharge Radiation Monitoring Systems. The maximum calculated noble gas release race following a steam generator tube rupture is 140 s i/cc which is well below 300 u i/ce. c c The new 300 uCi/cc range is considered to be acceptable for the Auxiliary Building Radiation Monitoring System for the following reasons : 1276 257

Page 27 of 49 a. The Auxiliary Building Radiation Monitoring System will remain on-scale for calculated recirculation leak rates of up to 70 gpm with 100 percent of the core noble gases retained in the recirculation fluid. This calculation is considered to be conservative since most of the gases will not be retained in the recirculation liquid but will remain inside the Containment. A leak race of 70 gpm is considered to be exceptionally large in view of the leakage control program instituted in response to NaC Position 2.1.6.a. b. The Auxiliary Building Radiation Monitoring System will remain on-scale for calculated CVCS leakage of up to 15 gpm with 100 percent of the core noble gases retained in the reactor coolant. This calculation is considered to be conservative because it is not conceivable that 100 percent of the core noble gases would be contained within the reactor coolant fluid alone. Also, a leak rate of 15 spm in the CVCS is considered unlikely in view of the leakage control program instituted in response to NRC Position 2.1.6.a. The Auxiliary Building Radiation Monitoring System will c. remain en-scale for the other accidents described in FSAR Chapter 15. The 300 yCi/cc range is considered to be acceptable for the Containment Radiatien Monitoring System for the following reasons : High activity discharges will not occur via the Contain-a. ment Purge System since the redundant isolation valves on each line in this system will be closed by the Con-tain=ent I. solation 3ignal or if high radiation levels are detected by the Containment Radiation Monitoring System. b. Radioactivity levels within the Containment can be deter-mined using the in-Containment radiation level monitors described in 3 below, c. The Containment Radiation Monitoring System would be used to monitor releases from the Hydrogen Vent System. However, since two redundant hydrogen recombiners are provided inside the Containment, it is very unlikely that the Hydrogen Vent System will ever be used follow-ing an accident. If the Hydrogen Vent Sy3 tem is used it will not be until 8.6 days after the accident (FSAR Section 6.2.5.3.4). At this time fellowing the accident, the concentration of Xe-133 inside the containment is conservatively calculated to be 1160 uCi/cc assuming 100 percent of the cora noble gases are released to the Containment atmosphere. Since the calculation is con-sidered to be very conservative, it is reasonable to 1276 258

Page 28 of 49 assume that the actual concentration will be less than 300 uCi/cc. Furthermore, grab samples can be taken and analyzed in accordance <ith PGE Response 2.1.8.b in the unlikely event the concentrations exceed the monitor range during the venting. The higher range detectors which are to be installed in the Auxiliary Building and Main Condenser Air Discharge Monitoring Systems will comply with the design criteria for those radiation monitoring systems specified in the FSAR Sectioa 11.4. 2. PCE Response 2.1.8.b, Position 2, addresses the procedure to be implemented for determining concentrations of radioiodine in effluents. In addition, existing Radiological Emergency Response Plan procedures require field monitoring for racioiodines in the event of an accident. No additional action is considered necessary. 3. Two physically separated high-range radiation monitors will be provided to detect post-accident radiation levels in the Contain-ment. Each monitor will have a range from 10 R/hr to 107 R/hr for gamma radiation maximum capacity will be equivalent to an integrated 5 R/hr for beta and ga=ma radiation. Detectors and dose rate of 10 associated cables up to the penetrations will be qualified to with-stand LOCA conditions discussed in FSAR Section 3.11.2. This moni-coring system will be designed and qualified to the existing Seismic Category I design criteria specified in the FSAR and provided with noninterruptible power. Each Containment monitor will be provided with indicators and recorders in seismically qualified panels in the control room and an integral electronic input to perform on-line testing. Implementation Schedule The high-range radiation detectors for the Auxiliary Building and Main Condenser Air Discharge Radiation Monitoring Systems will be installed by the startup of Cycle 3 in 1980. Based on currently estimated equipment delivery dates for the high-range Containment radiation monitor, installation will commence at the first cold shutdown after September 1, 1980 (assuming that the equipment is delivered according to currer.t estimates), and will be completed no later than startup of Cycle 4 in pring 1981. KM/TDW/4kk66.44A26

Page 29 of 49 Section 2.1.8.b - Increased Range of Radiation Monitors IRC Position 2 Interim Procedures for Ouantifying High Laval Accidental Radioactivity Releases Licensees to implement procedures for estimating noble gas and radio-iodine release rates if the existing effluent instrumentation goes offscale. PGE Response Current Trojan plant procedures require that grab samples of effluent streams be taken and analyzed by gamma spectroscopy if the Process and Effluent Monitors (PRMS) are offscale or nonfunctional. In addition, site boundary field surveys for external dose rate and gross iodine air concentration are required following accidental releases of radioactive effluents. In addition to the above-existing procedures, PGE will implement the following interim procedures for estimating noble gas and radiciodine release rates if the existing instrumentation goes offscale: 1. Noble Gases : A grab sample of the effluent from the appropriate Process and Effluent Radiation Monitor (PRM) will be analy:ed on in-Plant Ge-Li detector. If the in-Plant Ge-Li detector cannot be used because of high background or high sample activity, this sample will be surveyed with portable moni-toring equipment. Calibration factors will be developed to allow calculation of equivalent Ie-133 effluent concentra-tiors frem these measurements. Plant procedures will be developed to implement this monitoring method, including minimizing occupational exposures through remote handling of samples. 2. Radioiodines: Plait emergency procedures are currently being modified to re',uire measurement of a grab sample using portable moni-toring equipment if the PRM reads offscale and the in-Plant Ge-Li detect:r cannot be used due to high background or high sample activity. Modifications to Plant procedures will be implemented to require the use of silver zeolite cartridge as an iodine er, ping agent if interference from noble gases is suspected. Calibration factors will be developed to allow calculation of equivalent I-131 effluent concentrations from portable survey instrument measurements. i276 260

