ML16340B617

Safety Evaluation Report Related to the Operation of Diablo Canyon Nuclear Power Plant,Units 1 and 2.Docket Nos. 50-275 and 50-323.(Pacific Gas and Electric Company)

ML16340B617
Person / Time
Site:	Diablo Canyon
Issue date:	04/30/1981
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-0675, NUREG-0675-S14, NUREG-675, NUREG-675-S14, TAC-51638, TAC-51994, NUDOCS 8104160701
Download: ML16340B617 (84)

Text

NUREG-0675 Supplement No. 14 Safety Evaluation Report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2 Docket Nos. 50-275 and 50-323 Pacific Gas and Electric Company Supplemen't'o.

14 U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation April 1981 c~~A lOOq

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4/2/81 TABLE OF CONTENTS 1

TMI-2 RE(UIREMENTS 1.1 Introduction........

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2 FULL-POWER REQUIREMENTS.....................

2-1 I:I.C.1 I. C'. 7 I. C.8 I.G.1 II.

II.B. 1 II.B. 4 II.B. 7 II.B. 8 II.E.1.1 II~ E. 1. 2 II.E.3.1 II.E.4. 2 III.

III.A. 1. 1 III. D. 1. 1 III.D.2.4 III. D. 3. 3 III.D. 3. 4 IY.

Operational Safety..........

Guidance For Evaluation and Development of Procedures For Transients and Accidents........

NSSS Vendor Review of Procedures.................

Pilot Monitoring of Selected Emergency Procedures for Near-Term Operating License Applicants.....

Training During Low-Power Testing................

Siting and Design

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Reactor Coolant System Vents....................

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Training for Mitigating Core Damage..............

Analysis of Hydrogen Control Rulemaking Proceeding on Degraded Core Accidents Auxiliary Feedwater System Reliability Evaluation Auxiliary Feedwater Initiation and Indication....

Emergency Power Supply for Pressurizer Heaters...

Containment Isolation Dependability..............

Emergency Preparations and Radiation Protection..

Upgrade Emergency Preparedness...................

Primary Coolant Sources Outside Containment......

Offsite Dose Measurements Improved Inplant Iodine Instrument Under Accident Copditions.........................,..

Control Room Habitability........................

Practices and Procedures.........................

2-1 2"1 2-2 2-2 2-4 2-6 2-6 2-6 2-8 2"9 2-10 2-19 2-19 2-21 2-26 2-26 2-26 2-28 2-28 2-29 2"31 DATED RE(UIREMENTS....

3-1 I.

I.D.1 I.D.2 II.

II.B.2 II.B.3 II.D. 1 II.F.1 II.F.2 Operational Safety...........

Control Room Design Review..

Plant Safety Parameter Display Console........

Siting and Design.............

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Plant Shielding Postaccident Sampling.-........................

Performance Testing of Boiling Water Reactor and Pressurized Water Reactor Relief and Safety Valve...........................,.....

Additional Accident Monitoring Instrumentation Inadequate Core Cooling Instruments...........

3"2 3"2 3-2 3"4 3-4 3-9 3-12 3-14 3-17

II.K. 2 II.K.3 III.

III.A.1. 2 III.A.2 4

CONCLUSIONS TABLE OF CONTENTS (Continued)

Commission Orders on Babcock and Wilcox Plants (II.K.2. 13, II.K. 2.17 and II.K.2.19)......:..

Final Recommendations of B80 Task Force (II.K. 3.

II.K. 3. 5, II.K. 3. 25, and II.K. 3. 31)..........

Emergency Preparations and Radiation Protection Upgrade Emergency Support Facilities...........

Long-Term Emergency Preparedness...............

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3-20 3"21 3-23 3-23 3-24 4-1 APPENDIX A SUPPLEMENT TO THE CHRONOLOGY OF THE RADIOLOGICAL SAFETY REVIEW A-1 APPENDIX B EMERGENCY PREPAREDNESS EVALUATION REPORT...................

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- TNI-I REITIIIRENENTE

l. 1 Introduction In a letter dated June 26, 1980, we advised all applicants for construction permits and operating licenses of the Commission's guidance regarding the requirements to be met for current operating license applicationRs.

The require-ments are derived from NRC's Action Plan (NUREG-0660) and are found in NUREG-0694, "TMI-Related Requirements for New Operating Licenses."

The requirements discussed in NUREG-0694 were listed in four, categories:

th'ose required for fuel loading and low power testing; those required for full-power operation; those requiring internal NRC action; and those required to be implemented by a certain date.

Subsequently, by letter dated October 31, 1980, a compilation of those TMI-related items that have. been specifically approved by the Commission for implementation was issued to all licensees and applicants.

This letter trans-mitted NUREG-0737, "Clarification of TMI Action Plan Requirements,"

which included information about'schedules, applicability, method of implementation review, submittal dates, and clarification of technical positions.

Since requirements for fuel loading and low power testing were addressed in Supplement No.

12 to the Diablo Canyon Station Safety Evaluation Report, this supplement only addresses the full power requirements and dated requirements of NUREG-0694 as clarified and supplemented by NUREG-0737.

Each applicable full power requirement and appropriate dated requirements are discussed below and follows the numbering sequence used in NUREG-0694 and NUREG-0737.

, The staff's review of the issues described in this section are based on the explicit requirements contained in NUREG-0694 as updated in NUREG"0737.

Appendix A to this SER supplement is a continuation of the chronology of the principal events involved in the Commission staff s radiological safety review.

Appendix B provides our evaluation of emergency preparedness for the Diablo Canyon Facility.

Based on our review, we find that the applicant meets all the requirements of NUREG-0737 and is acceptable for full power operation pending compliance with the license conditions noted below.

Full Power License Conditions We have identified certain issues in our view where the full power license will be conditioned.

These items listed below are discussed further i,n the sections of this supplement as noted.

1.

NSSS Vendor Review of Procedures (I.C.7) 2.

Training During Low-Power Testing (I.G. 1) 3.

Training for Mitigating Core Damage (II.B.4) 4.

Containment.Isolation Dependability (II.E.4. 2) 5.

Emergency Preparation and Radiation Protection (III)

Moreover, the full power License will also be conditioned, as appropriate, to

'ddress the following TMI dated requirements.

Control Room Design Review (I.D. 1)

Plant Safety.Parameter Display Console (I.D.2).

Plant Shielding (II.B.2)

Postaccident Sampling (II.B.3)

I Performance Testing of Boiling Water Reactor and Pressurized Water Reactor and Pressurized Water Reactor Relief and Safety Valve (II.D.1)

Additional Accident Monitoring Instrumentation (II.F.1)

Inadequate'ore Cooling Instruments (II.F ~ 2)

Commission Or'ders on Babcock and Wilcox Plants (II.K.2.13, II.K.2.17 and II.K. 2. 19)

Final Recommendations of 880 Task Force (II.K.3.2, II.K.3.5, II.K.3.25, and II. K. 3. 31)

I Upgrade Emergency Support Facilities (III.A.1.2)

Long-Term Emergency Preparedness (III.A.2) 1"2

2 FULL-POWER RE UIREMENTS I.

0 erational Safet I. C. 1 Guidance for the Evaluation and Develo ment of Procedures for Transients and Accidents Position In letters of September 13 and 27, October 10 and 30, and November 9, 1979, the Office of Nuclear Reactor Regulation required licensees of operating plants, applicants for operating licenses'nd licensees of plants under con-struction to perform analyses of transients and accidents, prepare emergency procedure guidelines, upgrade emergency procedures, and to conduct operator retraining.

Emergency procedures are required to be consistent with the actions necessary to cope with the transients and accidents analyzed.

Analyses of transients and accidents were.to be completed in early 1980 and implementa-tion of procedures and retraining were to be completed 3 months after emergency procedure guidelines were established;

however, some difficulty in completing these requirements has been experienced.

Clarification of the scope of the task and appropriate schedule revisions were included in NUREG-0737, item I.C. 1.

Pending staff approval of the revised analysis and guidelines, the stqff will continue the pilot monitoring of emergency procedures described in Task Action Plan item I.C.8 (NUREG-0660).

For PWRs, this will involve review of the loss-of-coolant, steam-generator-tube

rupture, loss of main feedwater, and inadequate core cooling procedures.

The adequacy of each PWR vendor's guide-lines will be identified for each near term operating licensee (NTOL) during the emergency procedure review.

Discussion and Conclusions The staff has not approved the generic Westinghouse revised analysis and guidelines required by Task Action Plan item I.C. 1(3) and clarified in NUREG-0737.

Therefore, the pilot monitoring discussed in item I. C.8, Pilot Monitoring of Selected Emergency Procedures, was conducted using interim guidelines that have been approved by the staff.

The adequacy of the vendor provided guidelines was discussed in a meeting with PGLE personnel on November 3

and 4, 1980.

The draft procedures we reviewed for I.C.8 reflected the Westinghouse analysis of small-break LOCAs and inadequate core cooling in accordance with a license requirement and Task Action Plan (NUREG-0660) item I.C. l.

However, on November 10, 1980, the Westinghouse Owners Group provided additional guidelines for the mitigation of inadequate core cooling.

The staff has not completed its review of these guidelines.

The changes made to the draft procedures and any additional changes that may result from our review of the November 10, 1980 guidelines must be made and the Diablo Canyon operators must be trained on the changes and their bases prior to'peratio'n above 5X of the rated power level.

The Office of ICE will verify that these requirements are satisfied.

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Based on our review of the emergency procedures and our observation of the procedures being implemented on the simulator and in the, plant walk-through, we have concluded that when the required changes have been made to the procedures as specified in Section I. C.8 of this report, the Diablo Canyon emergency operating procedures will be acceptable for operation at power levels up to 100 percent of rated power.

I. C. 7 NSSS Vendor Review of Procedures Position Obtain NSSS vendor review of power ascension and emergency operating procedures to further verify their adequacy.

This requirement must be met before issuance of a full power license.

Discussion and Conclusion I

The NSSS,vendor, Westinghouse Electric Corporation, has reviewed the emergency procedures.

The changes recommended by Westinghouse have been incorporated

'nto the procedures.

This has been documented in a letter from PG8E to the NRC dated March 13, 1981.

This satisfies the requirements of item I.C.7 of NUREG-0694 in regard to emergency operating procedures.

However, we have not received verification of the Westihghouse review of the power ascension procedures.

We require that the applicant verify that 'the vendor has reviewed the power ascension procedures prior to the issuance of a full power license.

We conclude that the applicant meets the requirements of NUREG-0737 pending receipt of the verification cited above.

I.C.8 Pilot Monitorin of Selected Emer enc Procedures for NTOL A licants'osition Correct emergency procedures as necessary based on the NRC audit of selected plant emergency operating procedures (e.g.,

small-break LOCA, loss of feedwater, restart of engineeered safety features following a loss of ac power, steam-line

break, and steam generator tube rupture).

This action wi 11 be completed prior to issuance of a full power license.

Discussion and Conclusion During our review of emergency procedures, we met with PGLE personnel on November 3 and 4, 1980, to discuss the Diablo Canyon plant characteristics and the control room emergency procedures for loss-of-coolant accidents, steam generator tube rupture, loss of main feedwater, recovery from inadequate core cooling, reactor trip with safety injection and anticipated transient without trip.