Page 30 of 49 In addition to the above procedures, a procedure is being developed to allow the estimation of equivalent Xe-133 concentrations in the Containment using internal Area Radiation Monitor or external ge=ma dose rate measurements at specified locations outside the Containment. This procedure will be used as a backup to effluent monitors, grab samples, and field monitoring data. Imolementation Schedule The above procedural changes and the use of portable monitors will be implemented by January 1,1980. SGG/gah/4mg66.44A29 7

Page 31 of 49 Section 4.. 8.c - Improved In-Plant Iodine Instrumeni.ation NRC Position Each licensee shall provide equipment and associated training and procedures for accurately determining the airborne iodine concentration in areas within the facility where plant personnel may be present during an accident. PGE Response Current Trojan plant design includes the capability to analyze air samples for iodine using gamma ray spectroscopy in the counting room (ou Control Building Elevatior 45 ft). Plant procedures specify this method for final decarmisation of iodine levels in Plant air samples. In order to ensure the capability to analyze samples following an acci-dent, when the present counting room is rendered unusable by high-back-ground radiation levels due to airborne contamination, a mobile low-background counting facility will be established as discussed in Section 2.1.8.a. The mobile facility equipped with a Ge-Li gamma spectroscopf system with provisions for counting high activity sources normelly will be located onsite, but outside Plant buildings, and will be capable of being moved to ensure a sufficiently low background. Plant procedures will be written to implement the use of the backup facility. Plant procedures will be written to specify the use of silver zeolice cartridges for post-accident iodine air sampling if possible interfer-ecce from noble gases is suspected. As an interim measure, until the mobile counting facility is establi. hed, a backup counting facility will be established at the Visitors Inforua-tion Center (which is the Emergency Control Center following accidents). The equipment to be used consists of: 1. A 256 multichannel analyzer with a 2 x 2 NaI detector; and 2. A single channel analyzer with spectral scanning capa-bility with a 2 x 2 NaI detector. Procedures for calibrating and using this equipment will be developed. Implementation Schedule 1. The mobile counting facility will be operational by January 1, 1981. 2, The procedures for specifying the use of silver zeolite for post-accident iodine sampling will be implemented by January 1,1980. 3. The interim backup counting facility at the Visitors Infomation Center will be operational by January 1,1380. 1276 262 KM/SGG/4kk66.31

Page 32 of 49 Section 2.1.9 - Analysis of Design and O' dorsal Transients and Accidents NRC Position 1 Analyses, procedures, and training addressing the following are required: 1. Small break tass-of-coolant accidents ; 2. Inadequate core cooling; and 3. Transients and accidents. Some analysis requirements for small breaks have already been specified by the Bulletins and Orders Task Force. These should be completed. In . addition, pretest calculations of some of the Loss of Fluid Test (LOFT) small break tests (scheduled to start in September 1979) shall be per-for=ed as means to verify the analyses performed in support of the small break emergency procedures and in support of an eventual long term verification of compliance with Appendix K of 10 CFR Part 50. In the analysis of inadequate core cooling, the following conditions shall be analyzed using realistic (best-estimate) methods: 1. Low Reactor Coolant System inventory (two examples will be required - LOCA witn forced flow, LOCA without forced flow). 2. Loss of natural circulation (due to loss of heat sink). These calculations shall include the period of time during which inade-quate core cooling is approached as well as the period of time during which inadequate core cooling exists. The calculations shall be carried out in real time far enough that all important phenomena and instrument indications are included. Each case should then be repeated taking credit for correct operator action. These additional cases will provide the basis for developing appropriate emergency procedures. These calcu-lations should also provide the analytical basis for the design of any additional instrumentation needed to provide operators with an unambigu-ous indication of vessel water level and core cooling adequacy (see Section 2.1.3.b in this appendix). The analyses of transients and accidents shall include the design basis events specified in Section 15 of each FSAR. The analyses shall include a single active failure for each system called upon to function for a particular event. Consequential failures shall also be considered. Failures of the operators to perform reovired control manipulations shall be given consideration for permutations of the analyses. Operator actions that could cause the complete loss of function of a safety system shall also be considered. At present, these analyses need not address passive failures or multiple system failures in the short term. In the recent analysis of onsil break LOCAs, complete loss of auxiliary feed-water was considered. The complete loss of auxiliary feedwater may be 1276 263

Page 33 of 49 added to the failures being considered in the analysis of transients and accidents if it is concluded that more is needed in operator training beyond the short-term actions to upgrade Auxiliary Feedvater System reliability. Similarly, in the long tem, multiple failures and passive failures may be considered depending in part on staff review of the results of the short-tem analyses. The transient and accident analyses shall include event tree analyses, which are supplemented by computer calculations for those cases in which the system response to operator actions is unclear or these calculations c id be used to prr..de impor" ant quanticate:4 information not available fr m an event tree. For example, failure to initiate high-pressure injection could lead to core uncovery for somt transients, and a computer calculation could provide infomation on the amount of time available for correctisi action. Reactor simulators may provide some information in defining te event trees and would be useful in studying the infomation available to the operators. The transient and accident analyses are to be perfomed for the purpose of identifying appropriate and inappropriate operator actions relating to important safety considerations such as natural circulation, prevention of core uncovery, and prevention of more serious accidents. The infomation derived from ths preceding analyses shall be included in the planc emergency procedures and operatoi-training. It is expected that analyses perfomed by the NSSS vendorr will be put in the fom of emergency procedure guidelines and that the.hanges in the procedures will be implemented by each licensee or applicant. In addition to the analyses performed by the reactor vendors, analyses of selected transients should be perfomed by the NRC Office of Research, using the best available computer codes, to provide the basis for compari-sons with the analytical methods being used by the reactor vendors. These comparisons together with comparisons to data, including LOFT small break test data, will constitute the short-tem verification effort to assure the adequacy of the analytical methods being used to generate emergency procedures. PGE Response Analyses of small break loss-of-coolant accidents, symptoms of inadequate core cooling and required actions to restore core cooling, and analysis of transi' t and accident scenarios, including operator actions not previously analyzed, are being perfomed on a generic basis by the Westinghouse Owner's Group, of which Portland General Electric Company is a member. The small break analyses have been completed and are reported in WCAP-9600, which was submitted to the Bulletins and Orders Task Force by the Owner's Group on June 29, 1979. Incorporated in that report were procedure guidelines that were developed as a result of the small break analyses. These guidelines have been reviewed with the B&O Task Force and will be presented to the Owner's Group utility representatives at a seminar to be held on October 16-19, 1979. Following this seminar, each utility will develop plant-specific procedures and train their personnel on the new procedures. 1276 264

Page 34 of 49 The work required to address the other two areas of concern - inadequate core cooling and other transients and accident scenarios - is being performed in cociunction with the Bulletins and Orders Task Force, including establishment of information requirements to meet the schedule specified in Enclosure 6 to Mr. Eisenhut's letter dated September 13, 1979. Analyses related to the definition of inadequate core cooling and guidelines for recognizing the symptoms of inadequate core cooling based on existing plant instrumentation, and recovery from such a con-dicion will be provid'_4 to the B&O Task Force by October 31, 1979. Further work to better define the approaches to inadequate core cooling and recovery operatiors ./ be required and will be performed later. In the course of performing these analyses and developing corresponding operator guidelines, it is possible that a need for additional instrumenta-tion be will identified. Should this occur, we will notify you of our schedule for the procurement and installation of any instrumentation for which a naad is identified. The Owner's Group is also providing a pretest prediction analysis of the LOFT L3-1 nuclear small break experiment. This analysis will be sub-mitted by the required date of November 15, 1979, in accordance with the schedule established by th9 B&O Task Force. Implementation Schedule 1. It is intended that revised procedures and training for coping with small break LOCAs will be in place by December 31,1979 in accordance with the requirement of Enclosure 6 of Mr. Eisenhut's letter of Sepcember 13, 1979. 2. It is intended that guidelines for coping with inadequate core cooling vill be provided by October 31, 1919 and revisad pro-cedures and training wi'.i be in place by January 1,1930. These dates meet the requirements of Enclosure 6 of Mr. Eisethut's letter of September 13, 1979. 3. It is intended that the work related to other transients and accidents in Chapter 15 of the Trojan FSAR will be completed in es.rly 1980 in accordance with the requirement of Ecciosene 6 of Mr. Eisenhu:'s letter of September 13, 1979. KM/CJP/4kk66.4432 1?76 265