These discussions resulted in several minor revisions to the 'rocedures.

On November 13, 1980, these revised procedures were employed to respond to simulations of accident and transient conditions.

A team of NRC and contractor personnel observed Diablo Canyon operators participating in the simulations of several transients and accidents on the Zion Simulator.

The transients and 2-2

accidents included loss-of-coolant (LOCA) in a range of break sizes, steam generator tube rupture, loss of main feedwater, and recovery from inadequate

'core cooling.

Some transients and accidents were run more than once and equipment failures such as loss of offsite power and failure of one emergency diesel generator, failure of scram breakers to open (ATMS), and failure of individual components in the emergency core cooling systems and auxiliary feedwater systems were included in the simulated events.

During the simulation of the events and following each event, we discussed the operators'ctions and the procedures with the operators.

As a result of this exercise, some additional changes were made to the draft emergency operating procedures.

On November 18, 1980, we met again with PG8E representatives to discuss additional changes to the procedures.

As a result of performing simulator exercises of the LOCA procedures, we found that there could be a narrow range of small break sizes for which the procedures would require repeated termination and reinitiation of safety injection.

This procedure could cause the operator to postpone the cooldown and depressurization of the reactor coolant system.

PGRE personnel then modified the procedures to provide adequate charging flow to maintain high pressure in the reactor coolant system rather than cycling the pressure with repeated termination and reinitiation of safety injection.

The staff agrees with this change although it is not specifically provided for in the generic 'guidelines for the preparation of procedures.

The Diablo Canyon procedures'lso depart from the guidelines by not including the diagnostic chart provided in the guidelines.

Me have concluded that the procedures are adequate without the diagnostic chart.

Also on November 18, 1980, a team of NRC and Battelle Pacific Northwest Laboratories personnel observed a control ro'om team of one shift foreman, shift technical advisor, shift clerk, two reactor operators and an auxiliary operator participate in a 'trial of the LOCA procedures in the Diablo Canyon Control Room.

Simulated events included a small break LOCA, inadequate core cooling and a large break LOCA.

The procedures were discussed with operations personnel during and after each simulated event.

The efficient manner in which the procedures were executed indicated that the emergency procedures were generally clear, properly sequenced and compatible with the'ontrol room equipment and arrangement.

The remainder of the emergency operating procedures must be revised in accordance with our comments on the procedures reviewed and the operators trained on the revisions"within 30 effective full power days of operation.

The Office of ICE will verify that these requirements are satisfied.

Based on our review of emergency procedures and our observation of the procedures being implemented on the simulator and in the plant walk-thru, we have concluded that when the required changes have been made to the procedures"we

reviewed, the" Diablo Canyon emergency operating procedures, as revised by the staff's recommendations, will be acceptable for operation at power levels up to 100 percent of rated power.

Future actions required by Task Action Plan

'tems I.C.l.a(3), Transients and Accidents and I.C.9, Long-Term Program Plan for Upgrading of Procedures may require future revisions to the emergency procedures.

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I.G. 1 Trainin Durin 'ow-Power Testin Position Supplement operator training by completing the special low-power test program.

Tests may be observed by other shifts or repeated on other shifts to provide training to the operators.

This requirement shall be met before issuance of a full-power license.

Discussion and Conclusion Supplement 10 to the Safety Evaluation Report concluded that the special low-power test program proposed by the licensee to supplement operator training would satisfy this requirement.

The supplemental operator training will be obtained during the performance of the seven low-power tests on Diablo Canyon:

(1) natural circulation; (2) natural circulation with simulated loss of offsite power; (3) natural circulation with loss of pressurizer heaters; (4) effect of steam generator secondary side isolation on natural circulation; (5) natural circulation at reduced pressure; (6) cooldown capability of the charging and letdown system; and (7) simulated loss of all onsite and offsite ac power.

PG8E had previously committed to perform tests to show that Diablo Canyon Unit 1 conforms to Branch Technical Position RSB 5-1.

The tests listed below will be performed using decay heat after the 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> acceptance run at full power.

These tests are not part of the low power test program but are expected to provide significant technical information and training.

8.

Boron mixing during natural circulation.

9.

Cooldown and depressurization in natural circulation to the RHR setpoint.

10.

Measurement of reactor vessel head temperature during natural circulation cooldown.

Supplement 12, issued March 4, 1981, provides our evaluation and approval of the licensee's plans to perform the seven special low-power tests on Diablo Canyon during initial plant startup.

The staff reviewed the licensee's safety

, analysis and procedures for the tests, which include the operational safety criteria, effects of the exceptions to the Diablo Canyon Technical Specifica-

tions, and offsite dose analyses for postulated accidents.

On the basis of our review, we concluded that the licensee's plans for performance of the seven special low-power tests at Diablo Canyon are acceptable.

Should the special low-power test program'ot be completed prior. to the issuance of the'full-power license, we will condition the license to require (1) the results of the seven low-power reactor tests be submitted to the Office of Nuclear Reactor Regulation (NRR) prior to exceeding five percent of rated

power, and (2) the results of all tests be included in the Startup Test Report.

Based on our above-referenced approval to conduct the special low-power test

program, we conclude that the licensee will meet our requirement for supplemental operator training during low-power testing by performance of its proposed 2-4

program.

To assure satisfactory completion of the program, the full-power license will be conditioned to require that results of the special training program be submitted as noted in the preceding paragraph.

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II.

SITING AND DESIGN II.B.1 Reactor Coolant S stem Vents Position Design and install reactor coolant system (RCS) and reactor vessel head high point vents remotely operated from the control room.

Although the purpose of the system is to vent noncondensible gases from the RCS which may inhibit core cooling during natural circulation, the vents must not lead to an unacceptable increase in the probability of a loss-of-coolant accident or a challenge to containment integrity.

Since these vents form a'part of the reactor coolant pressure

boundary, the design of the vents shall conform to the requirements of Appendix A to 10 CFR Part 50, "General Design Criteria."

The vent system shall be designed with sufficient redundancy that assures a low probability of inadvertent or irreversible actuation.

In accordance with NUREG-0737, design and operating procedures shall be submitted July j., 1981, and installation shall be completed before July 1, 1982.

Discussion and Conclusions By letters dated February 11, 1980 and,February 26, 1980; the applicant provided a conceptual design for the.TMI Task Action Plan requirement II.'B.1 to install reactor coolant system vents.

The licensee has designed the vent system to be remotely controlled and monitored and has committed that the design will be safety grade, seismically qualified, and single-failure proof.

Finally, the applicant has confirmed that a break in the vent line will be within the envelope of present accident analyses and that these an'alyses are applicable to the vent line break.

'ur preliminary review of this information has concluded that this conceptual design adequately addresses the requirements of NUREG-0737 on vents.

However, a detailed evaluation of the design has not been completed.

Some areas that will require further detail are vent system qualification to operate under accident conditions, system testability, piping design, procedural guideliries and analyses.

The applicant has committed to provide this information prior to July 1, 1981; On the foregoing bases, we conclude that the applicant has provided an acceptable description of. the vent system conceptual design for full power but that further detailed review will be necessary as outlined above.

As stated in NUREG-0737, installation of the RCS vents may precede full approval provided that measures to preclude inadvertent operation are taken.

II.B.4 Trainin for Miti atin Core Dama e

Position NUREG-0737 requires that the applicant develop a program to ensure that all operating personnel are trained in the use of installed plant systems to control or mitigate an accident in which the core is severely damaged.

The training program shall include the following topics.

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A.

Incore Instrumentation 2.

Use of fixed or movable incore detectors to determin'e extent of core damage and geometry changes.

Use of thermocouples in determining peak temperatures; methods for extended range readings; methods for direct readings at terminal junctions.

B.

Excor e Nuclear Instrumentation NIS C.

Vital Use of NIS for determination of void formation; void location basis for NIS response as a function of core temperatures

'and density changes.

Instrumentation 2.

Instrumentation response in an accident environment; failure sequence (time to failure, method of failure); indication reliability (actual vs indicated level).

Alternative methods for measuring flows, pressures,

levels, and temperatures.

a.

Determination of pressurizer level if all level transmitters fail.

b.

Determination of letdown flow with a clogged filter. (low flow).

c.

Determination of other Reactor Coolant System parameters jf the primary method of measurement has failed.

D.

Primar Chemistr

,2.

3.

Expected chemistry results with severe core damage; consequences of transferring small quantities of liquid outside containment; importance of using leak tight systems.

Expected isotopic breakdown for core damage; for clad damage.

Corrosion effects of extended immersion in primary water; time to failure.

E.

Radi ati on Monitorin 2.

Response

of Process and Area Monitors to severe damage; behavior of detectors when saturated; method for detecting radiation readings by direct measurement at detector output (over-ranged detector);

expected accuracy of detectors at different locations; use of detectors to determine extent of core damage.

Methods of determining dose rate inside containment from measurements taken outside containment.

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F.

Gas Generation 1.

. methods of H~ generation during an accident; other sources of gas

'Ne, Kr); techniques for venting or disposal of noncondensibles.

2.

Hq flammability and explosive limit; sources of 0~ in containment or Reactor Coolant System.

Discussion and Conclusions The applicant's training program to teach personnel in the use of equipment and systems. to control or mitigate core damage. is currently in progress.

Preparation of the training material is progressing simultaneously with the training program.

The applicant has developed the training program using the INPO document STM-01-06-88 "Training Guidelines for Recognizing and Mitigating the Consequences of Severe Core Damage,"

and input obtained from the Sequoyah training program and the Rogovin report.

The training program will contain the following items as a minimum, as listed in Enclosure 3 of H.

R. Denton's March 28, 1980 letter:

(1) Incore Instrumentation, (2) Excore Instrumentation, (3) Vital Instrumentation, (4) Primary Chemistry, (5) Radiation Monitoring, and (6) Gas Generation.

The applicant's training material covering items 1, 2, 4, and thermocouples has been written, and training has begun in those areas.

Materials and training for all operators in all the six areas listed above will be completed by the applicant prior to full-power operation.

Pacific Gas and Electric Company is also initiating placement of a contract with Westinghouse Electric Corporation for their training program on recognizing and mitigating the consequences of severe core damage.

This course is expected to be available by July 1981.

Materials. from this course will provide a basis for various portions of the training program for mitigation of core damage.

In addition, Chapter 13.2 of the Diablo Canyon's FSAR will be revised to provide outline of the training program to mitigate core damage.

Based on the foregoing, we have concluded that the applicant's training and requalification program meets the requirements of NUREG-0737 for training personnel in the use of installed plant systems to control or mitigate an accident in which the core is severely damaged.

The applicant has committed to complete the training of all operating personnel in the use of installed systems to monitor and control accidents in which the core may be severely damaged before the issuance of full power license.

The Office of Inspection and Enforcement will verify completion of training prior to operation above five percent power.

II.B.7 Anal sis of H dro en Control Position Reach a,decision on the immediate requirements, if any, for hydrogen control in small containments, and apply, as appropriate, to new OLs pending completion of the degraded core rulemaking in II.B.8 of the Action Plan.

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Discussion and Conclusions The staff position on plants such as Diablo Canyon, which have dry containments, is that inerting is not required as an interim action and that continued operation and licensing of dry containment plants is justified using the current design basis, pending the rulemaking proceeding.