Page 35 of 49 Section 2.1.9 - Analysis of Design and Off-Normal Transients and Accidents NRC Position 2 Consistent with satisfying the requirements set forth in General Cri-terion 13 to provide the capability in the control room to ascertain Containment cenditions during the course of an accident, the following r4quirements shall be bnplemented: 1. A continuous indication of Containment pressure shall be provided in the control room. Measurement and indication capability shall include three times the design pressure of the Containment for concrete, four times the design pressure for steel, and -5 psig for all containments. 2. A continuous indication of hydrogen concentration in the Containment atmosphere shall be provided in the control room. Measurement capability shall be provided over the range of 0 to 10 percent hydrogen concentration under both positive and negative ambient pressure. 3. A continuous indication of Containment water level shall be provided in the control room for all plants. A narrow range instrument shall be provided for PWRs and cover the range from the bottom to the top of the Containment sump. Also for PWRs, a wide range instrument shall be provided and cover the range from the bottom of the Containment to the elevation equivalent to a 500,000 gal capacity. For BWRs, a wide range instrument shall be provided and cover the range from the bottom to 5 ft above the normal water level of the suppression pool. The Containment pressure, hydrogen concentration and wide range Contain-ment water level measurements shall meet the design and qualification provisions of Regulatory Guide 1.97, including qualification, redundancy, and testability. The narrow range Containment water level measurement instrumentation shall be qualified to meet the requirements of Regulatory Guide 1.89 and shall be capable of being periodically tested. PCE Response 1. Trojan currently has four safety-grade pressure transmitters for measuring Containment pressure. They are Barten Model 393 units, mounted i= mediately outside Containment and calibrated for a range of 0 to 75 psig. The transmitters are physically separated and are provided with power from separate safety grade emergency / vital power systems buses. The 4-20 mA outputs are displayed on safety grade main control board indicators which are scaled 0 to 75 psig. 1276 266

Page 36 of 49 Two new Barton dP transmitters will be physically connected in parallel to two (Channels A and B) of the existing four trans-mitters. Indicators, calibraced from -10 to +190 psig, will be provided on panels C19. Safety-grade cable runs between each new transmitter and the new control room indicator and qualified power supplies will be installed. 2. The current Trojan design complies with the NRC position. Trojan has redundant, Seismic Category I post-accident Hydrogen Analysis System panels. This system is discussed in FSAR Section 6.2.5 and illustrated in FSAR Figure 6.2-48. Analysis of the hydrogen gas is performed in each panel by the use of a Delphi thermo-conductivity analyzer. The hydrogen analyzer has a range of 0 to 10 percent hydrogen in air. The Hydrogen Analysis System is presently capable of running continuously through the Containment pressure range of 0 - 60 psig. Each control panel transmits a signal through Class lE cable to a remote indicator which is mounted on the Seismic Category I panel C41 in the control room. The indicators are safety grade. 3. Narrow range level indication - The existing level switches mounted in each of the two Containment sumps are not qualified for the post-LOCA environment. The Barton Model 764 trans-mitter has been qualified in accordance with IEEE 323-1974 and IEEE 344-1975 for the LOCA environment. We in' tend to mount one such transmitter, with Dow Corning 704-filled reference leg, in each Contain.nt sump with an upper range to Elevation 53 ft (approximate 500,000 gal water level). The design of the transmitters will be in accordance with the Trojan licensing commitment in the FSAR. Power for the units will be from the Class IE b ses and each 4-20 mA transmitter output will be routed to a Seismic Category 1 indicator mounted on Control Room Panel C19. Wide-range level indication - Two Barton Model 764 dP trans-mitters, with Dow Corning 704-filled reference legs, will be mounted near the bottom of the reactor cavity. Power for the transmitters will be from safety-grade buses which meet the r*quirements for Trejan licensing commitrent in the FSAR. Each transmitter outpve will be routed to an indicator mounted on Control Room Pacal C19. ihe bottom of the reactor cavity is at Elevation 16 ft 2 in. The upper sange of this instrument is at Elevation 53 ft (approximate 500,000 gal water level). Below this level is a drain sump, vita an existing (non-qualified) level instrument that alarms on high level on radvaste panel C-151. 1276 267

Pcge 37 of 49 Implementation Schedule Based on currently estimated equipment delivery dates, *.nstallation will commence at the first cold shutdown after September 1,1980 (assuming that the equipment is delivered according to current estimates) and will be completed no later than startup of Cycle 4 in Spring 1981. 7g }68 KM/4kk66.44BS

Page 38 of 49 Section 2.1.9 - Analysis of Design and Off-Normal Transiecis and Accidents NRC Position 3 Each applicant and licensee shall install Reactor Coolant System and reactor vessel head high point vents remotely operated fres the control room. Since these vents form a part of the reactor coolant pressire boundary, the design of the vents shall conform to the requiremen: s of Apper.dit A to 10 CFR Part 50 General Design criteria. In particu ar, thess vents shall be safety grade, and shall satisfy the single f ailure crittrion and the requirements of IEEE-279 in order to ensure a tow proba5ility of inadvertent actuation. Each applicaat and lir see shall provide the following information coreerning the desir,4 and operition of these high point vents: 1. A description of the c anstruction, location, size, and power supply for the vents along with results of analyses of Loss-of-Coolant Accidents initiated by a break in the vent pipe. The results of the analyses should be demon-strated to be acceptable in accordance with the acceptance criteria of 10 CFR 50.46. 2. Analyses demonstrating that the direct venting of noncon-densible gases with perhaps high hydrogen concentrations does not re: ult in violation of combustible gas concen-tration 1Lnits in Containment as described in 10 CFR Part 50.44, Regulatory Guide 1.7 (Rev. 1), and Standard Review Plan Seccion 6.2.5. 3. Procedural guidelines for tae operator's use of the vents. The information available to the operator for iniciating or terminating vent usage shall be discussed. PGE Response The pressurizer power-operated relief valves (PORVs), as presently designed, provide a safety-grade capability for venting noncondenaibles from the pressurizer steam spree. Power to operate the PORVs and their associated block valves is ptorided from safety-related circuits,.ss described in our response e s Itism 2.1.1. This design is such that venting of noncondensibles from the pressurizer would be possible even assuming a single failure of one of the two PORVs or block valves or one emergency power train. The Chemical and Voluue Control System can also be utilized for hydrogen reduction by bleed a*2d feed degasification via the volume control tank. The pressure boundaries, but not necessarily the control systems, for this CVCS function meet Seismic Category I requirements. We are currently developing the design for a remotely-operated reactor pressure vessel head vent using the existing manual head vent connection. 1276 269

Page 39 of A9 It is intended that the system design vill meet the NRC requirements set forth in NUREG-0578 and at the NRC Topical Meeting on PCS venting on October 11, 1979 (nonredundant but otherwise safety grade in accordance with the "_ ojan FSAR, with line size smaller than that corresponding to the normal system makeup capability, discharge to a well-mixed CoStain-ment atmosphere location, 10 CFR 50/ Regulatory Guide 1.7 hydrogen source term). Implementation Schedule A description of the design for the Reactor Vessel Venting System wila be provided by January 1, 1980. It is currently intended that the system will be installed, and operating procedures will be developed and imple-manted by January 1, 1981. KM/CJP/4kk66.44B8 i?76 270

Page 40 of 49 Section 2.2.1.a - Shift Supervisor's Responsibilities NRC Position 1. The highest level of corporate management of each licensee shall issue and periodically reissue a management directive that emphasizes the primary management respensibility of the shift supervisor for same operation of the plant under all conditions on his shift and that clearly establishes his connand duties. 2. Plant procedures shall be reviewed to assur ; that the duties, respon ' sibilities, and authority of the shift supervisor and control room operators are properly defined to effect the eetablishment of a defi-nite line of command and clear delineation of the ceccand decision authority of the shift supervisor in the control room relative to other plant management personnel. Particular emphasis shall be placed on the following: a. The responsi~aility and authority of the shif t supervisor shall be to maintain the broadest perspective of operational conditions affecting the safety of the plant as a matter of highest priority at all times when on duty in the control room. The idea shall be reinforced that the shift supervisor should not become totally involved in any single operation in times of emergency when multiple operations are required in the control roem. b. The shift supervisor, until properly relieved, shall remain in the control room at all times during accident situations to direct the activities of control room operators. Persons authorized to relieve the shift supervisor shall be spe:ified. c. If the shif t supervisor is temporarily absent from the control room during routine operations, a lead control room operator shall be designated to assume the control room consand tanc-tion. These temporary duties, re ponsibilities, and authority shall be clearly specified. 3. Training programs for shift supervisors shall emphasize and reinforce the responsibility for safe operation and the management function the shif t supervisor is to provide for assuring safety. 4. The administrative duties of the shift supervisor shall be reviewed by the senior officer of each utility responsible for plant operations. Administrative functions that detract from or are subordinate to the management recponsibility for assuring che safe operation of the plant shall be telegated to other operatious personnel not on duty in the control room. 1276 271

Page 41 of 49 ?CE Response 1. The Assistant Vice President for Thermal Plant Operations and Maintenance will issue, and reissue annually thereafter, a Management Directive emphasizing the Shift Supervisor's manage-ment responsibilities for safe operation of the Plant and command duties which will be defined in appropriate Trojan plant Administrative Orders. The Management Directive will be in the form of a formal documentation to the Plant management and Shif t Supervisors. A copy of the directive will be kept in the file at the Trojan plant and will be available for review by the NRC upon request. 2. Administrative Orders currently exist at the Trojan plant which identify the responsibilities of Shift Supervisors and Control Room Operators (i.e., Administrative Order, A0-1-4, " Operations Responsibilities"). In light of the concerns raised in NUREG-0578, these Administrative Orders will be reviewed to ensure that duties, responsibilities, and authority of the Shift Supervisors and Control Room Operators are clearly defined for the line of command and the command-decision authority. Particular emphasis will be placed on the responsibility of the Shift Supervisor to maintain a broad perspective of operational conditions affecting the safety of the Plant as a matter of highest priority at all times, without totally involving himself in any single opera-tion of the Plant. During accident situations, the Shift Supervisor may not be able to remain in the control room at all times, since his presence at the scene of some types of emergencies is essential to ensure safe operation of the Plant. Criteria governing his absence from the control room and delineating persons authorized to relieve him will be specified in the Administrative Orders. The Administrative Orders will also be reviewed to ensure that the Control Room Operator shall be designated for the command function when the Shif t Supervisor is abseat from the control room during routine operations. The temperary duties, responsi-bilities and authority of the designated Control Room Operator will be clearly specified in the Orders. 3. Indoctrination of Shif t Supervisors will emphasize and reinforce their respcasibilities and the management function for safe operation. 4. The administrative duties of the Shift Supervisor will be reviewed by the Assistant Vice President, Thermal Plant Opera-tions and Maintenance, on the same interval as the issuance of the Management Directives. His review will ensure that admin-istrative functions that detract from or are subordinate to the management responsibility for assuring safe operation of the Plant will be delegated to other personnel not on duty in the control room. 1276 272

Page 47 *of 49 I=olementation Schedule The first Management Directive and review of administrative duties for the Shift Supervisor by the Assistant Vice President. Thermal Plant Operations and Maintenance, will be completed on or before January 1, 1980. The Administrative Orders will be modified and the indoctrination of the Shif t Supervisor will be completed by January 1,1980. 1276 273 Ef/RPB/gah/4mg66.44Bil

• 4 Page 43 of 49 Section 2.2.1.b - Shif t Technical Advisor NRC Position Each licensee shall provide an on-shif t technical advisor to the shif t supervisor. The shift technical advisor may serve more than one unit at a multi-unit site if qualified to perf arm the advisor function for the various units.