II.B.8 Rulemakin Proceedin on. De raded-Core Accidents Position Issue an advance notice of rulemaking or requirements for design and other features "for accidents involving severely damaged cores.

These actions shall be completed before issuance of a full-power license.

Discussion and Conclusions The accident at Three Mile Island, Unit 2 resulted in a severely damaged core

'ccompanied by the generation and release to containment of hydrogen in excess of those limits allowed in current regulations.

This accident highlighted the difficulties associated with mitigating the consequences of an accident more severe than the current design basis accidents.

As a consequence, the TMI Action Plan (NUREG-0660), item II.B.B, calls for a rulemaking proceeding on consideration of degraded or melted cores in safety 'reviews to solicit comments.

The first steps in the resolution of item II.B.8 will be the issuance of an advance notice of proposed rulemaking and the issuance of an Interim Rule.

The advance notice was transmitted to the Commission in SECY 80-357, Degraded Cooling Rulemaking.

The Commission approved the advance notice on September 4, 1980.

The proposed Interim Rule was transmitted to the Commission on August 25, 1980 in SECY 80-399, "Proposed Interim Amendment to 10 CFR 50 Relating to Hydrogen Control and Certain Degraded Core Considerations,"

and was subsequently approved on September 4, 1980.

The Interim Rule, in summary, addresses the following areas:

1.

Requires inerting of all BWR Mark I and Mark II containments.

2.

3.

Requires all other plants to evaluate the effects of large amounts of hydrogen generation and to propose and assess mitigation techniques for control of hydrogen.

Codifies various Lessons Learned items to reduce the likelihood of degraded core accidents.

In addition to the effects related to the rulemaking, the staff has requested that a research program be initiated to investigate the effects of degraded/

melted core accidents for generic LWR plant designs, and to investigate various safety systems to reduce the effects of such accidents.

Additionally, the staff has contracted with the Lawrence Livermore National Laboratory for assistance on evaluating the effectiveness of distributed ignition sources within containment on an expedited basis.

The staff will, however, evaluate a

spectrum of mitigation techniques to control hydrogen and reduce the impact of severely degraded core accidents as part of the safety research program discussed above.

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Me estimate the end date of the rulemaking proceeding to be about 1983.

However, the projected end date for all the interim NRC actions identified above is January 15, 1982.

II.E.l. 1 Auxiliar Feedwater Reliabilit S stem Evaluation Position Perform a simplified AFW system reliability analysis that uses event-tree and faul,t-tree logic techniques to determine the potential for AFW system failure under various loss-of-main-feedwater-transient conditions.

Particular emphasis is given to. determining potential failures that could result from human errors, common causes, single-point vulnerabi lities, and test and maintenance outages; (2)

Perform a deterministic review of the AFM system using the acceptance criteria of Standard Review Plan Section 10.4.9 and associated Branch Technical Position ASB 10-1 as principal guidance; and (3)

Reevaluate the AFW system flowrate design bases and criteria.

Clarification The Three Mile Island Unit 2 (TMI-2) accident and subsequent investigations and studies highlighted the importance of the Auxiliary Feedwater System (AFWS) in the mitigation of transients and accidents.

As part of our assess-ment of the TMI-2 accident and related implications for operating plants, we evaluated the AFM systems for all operating plants having nuclear steam supply systems (NSSS) designed by Westinghouse (NUREG-0611) or Combustion Engineering (NUREG-0635).

Our evaluations of 'these system designs are contained in the NUREGs along with our recommendations for each plant and the concerns which led to each recommendation.

The objectives of the evaluation were to:

(1) identify necessary changes in AFW system design or related procedures at the operating facilities in order to assure the continued safe operation of these

plants, and (2) to identify other system characteristics of the AFW systems

which, on a long term basis, may require system modifications.

To accomplish these objectives, we:

(1)

Reviewed plant specific AFW system designs in light of current regulatory requirements (SRP) and,.

(2)

Assessed the relative reliability of the various AFW systems under various loss of feedwater transients (one of which was the initiating event of TMI-2) and other postulated failure conditions by determining the potential for AFW system failure due to common causes, single point vulnerabilities, and human error.

Discussion and Conclusions Licensee Res onse to osstions (1) and 2

and 3

Cited Above In accordance with the requirements of Item II.E. l. 1 of NUREG-0660, "NRC Action Plan Developed as a Result of the TNI-2 Accident," we have included the following results of the Diablo Canyon auxiliary feedwater

.system (AFMS) review in this supplement:

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2.

3.

4, Me have applied the generic results and recommendations from the above described reviews for operating plants to the Diablo Canyon AFMS.

We have reviewed the detailed Diablo Canyon AFWS reliability analysis submitted by the applicant.

The following is our evaluation of this reliability analysis.

In a letter dated October 9, 1980, the applicant provided a document.

entitled "Reliability Analysis of Diablo Canyon Auxiliary Feedwater

System, PLG-0140, Revision 3."

This report evaluated the AFWS reliability for the three postulated transient and accident scenarios identified for study in our March 10, 1980 letter utilizing fault tree methodology.

Overall numerical system unavailability for the three cases was determined using the NRC approved failure rate data base and again using plant specific data.

- Results of the applicant s analysis indicated that the Diablo Canyon AFWS ranked in the high reliability range for case 1,

Loss of Main Feedwater, and case 2,

Loss of Offsite Power, and in the medium reliability range for case 3,

Loss of All AC Power.

Dominant contributors to AFWS unreliability were also identified.

However, we disagree with the applicant's assigned failure rates and resulting system unavailability as a result of failure in the condensate storage tank common outlet valve (a dominant failure contributor).

We have determined a slightly lower system reliability (medium to high range) for cases 1 and 2.

Based on this, the applicant committed to modify the AFWS in accordance with the discussion under item GL-2 of this SER Supplement in order to improve system reliability and the effects of this dominant failure source.

Me conclude that the applicant has satisfactorily complied with the reliability study requirements of our March 10, 1980 letter, and the AFWS reliability assessment is acceptable.

We have reviewed the applicant's deterministic comparison of the Diablo Canyon AFWS against Standard Review Plan (SRP) Section 10.4.9 and Branch Technical Position (BTP) ASB 10-1, and find that the AFWS design is in compliance.

Our evaluation of the environmental qualification of the AFMS will be reported in a future supplement which is expected to be issued in mid-May, 1981.

We have reviewed the applicant's response to our request in Enclosure 2

of our letter dated March 10, 1980, regarding the design basi.s for AFWS flow requirements.

The applicant provided this information in a letter dated October 9, 1980.

Me conclude that the applicant's design basis for AFWS flow requirements is acceptable.

Me conclude that the implementation of the following recommendations identified from the above reviews have improved the reliability of the Diablo Canyon AFW system.

Im lementation of NRC Recommendations Short-erm Recommendatsons l.

Recommendatson GS "The licensee should propose modifications to the echnical Specifications. to limit the time that one AFM system pump and its associated flow train and essential instrumentation can 2-11

be inoperable.

The outage time limit and subsequent action time should be as required in current Technical Specifications; i. e.,

72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, respectively."

In response, the applicant indicated in a letter dated April 8, 1980, that the proposed draft Diablo Canyon Technical Specification, Section 3.7. 1.2, applies.

This specification limits the plant operation with one AFW train out of service to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and the subsequent action time to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Me conclude that this Technical Specification is in compliance with our recommendation and is, therefore,'cceptable.

2.

Recommendation GS "The licensee should lock open single valves or multiple valves in series in the AFW system pump suction piping and lock open other single valves or multiple valves in series tha't could interrupt all AFW flow.

Monthly inspections should be performed to verify that these valves are locked and in the open position.

These inspections should be proposed for incorporation into the surveillance requirements of the plant Technical Specifications.

The long term resolution of this concern is discussed in.

Recommendation GL-2 '

In response to this recommendation,'he

'applicant indicated in a letter dated April 8, 1980, that'there is a single normally open manual valve in the common suction piping to the AFW system pumps.

The applicant, by letter dated February 26, 1981, has proposed to modify this value as discussed under Item GL-2 of.this section.

Satisfactory resolution of Item GL-2 as indicated resolves this concern and the applicant's response is therefore, acceptable.

3.

Recommendation GS "The licensee has stated that it throttles AFM system f ow to avowed water hammer.

The'icensee should reexamine the practice of throttling AFW system flow to avoid water hammer.

The licensee should verify that the AFM system will supply on demand sufficient initial flow to the necessary steam generators to assure adequate decay heat removal following loss of main feedwater flow and a reactor trip from 100K power.

In cases where this reevaluation results in an increase in initial AFW system flow, the licensee should provide sufficient information to demonstrate 'that the required initial AFW system flow will not result in plant damage due to water hammer."

In response to this recommendation, the applicant indicated in a letter dated April 8, 1980, that on automatic start the AFWS control valves will be full open and, therefore, will deliver full flow until normal water level is established in the steam generators.

AFW flow will not be prevented or reduced because of water hammer considerations.

The steam generator feedwater spargers at Diablo Canyon have been modified by the installation ot "J-tubes" in order to minimize the possibility of a water hammer.

We, therefore, conclude that this recommendation does not apply to Diablo Canyon.

f If 4.

Recommendation GS "Emergency procedures for transferring.to alternate sources of AFW supply should be available to the plant 2"12

operators.

These procedures, should include criteria to inform the operator

when, and in what order, the transfer to alternate water sources should take place.

The following case should be covered by the procedures:

(1)

The case in which the primary water supply is not initially available.

The procedures for this case should include any operator actions required to protect the AF'W system pumps against self-damage before water flow is initiated.

(2)

The case in which the primary water supply is being depleted.

The procedure for this case should provide for transfer to the alternate water sources prior to draining.of the primary water supply."

In response to this recommendation, the applicant indicated in a letter dated April 8, 1980,,that plant emergency procedures are being revised to incorporate the above cases for transferring AFW supply to alternate sources.

We conclude that the applicant's response is acceptable

and, therefore, is in compliance with our recommendation pending verifi-cation of the emergency procedures by the Office of Inspection and Enforcement.

5.

Recommendation GS "The as-built plant should be capable of provldsng t e required AFW flow for at least two hours from one AFW pump train, independent of any alternating current power source.

If manual AFW system initiation or flow control is required following a complete loss of alternating current power,,emergency procedures should be established for manually initiating and controlling the system under these conditions.

Since the water for cooling of the lube oil for the turbine-driven pump bearings may be dependent on alternating current power, design or procedural changes shall be made to eliminate this dependency as soon as practicable.

Until this is done, the emergency procedures should provide for an individual to be stationed at the turbine-driven pump in the event of the loss of all alternating current power to monitor pump bearing and/or lube oil temperatures.

If necessary, this operator would operate the turbine-driven pump in a manual on-off mode until alter-nating current power is restored.

Adequate lighting powered by direct current power sources and communications at local stations should also be provided if manual initiation and control of the AFW system is needed.

See Recommendation GL-3 for the longer-term resolution of this concern."

In response to this recommendation, the applicant indicated in a letter dated April 8, 1980, that the turbine-driven AFW pump has no dependence on AC power.

Turbine bearing lube oil cooling is taken from the pump discharge.