The shift technical advisor shall have a bachelor's degree or equivalent in a scientific or engineering discipline and have received specific training in the response and analysis of the plant for transients and accidents. The shift technical advisor shall also receive training in plant design and layout, including the capabilities of instrumentation and controls in the control room. The licensee shall assign nor 31 duties to the shift technical advisors that pertain to the engineering aspects of assuring S sfe operations of the plant, including the review and evaluation of operating experience. PCE Response PGE will provide a Shift Technical Advisor onsite 24-hr per day to provide on-shift technical advice to the Shif t Supervisor and to provide the operating expe;ience assessment. The Shif t Technical Advisor will have a bachelors degree or equivalent in a scientific or engineering discipline and will receive specific train-ing in the response and analysis of the Plant for transients and accidents, plus training in Plant design and layout, including instrumentation and control capabilities in the control room; or in lieu of a bachelors degree, he vill have an SRO license. PGE does not believe that the above NRC requirement enhances the ability of the Shif t Supervisors er the Plant Operators to properly respond to an accident. The short-term actions required to properly respond to'an accident rely on a well-qualified and well-trained operating crew (Shift Supervisor and Operators). Interjection of advice, commenta, etc, during the initial short-term response actions will in all likelihood add confusion and may decrease the probability of properly responding to an accidsuc. For the proper long-term responses, the Shift Supervisor has the benefit of relying on qualified Plant Engineers and on Plant Managers who are always on call. Implementation Schedul,e The above-mentioned PGE position will be implemented by January 1,1980. Additional details concerning the STA will be provided by November 17, 1979, as we move towards implementation. We intend to reassess these plans following experience gained during 1980; any changes will be provided to the NRC prior to implementation. CPY/kk/kng66.44B14

Page 44 of 49 Section 2.2.1.c - Shift and Relief Turnover Procedures NaC Position The licensees shall review and revise as necessary the plant procedure for shift and relief turnover to assure the following: 1. A checklf st shall be provided for the encoming and offgoing control room operators and the onciming shif t stpervisor to complete and sign. The following f.tems, as a minimum, shall be included in the checklist: a. Assurance that critical plant parameters are within allowable limits (parameters and allow-able limits shall be listed on the enecklist). b. Assurance of the availability and proper align-ment of all systems essential to the prevention and mitigation of operacional transients and accidents by a check of the control console (what to check and criteria for acceptable status chall be included on the checklist). c. Identification of systems and components that are in a degraded mode of operation permitted by the Technical Specifications. For such systems and components, the length of time in the degraded mode shall be corpared with the Technical Specifications action scatement (this shall be recorded as a separate entry on the checklist). 2. Checklists or logs shall be provided for completion by the offgoing and oncoming auxiliary opera'. ors and technicians. Such checklists or logs shall include any equipment under maintenance and test that by themselves could degrade a sys-tem critical to the prevention and mitigation of operational transients and accidents or initiate operational transients (what to check and criteria for acceptable status shall be included on the checklisc). 3. A system shall be established to evaluate the effectiveness of the shif t and relief turnover pr'ocedure (for example, periodic independent verification of system alignments). PCE Response 1. Procedures for shift and relief turnover will be revised to specify how critical Plant parameters are verified to be within limits; how assurance of the availability and proper alignment of all systems essencial to the prevention and mitigation of operational transients and accidents is obtained by a check of the control console; and how systems and components that are in a degraded mode of operation permitted by the Technical Specifications are tracked. 1276 275

Page 45 of 49 For such systems and components, the length of time in the degraded mode will be compared with the Technical Specifications ac. tion state-ment. Checklists will be provided in the procedures for implementa-tion of these actions. 2. Maintenance and testing of equipment that by themselves could degrade a system critical to the prevention and mitigation of operational transients and accidents or initiate an operational transient is cur-rently controlled by a maintenance request. A maintenance request is required to provide Plant conditions for the work, instructions for the work, and safety considerations necessary for the work. The work in the maintenance request has to be approved by the Shift Super-visor prior to its initiation. Technicians and maintenance personnel are required to document !.he action taken when following work instruc-tions. The action taken and test results are reviewed and approved by the Shif t Supervisor prior to closing out each request. Therefore, no checklists are necessary for the auxiliary operator and-technician; the checklists used by the control operator and Shift Supervisor will suffice. 3. The effectiveness of turnover procedures and work controls is cur-rently monitored by periodic system line-up checks and operating tests. Additional requirements for independent spot checks of critical system and parameter status during their operation will be Laplemented in order to improve the verification mechanism of system alignments. Implementation Schedule The modification for procedures for shift and relief turnover and the establishment of independent spot checks will be completed and Laplemented by January 1, 1980. KM/gah/4mg66.44B15 ]}7g }Jg