The applicant further indicated that appropriate AFWS operating procedures wi 11 be prepared to cover the event of loss of all AC 2-13

power sources.

We conclude that the provisions available in the

. existing AFM system and appropriate operating procedures meet the requirements outlined in this recommendation and are, therefore, acceptable, pending verification of the procedures by the Office of Inspection and Enforcement,.

6.

Recommendation GS "The, licensee should confirm flow path avail-abs sty of an system flow train, that has been out of service to perform periodic testing or maintenance as follows:

(a)

Procedures should be implemented to require an operator to determine that the AFM system valves are pro'perly aligned and a

second operator to independently verify that the valves are properly aligned.

(b)

The licensee should propose Technical Specifications to assure that prior to plant startup following an extended cold shutdown, a flow test would be performed to verify the normal flow path from the primary AFW system water source to the steam generators.

The flow test should be'conducted with AFW system valves in their normal alignment."

In a letter dated February 26, 1981, in response to part (a) of this recommendation, the applicant has committed to revise plant procedures to include a second operator sign-off for independent verification of proper valve position fo'llowing restoration" of the AFMS by the first operator subsequent to testing or maintenance.

Based on the above commitment, we conclude.that the applicant's response is acceptable.

Our Office of Inspection and Enforcement will review the revised procedures prior to the issuance of a, full power license.

In response to part (b) of this recommendation, the applicant stated in a letter dated April 8, 1980 that the AFWS is required for plant startup following any cold shutdown.

Thus, the availability of an AFWS flow path from the primary water source to the steam generators is automatically.

verified for that flow path during the normal course of plant startup.

Me'conclude that the applicant's response to 'this recommendation is acceptable and, therefore, the applicant is in compliance with our recom-mendation.

7.

Recommendation GS "The licensee should verify that the automatic start M system signals and associated circuitry are safety-grade.

If this cannot'be verified, the AFW system automatic initiation

,system should be modified in the short-term to meet the functional requirements listed below.

For the longer term, the automatic initiation signals and circuits should be upgraded to meet safety-grade requirements as indicated in Recommendation GL-5.

.(1)

The design should provide for the automatic initiation of the auxiliary feedwater system flow.

(2)

The automatic initiation signals and'ircuits should be designed so that a single failure will not result in the loss of auxiliary feedwater system function.

2-14

(3)

Testability of the initiation signals and circuits shall be a feature of the design.

(4)

The initiation signals and circuits should be powered from the emergency buses.

(5)

Manual capability to initiation of the auxiliary feedwater system from the control room should be retained and should be implemented so that a single failure in the manual circuits will not result in the loss of system function.

(6)

The alternating current motor-driven pumps and v'alves in the auxiliary feedwater system should be included in" the automatic actuation (simultaneous and/or sequential) of the 1'oads to the emergency buses.

(7)

The automatic initiation signals and circuits shall be designed so that their failure will not result in the loss of manual capability to initiate the AFM system from the control room."

In response to this recommendation, the applicant stated in a letter dated April 8, 1980, that the Diablo Canyon AFM system is designed for automatic start and meets all the above indicated functional requirements.

The applicant further states that the AFW actuation signals and associated circuitry are safety.grade.

Me have reviewed the applicant s design and find that't satisfies the requirements of NUREG-0737 (refer to Item II.E; l. 2 of this supplement)..

8.

Recommendation GS This recommendation is not applicable to Diablo

~

anyon as the AFMS design currently includes automatic initiation of AFM system flow.

The recfuirements'for the automatic initiation design are covered by Recommendation GS-7.,

B.

A'dditional Short Term Recommendations Recommendation

- "The licensee should provide redundant level sndscatson and,low level alarms in the control room for the AFW system primary water 'supply, to allow the operator to anticipate the need to make up water or transfer to an alternate water supply and prevent a low pump suction pressure condition from occurring.

The low level alarm setpoint should allow at least 20 minutes for operator

action, assuming that the largest capacity AFM pump is operating."

II In response to this recommendation, the'applicant indicated in a letter dated April 8, 1980, that condensate storage tank (the primary AFW source) level indication is provided locally at the tank, at the remote shutdown panel, and in the control room.

The level indication instrument channels are seismically qualified.

The applicant has committed to upgrade the instrumentation to be redundant and safety grade.

The existing condensate storage tank low-low level alarm i's annunciated in the control room.

This alarm instrument circuit is safety grade.

The applicant has committed.to reset the alarm setpoint from its current level which provides 16 minutes for operator action to transfer AFW supply to the secondary source to a level which allows 20 minutes for this operator action.

2-15

2.

Me have reviewed the applicant's response and. conclude that it meets the requirements of this recommendation and is, therefore, acceptable.

Recommendation This recommendation has been revised from the 'ori inal recommendatson sn NUREG-0611 he licensee shou d perform a 4 -hour endurance test on a 1

AFM system pumps,'f such a test or continuous period of operation has not been accomplished to date.

Following the 48-hour pump run, the pumps should be shut down and cooled down and then restarted and run fo'0 one hour.

Test acceptance criteria should include demonstrating that the pumps remain within design limits with respect to bearing/bearing oil temperatures and vibration and that pump room ambient conditions (temperature, humidity) do not exceed environmental qualification limits for safety-.related equipment in the room."

In response to this recommendation',

the applicant i'ndicated in a letter dated April 8, 1980, that an endurance test will be performed prior to plant startup.

Since that time, the applicant has informed the NRC staff that the endurance test of the motor operated auxiliary feedwater pumps will be performed prior to plant startup.

However, due to lack of steam availability of steam prior to plant startup, the t'est of the steam driven auxiliary feedwater pump will be performed during the low power test program, we consider this to be acceptable.

The applicant further indicated that the endurance test procedures and acceptance criteria will be made available to the NRC for comment prior to performing the t'est.

As noted above, this recommendation has been revised from the original recommendation in the March 10, 1980 letter to require a 48-hour endurance test on all AFM system pumps.

In addition, the applicant should provide a copy of test results including:

(1) a brief description of the test method and instrumentation

used, (2) a plot of bearing temperature vs time for each pump demonstrating that the temperature design limits were not

exceeded, (3) a plot of pump room ambient temperature and humidity vs.

time to demonstrate that the pump room ambient conditions do not

'exceed environmental qualification limits for safety-related equipment in the room,'nd (4) a statement confirming that the pump vibration limits were not exceeded.

Based on the applicant's commitment, we conclude that the response to this recommendation is acceptable.

If the test results are not acceptable, we will require modifications and provide a safety evaluation regarding the tests and modifications.

Recommendation

- "The licensee should implement the following requirements as spec>fred by Item 2. 1. 7. b on page A-32 of NUREG-0578:

"Safety-grade indication of auxiliary feedwater flow to each steam generator shall be provided in the control room.

The auxiliary feedwater flow instrument channels shall be powered from the emergency buses consistent with satisfying the emergency power diversity requirements for the auxiliary feedwater system set forth in Auxiliary Systems Branch Technical Position 10-1 of the'tandard Review Plan, Section 10.4.9."

2-16

4.

In response to this recommendation, the applicant indicated in a letter dated April.8, 1980,,that the AFWS design includes indication of AFW flow to each steam generator in the control room and at the remote shutdown panel.

The instrument channels are safety grade and powered from diverse emergency vit'al buses.

As stated under Item II.E. 1. 2=-in Supplement 12,

. the applicant has stated that replacement transmitters meeting the safety-grade requirements are on order and should be installed in mid-1981.

Me will require this action. to be completed by July 1, 1981.

Based on the

.applicant s commitment and the above required action data, we find this to be acceptable.

Recommendation

-. "Licensees with plants which require local manual realignment of valves to conduct periodic tests on one AFW system train, and there is only one remaining AFW train available for operation, should propose Technical Specifications to provide that a dedicated individual who is in communication with the control room be stationed at the manual valves.

Upon instruction from the control room,,this operator would realign the valves in the AFW system train from the test mode to their operational alignment."

4 In response to this recommendation, the applicant indicated in a letter dated April 8, 1980, that there are three AFM trains.

During the periodic tests of the AFM system, only one pump at a time is run, and, therefore, there are still two trains available.

The applicant further stated that periodic testing of the AFWS does not require local manual valve realignment.

Realignment of system valves from the normal operational mode to the test mode and back is accomplished from the control room.

Based on the applicant's response,~

we concludethat this recommendation is not applicable to Diablo Canyon.

C.

Lon Term Recommendations Recommendation GL "For plants with a manual starting AFW system, the licensee shoul install a system to automatically initiate the AFW system flow.

This system and associated automatic initiation signals should be undesigned and installed to meet safety-grade require-ments.'anual AFM, system start and control, capability should be retained with manual start, serving as backup to automatic AFW system intiation."

2.

Because the applicant's response to Recommendation GS-7 stated that the Diablo Canyon AFW design includes automatic star t, Recommendation GL-1 is not applicable to Diablo Canyon.

Recommendation GL "Licensees with plant designs in which all primary and a ternate) water supplies to the AFM systems pass through valves in a single flow path should install redundant parallel flow paths (piping and valves).

Licensees with plants in which the primary AFW system water supply passes through valves in a single flow path, but the alternate AFW system water supplies connect to the AFW system pump suction piping to the above valve(s) or provide automatic opening of the valve(s) from the alternate water supply upon low pump suction pressure.

2-17

The licensee should propose Technical Specifications to incorporate appropriate periodic inspections to verify the valve positions."

In response to this recommendation, in a letter dated February 26, 1981, the applicant proposed the following modification:'he common suction supply valve is a manual gear operated butterfly valve.

The applicant proposes to render this valve inoperable by replacing the existing pointer on the valve shaft with one made of steel, plate.

The plate and shaft will be drilled and pinned with a steel pin which will be tack welded in place.

After assuring that the valve is fully open, the plate and oper'ator cover will be drilled and secured together with a bolt which will also be tack welded.

In addition, the valve will be included in the plant sealed valve list.

The applicant has further indicated that the valve disc is secured to the shaft by vibration proof taper pins equal in strength to the shaft itself thus reducing the possibility of the disc freeing itself from the shaft and rotating in a manner which might block pump suction flow.

Based on the above, we conclude that the applicant's response is acceptable.

3.

Recommendation GL "At least one AFW system pump and its associated f ow.path an essential instrumentation should automatically initiate AFW system flow and be capable of being operated independently of any AC power source for at least two hours.

Conversion of DC power to AC power is acceptable."

In response to this recommendation, the applicant committed in a letter dated April 8, 1980, to modify one AFWS train to meet the above requirement at or prior to the first refueling.. This train will consist of the existing turbine driven AFM pump which is capab'le of delivering AFW flow to all four steam generators, a steam supply valve powered from a vital OC bus, automatic AFMS actuation instrumentation powered from a vital (battery backed) instrument bus, and steam generator level and AFW fl.ow indication instrumentation powered from a vital (battery backed) bus.

Me

, have reviewed this response and conclude that it meets the requirements

'f this recommendation and is, therefore, acceptable.

4.

Recommendation GL "Licensees having plants with unprotected normal AFM system water supplies should evaluate the design of their AFM systems to determine if automatic protection of the pumps is necessary following a seismic event or a tornado.

The time available before pump damage, the alarms and indications available to the control room operator, and the time necessary for assessing the problem and taking action should be considered in determining whether operator action can be rel,ied on to prevent pump damage.

Consideration should be given to providing pump protection by means such as automatic switchover of the pump suctions to the alternate safety-grade source of water, automatic pump trips on low suction pressure, or upgrading the normal source of water to meet seismic Category I and tornado protection requirements."

2-18

In response to this recommendation, the applicant indicated in a letter dated April 8, 1980, that the primary water source for AFM system supply is seismic Category I and protected against potential damage due to tornados.

Based on the applicant's

response, we conclude that this recommendation is not applicable to Diablo Canyon.

5.

Recommendation GL "The licensee should upgrade the AFM system auto-matic snltiat)on signals and circuits to meet safety-grade requirements."

In response to this recommendation, the applicant indicated in a letter dated April 8, 1980, that the present AFM system automatic initiation signals are safety grade.

As stated in Item II.E. 1. 2 of Supplement 10, the AFW system initiation signals and circuits meet safety-grade requirements and is, therefore, acceptable.

II.E. 1. 2 Auxiliar Feedwater Initiation and Indication As stated in Supplement 10, the Auxiliary Feedwater flow initiation meets safety-grade requirements.

The auxiliary feedwater flow rate indication was also addressed in Supplement 10 of the Diablo Canyon Safety Evaluation Report.

As noted therein, we concluded that the system for the indication of auxiliary feedwater flow satisfied the short-term control grade requirements, for a fuel loading license, until January 1, 1981 when the safety grade requirements must be met.

In NUREG-0737, "Clarification of TNI-Action Plan Requirements,"

the implementation date was extended until July 1, 1981, for operating plants.

The applicant has indicated that replacement transmitters meeting the safety grade requirements are on order and should be installed by mid-1981.

As stated in Supplement 12, the low power operating license will be conditioned to require this action to provide conformance to the safety grade requirements by July 1, 1981.

II.E.3. 1 Emer enc Power<<Su 1

for Pressurizer Heaters Position Ib Consistent with satisfying the requirements of General-Design Criteria 10, 14, 15, 17, and 20 of Appendix A to 10 CFR Part 50 for the event of loss of offsite power, the following positions shall be implemented:

(1)

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

a predetermined number of pressurizer heater s 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)

Procedures and training shall be established to make the operator aware of when and how the required pressurizer heaters shall be connected to the emergency buses.

If required, the procedures shall identify under what conditions selected emergency loads can be shed from the emergency power source to provide sufficient capacity for the connection of the pressurizer heaters.

2-19

(3)

The time required to accomplish the connection of the preselected pressurizer hea er to the emergency 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.

Clarification (1)

Redundant heater capacity must be provided, and each redundant heater or group of heaters should have access to only one Class IE division power supply.

(2)

The number of heaters required to hav'e access to each emergency power source is that number required to maintain natural circulation in the hot.

standby condition.

(3)

The power sources needed not necessarily have the capacity to provide power to the heaters concurrently with the loads required for loss-of-coolant accident.

(4)

Any changeover of the heaters from normal offiste power to emergency onsite power is to be accomplished manually, in the control room.

(5)

In establishing procedures to manually load the pressurizer heaters onto the emergency power sources, careful consideration must be given to:

(a) which ESF loads may be appropriately shed for a given situation;

( )

reset of the safety injection actuation signal to permit the operation of the heaters; and (c) instrumentation and criteria for operator use to prevent overloading a diesel generator.

(6)

The Class IE interfaces for main power and control power are to be protected by safety-grade circuit breakers (see also Regulatory Guide 1.75).

(7)

Being non-Class IE loads, the pressurizer heaters must be automatically shed from the emergency power sources upon the occurrence of a safety injection actuation signal (see item 5.b above).

Discussion and Conclusions All four pressurizer heater groups in the Diablo Canyon design can be supplied from the offsite power sources when they are available, and two of the four groups can be transferred to the emergency power sources when the offsite sources are not available.

Each of the two selected groups (12 & 13) has access to only one Class IE diesel generator and their controls are likewise supplied from separate safety grade dc power supplies.

Each heater group (12 & 13) is comprised of seven 69 kW heaters (483 kW total).

Therefore, energizing three of the seven heaters (207 kW) will be more than adequate to fulfillthe calculated minimum heater requirement of 150 kW.

This is in accordance with NRC position 1 and clarification items 1 and 2.

2-20

In accordance with position 4 and clarification 6, the connection of the pressurizer heater 'elements and controls to the Class IE buses is through safety grade circuit breakers, and the heaters are automatically tripped off of the emergency buses upon occurrence of a safety injection (SI) signal in accordance with clarification 7.

l PGLE has developed procedures and implemented the training of their operators to make them aware of when and how the required heaters should be connected to the emergency buses.

The procedures identify under what conditions selected loads can be shed from the emergency bus to prevent overloading the diesel generator when the pressurizer heaters are connected and also include provisions to reset the SI signal to permit the operation of the heater's.

Diesel generator and pressurizer heater loading information is displayed in the control room.

This covers position 2 and clarifications 3 and 5.

In order to align the pressurizer heaters to the emergency power source, an operator must be dispatched to the 100 ft level of the Auxiliary Building, three floors directly below the main, control

room, and manually throw a transfer switch.

This action in itself does not connect the heaters to the emergency buses but allows subsequent operation of the heaters from the emergency buses using the normal control devices provided on the main control console.

Although not in strict accordance with clarification item 4, this is considered an acceptable alternative since the heaters can be supplied with emergency power well within the one hour limit recommended by Mestinghouse and therefore meets the criteria of position 3.

Based on our review, we conclude that the design for providing emergency power to the pressurizer heaters at Diablo Canyon Nuclear Generating Station Units l and 2 is consistent with the NRC positions and clarifications in NUREG-0737 and is acceptable.

II.E.4. 2 Containment Isolation De endabilit Position (2)

Containment isolation system designs shall comply with the recommendations of Standard Review Plan Section 6.2.4 (i.e., that there be diversity in the parameters sensed for the initiation of containment isolation).

All plant personnel shall give careful consideration to the definition of essential and nonessential

systems, identify each system determined to be essential, identify each system determined to be nonessential, describe the basis for selection of each essential

system, modify their containment isolation designs accordingly, and report the results of the reevaluation to the NRC.

(3)

(4)

All nonessential systems shall be automatically isolated by the containment isolation signal.

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.

Reopening of contain-ment isolation valves shall require deliberate operator action.

2-21

(5)

(6)

(7)

The containment setpoint pressure that initiates containment isolation for nonessential penetrations must be reduced to the minimum compatible with normal operating conditions.

Containment purge valves that do not satisfy the operability criteria set forth in Branch Technical Position CSB 6-4 or the Staff Interim Position of October 23, 1979 must be sealed closed as defined in SRP 6.2.4, item II.3.f during operational conditions 1, 2, 3, and 4.

Furthermore,'hese valves must be verified to be closed at least every 31 days.

Containment purge and vent isolation valves must close on a high radiation signal.

Clarification The reference to SRP 6.2.4 in position 1 is only to the diversity requirements set forth in that document.

(2)

For postaccident situations, each nonessential penetration (except instru-ment lines) is required to have two isolation barriers in series that meet the requirements of General Design Criteria 54, 55, 56, and 57, as clarified by Standard Review Plan, Section 6.2.4.

Isolation must be performed automatically (i.e.,

no credit can be given for operator action).

Manual valves must be sealed

closed, as defined by Standard Review Plan, Section 6.2.4, to qualify as an isolation barrier.

Each automatic isola-tion valve in a nonessential penetration must receive the diverse isolation signals.

(3)

Revision 2 to Regulatory Guide l. 141 will contain guidance on the classifi-cation of essential"versus nonessential systems and is due to be issued by June 1981.

Requirements for operating. plants to review their list of essential and nonessential systems will be issued in conjunction with this guide including an appropriate time schedule for completion.

(4)

Administrative provisions to close all isolation valves manually before resetting the 'isolation signals is not an acceptable method of meeting position 4.

(5)

(6)

Ganged reopening of containment isolation valves is not acceptable.,

Reopening of isolation, valves must be performed on a valve-by-valve

basis, or on a line-by-line basis, provided that electrical independence and other single-failure criteria continue to be satisfied.

The containment pressure history during normal operation should be used as a basis for arriving at an appropriate minimum pressure setpoint for initiating containment isolation.

The pressure setpoint selected should be far enough above the maximum observed (or expected) pressure inside containment during normal operation so that inadvertent containment, isolation does not occur during normal operation from instrument drift or fluctuations due to the accuracy of the pressure sensor.

A margin of 1 psi above the maximum expected containment, pressure, should be adequate to account for instrument error.

Any proposed 'values greater. than 1 psi will require detailed justification.

Applicants for an operating license 2"22

and operating plant. licensees that have operated

.less than one year should use pressure history data from similar plants that have operated more than one year,- if possible, to arrive at a minimum containment setpoint pressure.

(7)

Sealed-closed purge.,isolation valves shall be under administrative control to assure that they cannot be inadvertently opened.

Administrative control includes mechanical devices to seal or lock the,valve closed, or to prevent power from being supplied to 'the valve operator.

Checking the valve position light'in the control room is an adequate method for verifying every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> that the purge valves are closed.

Discussion and Conclusion The staff's discussion on and evaluation of each position listed above follows:

Diverse Isolation Si nal (Position(l There are two phases of containment isolation at Diablo Canyon.

Phase A

isolates all nonessential process lines but does not affect safety injection, containment spray, containment pressure

monitors, component cooling water supplied to the reactor coolant pumps and containment fan coolers, auxiliary feedwater, main steam, steam

dump, and steam to auxiliary feedwater pump turbine.

Phase B isolates the main steam isolation valves and component cooling water-supplied to the reactor coolant pumps.

Phase A isolation is initiated'y high containment pressure, high differential pressure between st'earn lines, low pressurizer

pressure, high steam flow with low steam pressure or'low-low T average, or manual initiation.

Phase B isolation is initiated by high-high containment pressure or manually.

Me conclude that this system complies with the requirements of SRP 6.2.4 regarding diversity of parameters sensed for containment isolation.

Essential and Nonessential S stem List Position 2 )

The applicant (PG8E) has defined three levels of containment process pene-

trations, which are as follows:

1.

"Nonessential" process lines are defined as those which do not increase the potential for damage for in-containment equipment when isolated.

These are isolated on Phase A isolation.

2.

"Essential" process lines are those providing cooling water and seal water flow to the reactor coolant pumps.

These services should not be interrupted while the reactor coolant pumps are operating.

Hain steam lines are also "essential," for reasons given below.

These lines are isolated on Phase B isolation.

3.

Safety system process lines are those required to perform the function of the Engineered Safety Features System.

The staff finds acceptable the systems which the applicant has designated as nonessential (Category 1) systems and the provisions for sensing diverse 2-23

parameters to isolate them.

The safety systems in Category 3 above are essential systems which do not require automatic isolation.

Steam dump, whi'ch is a Category 3 system, is an essential system because 10K steam dump i' required for safe shutdown.

Me agree with the applicant's view that the systems listed in Category 2 above are essential systems; these

systems, while not engineered safety feature (ESF), systems required for accident mitigation, may nonetheless be considered important to postaccident plant safety and valuable in accident mitigation.