Page 46 of 49 Section 2.2.2.a - Control Room Access NRC Position The licensee shall make provisions for limiting access to the control room to those individuals responsible for the direct operation of the nuclear power plant (e.g., operations supervisor, shif t supervisor, and control room operators), to technical advisors who may be requested or required to support the operation, and to predesignated NRC personnel. Provisions shall include the following: 1. Develop and Laplement an administrative procedure that establishes the authority and responsibility of the person in charge of the control room to ILnic access. 2. Develop and implement procedures that establish a clear line of authority and responsibility in the control room in the event of an emergency. Tha line of succession for the person in charge of the control room shall be established and Ibnited to persons possessing a current senior reactor operator's license. The plan shall clearly define the lines of communication and authority for plant management personnel not in direct command of operations, including those who report to stations outside of the control room. PCE Response 1. Current Administrative Order, A0-3-8, " Control Room Opera-tions", at the Trojan plant establishes the authority and responsibility of the person in charge of the ecstrol room to limit acces ;. This Administrative Order will be revised to explicitly state that the Control Operator is responsible and is authorized to restrict access to the control room during both normal operation-d emergencies, although his decision may be overruled by the Shift Supervisor. 2. The existing procedure A0-3-8 vill be modified to include a clear line of authority and responsibility in the control room in the event of an emergency. The modified procedure will specify personnel who will be allowed access during emergency conditions, as well as the line of succession of the person in charge of the control room which will be restricted to persons possessing a current senior reactor operator's license. The line of communication and authority for Plant management personnel not in direct command of operations will also be defined in the procedure. Imolementation Schedule Administrative Order A0-3-8 incorporating the above changes will be revised and Laplemented by January 1,1980. KM/ RPS/gah/4mg66.44317 1276 277

Page 47 of 49 Section 2.2.2.b - Onsite Technical Support Center NRC Position Each operating nuclear power plant shall maintain an ensite Technical Support Center (TSC) separate from and in close proximity to the control room that has the capability to display and transmic plant status to those individuals who are knowledgeable of and responsible for engineer-ing and management support of reactor operations in the event of an accident. The center shall be habitable to the same degree as the control room for postulated accident conditions. The licensee shall revise his emergency plans as necessary to incorporste the role and location of the technical support center. Records that pertain to the as-built conditions and layout of structures, systems and components shall be stored and filed at the site and acces-sible to the technical support center under emergency conditions. Exam-ples of such records include system descriptions, general arrangement drawings, piping and instrument diagrams, piping system isometrics, elec-trical schematics, wire and cable lists, and single line electrical diagrams. It is not the intent that all records described in ANSI-N45.2.9-1974 be stored and filed at the site and accessible to the technical support center under emergency conditions; however, as stated in that standard, storage systems shall provide for accurate retrieval ef all pertinent information without undue delay. PGE Response PGE will establish an onsite Techaical Support Center (TSC). This center will be provided with the following: 1. Sufficient space to accommodste approximately 25 key technical personnel; 2. Communication links to required locations such as the control room and Emergency Control Center (ECC); 3,. Appropriate Plant instrument results; and 4. Copies of appropriate Plant drawings and procedures. The TSC will consist of a well-engineered structure that is habitable for a realistic spectrum of potential nuclear plant accidents; in particular, during the diagnostic period prior to potentially significant radioactive material releases. The control room, the Company Control Center (in Forciand) and the ECC (at the VIC) serve as backup facilities to the TSC in the event to significant radioactive material releases resulting in deactivation of the TSC. Until the permanent TSC is established, the Plant Administration Building vill be designated as ihe TSC. The Administration Building is located in close proximity to the control room, has telephone and in-Plant communica-tion to the control room and ECC, and contains copies of the Plant drawings. 1276 278

Page 48 of 49 The Tro'4an Radiological Emergency Response Plan vill be revised to reflect the role of the TSC. Imolementation Schedule By January 1, 1980: 1. The role, location, communication facilities, and drawing storage at the interim TSC will be established; 2. The design criteria for the p ;aanent TSC, including infor-mation stored there, will be determined; 3. The Radiological Emergency Response Plan will be revised to reflect the TSC. We vill endeavor to establish a penadnent TSC by January 1,1981; however, since the design criteria and the design of this facility is currently in an early stage, it may not be possibli to complete all work by this date. Additional schedular data vill be provided as it becomes available. KM/TDW/kk/4eg66.44Bla 1276 279

Page 49 of 49 Section 2.2.2.c - Onsite Operational Support Center NRC Position An area to be designated as the onsite opt ratioral support center shall be established. It shall be separate from the control room and shall be the place to which the operations support personnel vill report in an emergency situation. Consunications with the control room shall be provided. The emergency plan shall be revised to reflect the existence of the center and to establish the methods and lines of communication and management. PCE Response To reduce congestion in the control room, PGE will sacablish an onsite Operational Support Center (OSC) in the Control Room viewing gallery and proximate area of the Turbine Building operating deck. The following personnel vill report to the OSC:

1. Designated operations personnel.

2. Designated maintenance and instrument control personnel

3. Designated radiation protection personnel. (Certain radiation protection personnel will report to the Access control Point if it is habitable to ensure positive access control of potential high radiation areas within the Auxiliary and Fuel Building.)