For such essential

systems, the sensing of diverse parameters for the initiation of'ontainment isolation is not.required.

Further justification for this is delineated below.

a.

Reactor Coolant Pum Coolin Mater and Seal. Water Su lies This service is maintained to prevent possible pump motor or seal

'amage following events, such as a main steam line break" outside containment or small isolable LOCA, which may not result in contain-ment pressure reaching the high-high setpoint, and for which continued reactor coolant pump operation may be desirable or even necessary.

b.

Main Steam Isolation Valve Closure Closure of the-main steam isolation valves results in loss of the normal heat sink, loss of normal pressure control, and actuation of main steam safety valves, with associated system transients.

In order to minimize unnecessary

closures, these valves are closed upon receipt of the phase B isolation signal.

We conclude that P68E has met the requirements f'r classification of systems as essential and nonessential.

Automatic Isolation of Nonessential S stem (Position(3 All nonessential systems use either manually sealed closed valves or else the valves are automatically isolated on a Phase A containment isolation signal.

Additionally, the essential systems listed in Category 2 of Position 2 above (i.e., cooling and seal water to the reactor coolant"pumps, and main steam lines) are automatically isolated on a phase 8 containment isolation signal.

Me conclude that these provisions comply with the requirements of Position (3) above.

Prevent Automatic Reo enin on Reset Position 4

Our review indicates that, with one exception, the Diablo Canyon design meets the requirements of this item.

Resetting the isolation signal will not result in the automatic reopening of any containment isolation valves.

However, contrary to Clarification (5) above, ganged reopening of several Liquid Ra'd-waste isolation valves can result from a single operator action, after the isolation'ignal has been reset.

Four isolation valves inside containment can be reopened by a single action, or five valves outside containment (associated with the four valves just mentioned inside containment) can be reopened by a different single action.

Therefore, we will require that modifications be completed, before full power operation begins,'to bring the isolation system into compliance with the requirements of Clarification (5).

Me conclude that, with the 'exception noted above, the Diablo Canyon design meets the requirements of this item.

Containment Set 'oint Pressure (Position 5

I The containment setpoint pressure that initiates containment isolation for nonessential, penetrations is set at less than or equal to 3 psig.

The appli-cant states that, after the plant becomes operational, the containment pressure history during normal operation will be, evaluated to arrive at an appropriate minimum pressure setpoint.

This approach does not meet the requirements.

'The applicant must provide, and justify, the minimum containment pressure that will be used for initiating containment isolation as stated in Position 5 above, and this pressure set-point must be implemented at Diablo Canyon by July 1, 1981, or before issuance of a full power operating license, whichever is later.

Containment Pur e Valves Position 6

As we stated in Section 6.2.3 of Supplement 9 to the staff's SER for Diablo Canyon, the Diablo Canyon purge.system valves satisfy the operability criteria set forth in Branch Technical Position CSB 6-4.

Me also stated that the applicant would block the 12-inch vacuum/overpressure relief valves to no more than 50 degrees open until the app'licant has submitted analyses demonstrating operability of these valves.

The applicant also committed to limit purge/vent operations to no more than 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> per year.

Thus, the 12-inch vacuum/overpressure relief line valves meet the operability criteria set forth in the Staff Interim Position of October 23,

1979, and we find that the Diablo Canyon plant meets the requirements of Position (6) above for full power operation.

However, the rest'rictions on purge/vent operations detailed in Section 6.2.3 of Supplement 9 to the SER must remain in effect.

Also, when the applicant.submits analyses demonstrating operability of the vacuum/overpressure relief line valves, we will report our findings.

Containment Pur e and Vent Isolation Valves Position(7 The containment purge and ventilation isolation valves are closed automati-cally by any one of the following:

1) 2)

3)

Phase A isolation signal High gaseous or air particulate radioactivity in containment High radiation at the plant vent Me find that the requirements of Position (7) have been met.

Me conclude that the applicant has met all the requirements of Positions (1),

(2), (3), (6), and (7).

The requirements of Positions (4) and (5) have been met, except as noted in the appropriate sections above.

Me require resolutions of our concerns with regar d to ganged reopening of certain isolation valves and the containment pressure setpoint prior to the issuance of a full power operating license.

On the basis of our evaluation and the above cited require-

ments, we conclude that the design is acceptable.

2-25

III.

EMERGENCY PREPARATIONS AND RADIATION PROTECTION III.A.l. 1 U

vade Emer enc Pre aredness Position Provide an emergency response plan in compliance with NUREG-0654, Rev.

1 (November 1980) "Criteria for Preparation and Evaluation of Radiological Emergency

Response

Plans and Preparedness in Support of Nuclear Power Plants."

NRC will give substantial weight to FEMA findings on offsite plans in judging the adequacy against NUREG-0654.

Perform an emergency exercise to test the integrated capability and a major portion of the basic elements existing within emergency preparedness plans and organizations.

This requirement shall be met before issuance of a full-power license.

Discussion and Conclusions Me have reviewed the applicant's revised emergency plan against the current regulatory requirements contained in 10 CFR Part 50 and the guidance criteria in NUREG-0654 dated November 1980.

A discussion of our evaluation is provided.

in Appendix B to this supplement.

Upon 'satisfactory completion of the items identified below, the staff will issue a favorable finding with respect to emergency preparedness matters for full power operation.

1.

Revision of site plan to address protective action determination and implementation during an earthquake.

2.

NRC review of FEMA findings on the applicant's recommendations during earthquakes.

3.

4.

5.

Performance of an emergency response exercise that tests the integrated capability and a major portion of. the basic elements existing within the emergency preparedness plans and organizations (NUREG-0694, item III.A.1. 1).

NRC review of the Federal Emergency Management Agency findings and deter-minations as to whether State and local emergency plans are adequate and capable of being implemented (10 CFR 50.'47a).

Final NRC approval of the state of emergency preparedness for the Diablo Canyon site will be made following implementation of the emergency plans to include development of procedures, training and qualifying of personnel, installation of equipment and facilities, and a joint. exercise of all the plans (site, State, and local).,

III.D.1. 1 Primar Coolant Sources Outside Containment Position Applicants shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as-low-as-practical levels.

This program shall include the following:

2"26

(1)

Immediate leak reduction (a)

Implement all practical leak reduction measures for all systems that carry radioactive fluid outside of containment.

(b)

Measure actual leakage rates with system in operation and report them to the NRC.

(2)

Continuing Leak Reduction Establish and implement a program of preven-tive maintenance to reduce leakage to as-low-as-practical levels.

This program shall include periodic integrated leak tests at intervals not to exceed each refueling cycle.

Clarification Applicant shall provide a summary description, together with initial leak-test

results, of their program to reduce leakage from systems outside containment that would or could contain primary coolant or other highly radioactive fluids or gases during or following a serious transient or accident.

(1)

Systems that should be leak tested are as follows {any other plant system which has similar functions or postaccident characteristics even though not specified herein, should be included):

Residual heat removal (RHR)

Containment spray recirculation High-pressure injection recirculation Containment and primary coolant sampling Reactor core isolation cooling Makeup and letdown (PMRs only)

Maste gas (includes headers and cover gas system outside of containment in addition to decay or storage system)

Include a list of systems containing radioactive materials which are excluded from program and provide justification for exclusion.

{2)

Testing of gaseous system should include helium leak detection or equiva-lent testing methods.

(3)

Should consider program to reduce leakage potential release paths due to design and operator deficiencies as discussed in our letter to all operating nuclear power plants regarding North Anna and related incidents, dated October 17, 1979.

(4)

This requirement shall be implemented by applicants for operating license prior to issuance of a full-power license.

Discussion and Conclusions In the document, "Response to NUREG-0578, Short Term Lesson Requirements" as amended through January 26, 1981, and the FSAR, the applicant committed to a program to reduce leakage from systems outside of containment prior to the fuel loading date, in order to satisfy the requirement of III.D.l.l.

The systems to be initially tested, prior to full-power operation, and the 2-27

applicable methods of obtaining actual leak rates has been given.

The commit-ment addresses the North Anna and related incidents letter and provides a

continuing leak reduction program for the systems outside of containment.

We find the proposed program to reduce leakage from systems outside of contain-ment meets the requirements for III.D.l.1 given above and in NUREG-0578,

0660, 0690 and 0737, therefore the program is acceptable for full power operation.

III.D. 2.4 Offsite Dose Measurements Position The NRC will place approximately 50 thermoluminescent dosimeters (TLDs) around the site in coordination with the applicant's and state's environmental moni-toring program.

This action shall be completed prior to issuance of a full power license.

Discussion and Conclusions The Office of Inspection and Enforcement has determined that only 14 TLOs are, required for Diablo Canyon because of its remoteness and its coastal location.

The 14 TLDs have been placed around the plant site.

A program has been established as part of the state of California environmental program to collect the TLDs quarterly and send them to NRC for processing.

Based on the above, we conclude that the Section III.D.2.4 full power requirements of NUREG-0694 have been met.

III.D. 3 '

Im roved In lant Iodine Instrumentation Under Accident Conditions Position (1)

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.

II (2)

Each applicant for a fuel-loading 1 icense to be issued prior to January 1,

l981 shall provide the equipment, training, and procedures necessary to accurately determine the presence of airborne radioiodine in areas within the plant where plant, personnel may be present during an accident.

Clarification Effective monitoring of increasing iodine levels in the buildings under accident conditions must include the use of portable 'instruments using sample media that will collect iodine selectively over xenon (e. g., silver zeolite) for the following reasons:

(1)

The physical size of the auxiliary and/or fuel handling building precludes locating stationary monitoring instrumentation at all areas where airborne iodine'oncentration data might be required.

(2)

Unanticipated isolated "hot spots" may occur in locations where no stationary monitoring instrumentation is located.

2-28

(3)

Unexpectedly high background radiation levels near stationary monitoring instrumentation:after an accideiit may interfere with filter radiation readings.

(4)

The time required to retrieve samples after an accident may result in high personnel exposures if these'ilters are located in high-dose-rate areas,"

'fter January 1, 1981, each applicant and licensee shall have the capability to remove the sampling cartridge to a low-backgrou'nd, low-contamination area for further analysis.

Normally, counting rooms in the auxiliary building will not have sufficiently low backgrounds for such analyses following an accident.

In the low background area, the sample should first be purged of any entrapped noble gases using nitrogen gas or clean air free of noble gases.

The licensee shall have the capability to measure accurately the iodine concentrations present on these'amplers under accident conditions.

There should be 'sufficient samplers to sample all vital areas.

For applicants with fuel-loading dates prior to January 1, 1981, provide by fuel loading (until January 1, 1981) the capability to accurately detect the presence of iodine in'the region of interest following ari accident.

This can be accomplished by using a portable or cart-mounted iodine sampler with attached single-channel analyzer (SCA).

The SCA window should be calibrated to the 365 KeV of iodine-131 using the SCA.

This will give an initial conservative estimate of presence of iodine and can be used to determine if respiratory protection is required.

Care must be taken to assure that the counting system is not saturated as a result of too much activity collected on the sampling cartridge.

Discussion and Conclusions By letters dated 2/29/80, 3/17/80, 3/31/80, 4/9/80. 4/11/80, and 9/22/80, the applicant has submitted commitments and documentation of actions to be taken at Diablo Canyon to implement short-term lesson learned items from NUREGs-0578,

0660, 0694, and 0737.