All other support personnel will report tc the Visitors Information Center (VIC) in the event of a site evacuation. These support personnel may be dispatched from the VIC to the OSC if additional personnel are required in the Plant. The OSC will be provided with communications to the control room. The Trojan Radiological Emergency Response Plan will be revised to reflect the OSC. Implementation Schedule By January 1. 1980: 1. The OSC will be established; and 2. The Radiological Emergency Response Plan will be revised to reflect th' OSC. 1776.28H0 KM/TDW/kk/4mg66.44B20 e

a ATTACHMENT 2 Page 1 of 3 Responses to Near-Term Requirements for Improving Emergency Preparedness of D. G. Eisenhut Letter of September 13, 1979 1. NRC Position Upgrade licensee emergency plans to satisfy Regulatory Guide 1.101, with special attention to the development of uniform action level criteria based on plant parameters. PGE Responge PGE and applicable State and local government agencies are currently in the process of reviewing the Trojan Radiological Emergency Response Plan (RERP) against the recommendation of RG 1.101, including protec-tive action level criteria. A meeting with the NRC on this subject is scheduled for October 22-24, 1979. We intend to revise the RERP to incorporate the requirements of Regulatory Guide 1.101 as soon as possible, but in any event, no later than by July 1980. 2. NRC Position, Assure the Laplementation of the related recommendations of the lassons Learnsd Task Force involving instrumentation to follow the course of an accident and relate the information provided by this instrumentation to the emergency plan action levels. This will include instrumentation for post-accident sampling, high range radioactivity monitors, and improved in-Plant radioiodine instru-mentation. The inplementation of the Lessons Learned Task Force's recommendations on instrumentation for detection of inadequate core cooling will also be factored into the emergency plan action level criteria. PGE Response Attachment I addresses PGE response to the NRC Lessons Learned Task Force recommendations. Information relating Plant accident instrumen-tacion readings to emergency action levels will be implemented by July 1980 for currently installed instruments and by the operational date for new or augmented instruments. 3. NRC Position Determine that an emergency operations center for Federal, State and local personnel has been established wi*.h suitable communications to the plant, and that upgradicg of the facility in accordance with the Lassons Learned Task Force's recommendation for an in-Plant technical support center is underway. PGE Resconse A Decision Center has been designated by PGE and the State of Oregon for use by senior PGE, State, local and Federal officials as an 1276 281

Page 2 of 3 Emergency Operations Center (EOC). A final location for the Decision Center has not been determined, but it will be a designated area in either the State Capitol in Salem or at the State Office Building in Portland. Dedicated and redundant communication links will be provided between the Decision Center and the PGE, State and local emergency centers during the first quarter of 1980. The EOC is a separate facility from the in-Plant Technical Support Center. The EOC is a facility remote from the Plant for decision makers to assemble and for decisions to be made based on information that is communicated to it. Special design features are not anticipated other than for communications equipment and building space. 4. NRC Position Assure that improved licensee offsite monitoring capabilities (includ-ing additional thermoluminiscent dosimeters or the equivalent) have been provided for all sites. PGE Response Additional thennoluminescent dosimeters (TLDs) have been placed around the Trojan site to improve post-accident offsite monitoring capability. These TLD's are located at or near site boundary loca-tions, and at 5 miles and 10 miles distance in each 16.5 degree sector around the site. 5 NRC Position Assess the relationship of State / local plans to the licensees' and Fe Aoral plans so as to assure the capability to take appropriate emergency actions. Assure that this capability will be extended to a distance of 10 miles. This item will be performed in conjunction with the Office of State Programs and the Office of Inspection and Enforcement. PGE Response The State of Washington's emergency plan has received NRC concur-rence. The State of Oregon intends to receive NRC concurrence by mid-1980. PGE will work with the State of Oregon in their efforts to obtain NRC concurrence and will work with both states in their efforts to incorporata upgraded NRC criteria intp their plans, includ-ing emergency action planning out to 10 miles. 6. NRC Position Require test exercises of approved emergency plans (Federal, State, local and licensees), review plans for such exercises, and participete in a limited number of joint exercises. Tests of licensee plans will be required to be conducted as soon as practical for all facilities and before reactor startup for new licensees. Exercises of State 1276 282

= 6 Page 3 cf 3 plans will be performed in e junction with the concurrence reviews of the Office of State Programs. As a preliminary planning bases, assume that joint test exercises involving Federal, State, local and 1!censees will os conducted at the rate of about 10 per year, which would result in all sites being exercised once s&ch 4 years. Revised plannirg guidance may result from the ongoing rulemaking. PCE Resposse PGE is committed to condu: ting test exercises annually. The uszt exercise, which will involve 3 tate and local officials, is scheduled for late October 1979. PGE will participate in the proposed joint exercise. TDW/gah/4mg66.44B24 1276 283}}