The applicant has portable air samplers which can be fitted with silver zeolite or impregnated charcoal filters suitable for radibiodine sampling.

Analysis facilities include gamma spectrum analysis systems at the radioanalytical counting room and whole body counting room, with low background counting capability at the Technical Support Center Laboratory; a mobile radioanalytical van; and California Polytechnic University.

An additional portable system will be available prior to fuel loading.

Sample results can be obtained within 10 minutes of start of counting.

Training will be performed for post-accident sampling upon completion of procedures.

The actions and commitments for Diablo Canyon meet our positions in NUREGs-0578,

0660, 0694, 0737 and are therefore'acceptable for full power operations.

An inspection of the sampling and analysis facilities will be performed during a routine. inspection.

III.D.3.4 Control Room Habitabi 1it Position In May 1980 the staff issued the report NUREG-0660, "NRC Action Plan Developed as a Result of TMI-2 Accident."

The report includes Item III.D. 3.4, "Control Room Habitability."

As stated in the report, the applicant must "Identify '5nd evaluate the potential hazards in the vicinity of the site as described in SRP Sections 2 2. 1, 2.2.2, and 2.2.3, confirm that operators in the control room are adequately protected from these hazards,and the release of radioactive gases as described in SRP Section 6.4, and, if necessary, provide the schedule for modifi-cations to achieve compliance with SRP Section 6.4."

Discussion and Conclusions In a letter of May 7, 1980, we advised the applicant of the requirement.

In June of 1980 the staff issued NUREG-0694, "TMI-Related Requirements for New Operating Licenses,"

which states that the requirements of Item III.D. 3.4 must be met prior to the NRC issuing a license to operate Units 1 and 2 of the Diablo Canyon Nuclear Power Station at full power.

Me advised the applicant of thehe requirements and requested that it inform us by letter of the results of the independent:review, and, if necessary, of the schedule for modifications, In a letter dated September 3, 1980, the applicant submitted its 'response to NUREG-0660 and NUREG-0694.

The applicant stated that the habitability of the control room had been reviewed with respect to the effects of offsite

hazards, in accordance with the appropriate sections of the Standard Review Plan, including Regulatory Guides 1.78 and 1.95.

The applicant determined that no design changes are necessary.

Me therefore have concluded that the full power license requirements of Item III.D.3.4 of NUREG-0660, NUREG-0694 and NUREG-0737 have been met.

2-30

IY.

Practices and Procedures IV. F. 1 Power-Ascension Test=

Position IE will monitor the power-ascension test program to confirm that safety is not compromised because of the expanded startup test program and economic 'costs of the delay in commerical operation.

'his action shall be taken 4uring the startup and power-ascension program.

Discussion IE will monitor the power-ascension test program.

~

g 2-31

i'

~

ff' I

D DETER REIR IREMEMTE Mith respect to THI-2 dated requirements, experience with implementation of the dated requirements for licensees of operating reactors and applicants for near-term operating licenses has indicated to the staff that deadlines may be too tight in some cases to allow reasonable time for completion of th work required.

The staff has allowed and would intend to continue to allow case-by-case exceptions to the deadlines 1f good cause is shown.

e The dated requirements are not preconditions for licensing of new plants.

That is, if a completion deadline falls later than the operating license date for a

new plant, then that requirement need not be met by the newly licensed plant until the completion deadline.

If a completion deadline falls before an operating license issuance

date, then that requirement is a prerequisite for the new operating license, except when good cause is shown for the exception.

The licensee will meet the required dates for all of these requirements except for the one listed below.

NUREG-0737 Licensee

~Ri DD Dt II.F.2 Inadequate Core Cooling Instruments 1/1/82 (Incore thermocouple in-containment connectors and junction boxes)

First extended outage after avail abi 1 ity of components II.K. 3. 2 Report on PORV Failures X/Z/8X 3/81 3"1

I.

0 erational Safet I.D.j.

Control Room Desi n Review Pos 1tion Conduct a detailed control-room design review to identify and correct design deficiencies.

This detailed control-room design review is expected to take about a year.

Therefore, 'applicants for operating licenses who are unable to complete this review prior to'ssuance of a license shall make preliminary assessments of their control rooms to identify significant human-factors and instrumentation problems and establish a schedule approved by NRC for correcting deficiencies.

These applicants will be required to complete the more detailed control-room reviews on the same schedule as licensees with operating plants.

This requirements shall be met as follows:

(1}

Applicants for operating licenses whose schedules do not permit a detailed review prior to licensing shall complete a preliminary 'assessment, and it shall be approved'by NRC prior to issuance of the operating license.'2)

Licensees and applicants for'perating licenses shall complete the detailed review, using NRC guidelines (NUREG-0700 to be issued in 1981} on a schedule that will be determined upon issuance of the guidelines.

See letter of October 31, 1980 (NUREG-073?).

Discussion and Conclusion By letter dated March 3.9, 1981 the applicant stated that it will address the review requirements of the main control room in'ccordance with NUREG"0700 after it is issued.

We conclude that the licensee has taken adequate steps to date to meet this requirement.

Based on our review of the applicant's preliminary assesment of its control

room, our onsite preliminary design review and the applicants implementation of measures to correct human factor deficiencies identified in Supplement 10 as noted in the applicant's submittal dated February 26, 1981.

We conclude that the improvements made to the control room will enhance the operation's detection and response capability and will lessen the probability of operator error under stressful conditions to permit safe operation during full power operation.

I.D.2 Plant Safet Parameter Dis la Console Position Install a safety parameter display system that will display to operating personnel a minimum set of parameters which define the safety status of the plant.

This can be attained through continuous indication of direct and derived variables as necessary to assess plant safety, status.

Clarification and functiona1 criteria were provided in NUREG-0696.

Discussion and Conclusion By letter dated March 19, 1981 the applicant has acceptably committed to address the functional criteria of NUREG-0696 when this becomes a requirement.

II.

Sitin and Desi n

II.B.2

~P1 Shi 1di Position With the assumption of a postaccident release of radio'activity equivalent. to that described in Regulatory Guides l. 3 and 1.4 (i.e., the equivalent of 50K of the core radioiodine, 100K of the core noble gas inventory, and 3X of the core solids are contained in the primary coo1ant),

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 review should identify the location of vital areas and equipment, such as the control room, radwaste control stations, emergency power supplies, motor control centers, and instrument areas, in which personnel occupancy may be unduly limited or safety equipment may be unduly degraded by the radiation fields during postaccident operations of these systems.

Each licensee shall provide for adequate access to vital areas and protection of safety equipment by design

changes,

'increased permanent or temporary shielding, or postaccident procedural controls.

The design review shall determine which types of corrective act'ions are needed for vital areas throughout the facility.

V Chan es to Previous Re uirements and Guidance This requirement was originally issued by letters to all operating nuclear power plants, dated September 13 and October 30, '1979, and was incorporated into NUREG-0660.

Significant changes in requirements or guidance are:

(1)

Adds several areas to be evaluated for access to ensure that these areas are not overlooked.

(2)

Specifies that the source term for recirculated depressurized coolant need not be assumed to contain noble gas.since this gas wi'll be released from the liquid when it is depressurized.

(3)

Specifies that certain systems be considered as potential sources and that leakage from systems outside of containment need not be considere'd as potential sources.

(4)

Allows averaging over 30 days of the dose rate criteria for areas requiring continuous occupance and that the control room and techni'cal support center should be considered areas requiring continuous occupancy.

This ensures that the dose rate criteria is applied correctly to these areas; (5)

Specifies source terms to be used in conjunction with Commission Order and Memorandum dated May 23, 1980 (CLI-80-21} on equipment qualification, and specifies schedule in 'above order.

H (6)

Because of difficulty in obtaining equipment (e,g.,

remote-operated valves),

the implementation date is moved to January 1, 1982, or the first outage of sufficient duration thereafter, but no later than July 1, 1982.

3-4

Clarification The purpose of this item is to ensure that licensees examine their plants to determine what actions can be taken over the short-term to reduce radiation, levels and increase the capability of operators to control and mitigate the consequences of, an accident.

These actions should be taken pending conclusions resulting in the long term degraded core rulemaking, which may result in a need to consider additional sources.

4

~

Any area which will or may require occupance to permit an operator to aid in the mitigation of or,recovery from. an accident is designated as a vital area.

For the purposes of this evaluation, vital areas and equipment are not necessar-ily the same vital areas or equipment defined in 10 CFR 73.2 for security purposes.

The security center is listed as an area to be considered as poten-tially vital, since access to this area may be necessary to take action to give access to other areas in the plant.

The control room, technical support center (TSC), sampling station and sample analysis area must be included among those areas where access is considered vital after an accident.

(See Item III.A.1.2 for discussion of the TSC and emergency operations facility.)

The evaluation to determine the necessary vital areas should also include, but not be limited to, consideration of the post-LOCA hydrogen control system, containment isolation reset control area, manual ECCS alignment area (if any), motor control centers, instrument panels, emer gency power supplies, security center, and radwaste control panels.

Dose rate determinations need not be for these areas if they are determined not to be vital.

As a.minimum, necessary modifications must be sufficient to provide for vital system operation and for occupancy of the control

room, TSC, sampling station, and sample analysis area.

~

~

In order to assure that personnel can perform necessary postaccident operations in the vital areas, the following guidance is to be used by licensees to evaluate the adequacy of.radiation protection to the operators:

i

{1)

'Source Term The minimum radioactive source term should be equivalent to the source terms recommended in Regulatory Guides 1.3, 1.4, 1.7 and Standard Review Plan 15.6.5 with appropriate decay times based on plant design (i.e., you may assume the radioactive decay that occurs before fission products can be transported to various systems).

(a)

Liquid-Containing Systems:

100K of the core equilibrium noble gas inventory, 50X of the core equilibirum halogen inventory, and lX of all others are assumed to be mixed in the reactor coolant and liquids recirculated by residual heat removal (RHR), high-pressure coolant injection {HPCI), and low-pressure coolant injection (LPCI), or the equivalent of these systems.

In determining the source term for recirculated, depressurized cooling water, you may assume that the water contains no noble gases.

{b)

Gas-Containing Systems:

100K of the core equilibrium noble gas inventory and 25K of the core equilibrium halogen activity are assumed to be mixed 3-5

in the containment atmosphere.

For vapor-containing lines connected to the primary system (e. g.,

BWR steam lines), the concentration of radio-activity shall be determined assuming the activity is contained in the vapor space in the primary coolant system.

(2)

Systems Containing the Source

)I Systems assumed in your analysis to contain high levels of radioactivity in a postaccident situation should include, but not be limited to, containment, residual heat removal

system, safety injection systems, chemical and volume control system (CVCS), containment spray recirculation, system, sample lines, gaseous radwaste

systems, and standby gas treatment systems (or equivalent of these systems).

If any of these systems or others that could contain high levels of radioactivity were excluded, you should explain why such systems were excluded.

Radiation from leakage of systems located outside of containment need not be considered for this analysis.

Leakage measurement and reduction is treated under Item III.D.l. 1, "Integrity of Systems Outside Containment Likely To Contain Radioactive Material. for PWRs and BWRs."

Liquid waste systems need not be included in this analysis.

Modifications to liquid waste systems will be considered after completion of Item III.D.1.4, "Radwaste System Design Features To Aid in Accident Recovery and Decontamination."

(3)

Dose Rate Criteria The design dose rate for personnel in a vital area should be such that the guidelines of GDC 19 will not be exceeded during the course of the accident.

GDC 19 requires that adequate radiation protection be provided such that the dose to personnel should not be in excess of 5 rem whole body, or its equivalent to any part of the body for the duration of the accident.

When determining the dose to an operator, care must be taken to determine the necessary occupance times in a specific area.

For example, areas requiring continuous occupance will require much lower dose rates than areas where minimal occupancy is required.

Therefore, allowable dose rates will'be based upon expected occupance, as well as the radioactive source terms and shielding.

However, in order to provide a general design objective, we are providing the following dose rate criteria with alternatives to be documented on a case-by-case basis.

The recommended dose rates are average rates in the area.

Local hot spots may exceed the dose rate guidelines.

These doses are design objectives and are not to be used to limit access in the event of an accident.

(a)

Areas Requiring Continuous Occupancy:

<15 mrem/hr (averaged over 30 days).

These areas will require full-time occupance during the course of the accident.

The control room and onsite technical support center are areas where continuous occupance will be required.

The dose rate for these areas is based on the control room occupancy factors contained in SRP 6.4.

(b)

Areas Requiring Infrequent Access:

GDC 19.

These areas may require access on an irregular basis, not continuous occupance.

= Shielding should be provided to allow access at a frequency and duration estimated by the licensee.

The plant radiochemical/chemical analysis laboratory, radwaste

panel, motor control center, instrumentation locations, and reactor coolant and containment gas sample stations are examples of sites where occupance may be needed often, but not continuously.

3"6

(4)

Radiation gualification of Safety-Related Equipment The review of safety-related equipment which may be unduly degraded by radiation during postaccident operation of this equipment relates to equipment inside and outside of the primary 'containment.

Radiation source terms calculated to determine environmental qualification of safety-related equipment consider the following:

~

(a)

LOCA events which completely depressurize the primary system should consider releases of the source term (100K noble gases, 50K iodines, and particulate) to remain in the primary coolant.

This method is used to determine the qualification doses for equipment in close proximity to recirculating fluid systems inside and outside of containment.

Non-LOCA events both inside and outside of containment should use 10K noble gases, 10K iodines, and OX particulate as a source term.

The following table summarizes these considerations:

Containment LOCA Source Term (Nob1 e Gas/Iodine/

Particulate)

Non-LOCA High-Energy Line Break Source Term (Noble Gas/Iodine/Particulate}

Outside Inside x

(100/50/1) in RCS

~Lar er of in containment x

(1O/1O/O) in RCS (10/10/0) in RCS or (1OO/5O/1) in RCS Discussion and Conclusion Desi n Review of Plant Shieldin and Vital Area Access By letters dated February 20, 1981 and March 17, 1981, the applicant has submitted commitments and documentation of actions to'be taken at Diablo Canyon by PG8E to implement short'erm lesson learned item 2. 1.6.b/II.B.2 of NUREG-0578/0737.

The applicant's Shielding Design Review for the Diablo Canyon facility has been evaluated in accordance with NUREGs-0578/0737.

The descriptions meet our positions and requirements for full power operations for post-accident plant shielding and vital area access.

The source terms used by the applicant in its evaluation are those established in NUREG-0578 and TID 14844.

To determine maximum dose rates and integrated 3-7

doses, minimum delay times for release of radioactive materials and minimum dilution factors were used.

A delay time of 30 minutes, based on Design Base Accident Transport times, was utilized for transport criteria,

however, zero time delay was assumed for dose calculations.

Dilution of the source term for containment release included the Reactor Coolant System volume, the Boric Acid Injection Tank Volume, and a portion of the Refueling Mater Storage Tank volume.

For radiation dose calcula'tions, only the Reactor Coolant System volume was utilized.

Five separate, accident.scenarios were selected for analysis by PG5E, and all systems involved in the scenarios were assumed to be sources for dose calculations.

Computer codes used in calculations included ORIGEN,.gAONOD, and G~.

Those systems assumed to carry highly radioactive materials in a postaccident situation are the charging portion of the chemical and volume control system (including its safety injection flowpath through the Boron Injection Tank),

the Safety Injection system, the Residual Heat Removal

system, the Containment Spray system discharge

headers, and the sampling systems associated with the Emergency Sampling Compartment.

Redundant

systems, the containment, and the plant vent stack (in special cases),

were also identified and evaluated.

Process

systems, including gase'ous radwaste systems were not included, since they will be iso1ated by containment isolation and will not be subjected to the source term.

Those systems and areas where PG8E has determined that access or occupancy are necessary for postaccident situations are the Control Room, the Technical Support Center, the Hot Shutdown Panel, and the Emergency Sampling Compartment.

Analysis shows that all of these areas are accessible and can be occupied consistent with criteria specified in NUREG-0737 and Regulatory Guide 1.4.

Dose estimates for occupancy or operations in these areas are well within the applicable dose criteria of Standard Review Plan 6.4 and General Design Criteria 19.

Other areas do not-require access due to systems redundancy and adequate control capability from accessible areas.

PGSE has also provided estimates of the projected. doses to individuals for expected occupancy of vital areas, including access to these areas.

Area dose rate maps for potentially occupied areas are included in the analysis;

however, these should be expanded in scope to include projected post accident dose rates for occupied areas, potentially occupied areas, and travel routes at the facility.

The licensee has committed to provide this information prior to the issuance of a full power license.

No shielding construction or modifications, except those committed for January 1;

1982 under item II.B.3 of this supplement for postaccident sampling are needed at Diablo Canyon.

Onsite inspection of the Diablo Canyon shielding and access provisions will be conducted during routine inspections.

Plant Shielding and-vital area access for the Diablo Canyon plant meet our positions in NUREGs-0578/0737 and are acceptable for full power operations.

Our 'evaluation of the radiation qualification of safety-related equipment will be addressed in a future supplement which is expected to be issued in mid-May, 1981 3-8

II.B.3 Postaccident Sam lin Position A design and operational, review of the reactor coolant and containment atmosphere sampling line systems shall be performed to determine the capability of personnel to promptly obtain (less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) a sample 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, respectively.

Accident conditions should assume a Regulatory Guide 1.3 or 1.4 release of fission products.

If the review indicates that personnel could not promptly and safely 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 quantify (in less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) certain radionuclides 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 nonvolatile isotopes (which indicate fuel melting).

The initial reactor coolant spectrum should correspond to a Regulatory Guide'.3 or 1.4 release.

The review should also consider the effects of'irect radiation from pip'ing 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 modifica-tions 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 provided to perform boron and chloride chemical analyses assuming a highly radioactive initial sample (Regulatory Guide 1.3 or 1.4 source term).

Both analyses shall be'apable of being completed promptly (i.e., the boron sample analysis within an hour and the chloride sample analysis within a shift).

Chan es to Previous Re uirements and Guidance This requirement was originally issued to all operating plants by letters dated September 13 and October 30, 1979; Significant changes in requirements or guidance are:

(l)

Specifies that licensee may use online sampling and analysis to meet the 3-hour time requirement but must provide capability to remove grab samples of reactor coolant and containment atmosphere for separate analysis.

(2)

Implementation date has been changed to January j., 1982.

(3)

Provides design guidance for sampling and analytical capability.

Clarification The following items are clarifications of requirements identified in NUREG-0578, NUREG-0660, or the September 13 and October 30, 1979 clarification letters.

3"9

(1)

The licensee shall have the capability to promptly obtain reactor coolant samples and containment atmosphere samples.'he combined time allotted for sampling and analysis should be 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> or less from the time a decision is made to take a sample.

44 (2).

The licensee shall establish an onsite radiological and chemical analysis capability to provide, within the 3-hour time frame established

above,

'uantification of the following:

(a) certain radionuclides in the reactor coolant and containment atmosphere that may be indicators of the degree of core damage (e.g.,

noble gases; iodines and cesiums, and nonvolatile isotopes);

{b) hydrogen levels in the containment atmosphere; (c) dissolved gases (e. g., H~), chloride {time allotted for analysis subject to discusson below),

and boron concentration of l.iquids.

(d)

Alternatively, have inline monitoring capabilities to perform all or part of the above analyses.

(3)

(4)

(5)

Reactor coolant and containment atmosphere sampling during postaccident conditions shall not require an isolated auxiliary system [e.g., the letdown system, reactor water cleanup system (PWCUS)] to be placed in operation in order to use the sampling system.

Pressurized reactor coolant samples are not required if the licensee can quantify the amount of dissolved gases with unpressurized reactor coolant samples.

The measurement of either total dissolved gases or H2 gas in reactor coolant sample is considered adequate.

measuring the 02 concen-tration is recommended, but is not mandatory.

The time for a chloride analysis to be performed is dependent upon two factors:

(a) if the plant's coolant water is seawater or brackish water and (b) if there is only a single barrier between primary containment systems and the cooling water.

Under both of the above conditions the licensee shall provide for a chloride analysis within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the sample being taken.

For all other cases, the licensee shall provide for the analysis to be completed within 4 days.

The chloride analysis does not have to be done onsite.

(7)

The design basis for plant equipment for reactor coolant and containment atmosphere sampling and analysis must assume that it is possible to obtain and analyze a sample without radiation exposures to any-individual exceeding the criteria of GDC 19 {Appendix A, 10 CFR Part 50) {i.e.,

5 rem whole body, 75 rem extremities).

(Note that the de'sign and operational review criterion was changed from the operational limits of.10 CFR Part 20 (NUREG-0578) to the GDC 19 criterion (October 30, 1979 letter from H.

R.

Denton to all licensees).)

The analysis of primary coolant samples for boron is required for PWRs.

{Note that Revision 2 of Regulatory Guide 1.97, when issued, will likely specify the need for primary coolant boron analysis capability at 8WR p'l ants. )

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(8) If'inline monitoring is used for any sampling and analytical capability specified herein, the licensee shall provide backup sampling through grab

samples, and shall demonstrate the capabiIity of analyzing the samples.

Established planning for. analysis at offsite facilities is acceptable.

Equipment provided for backup sampling shall be capable of providing at least one sample per week until the accident condi ton no longer exi sts.

(9)

The licensee's radiological and chemical sample analysis capability shall include provisions to:

(a)

Identify and quantify the isotopes of the nuclide categories discussed above to:levels corresponding to the source terms given in Regulatory Guide 1.3 or 1.4 and 1.7.

Mhere necessary and practicable, the ability to dilute samples to provide capability for measurement and reduction of personnel exposure should be provided.

Sensitivity of onsite liquid sample analysis capability should be such as to permit measurement of nuclide concentration in the range from approximately 1 pCi/g to 10 Ci/g.

(b)

Restrict background levels of radiation in the radiological and chemical analysis facility from sources such that the sample analysis will provide results with an acceptably small error (approximately a

factor of 2).

This can be accomplished 'through the use of sufficient shielding around samples and outside sources, and by the use of ventilation system design which will control the presence of airborne radioactivity.

(10) Accuracy, range and sensitivity shall be adequate to provide pertinent data to the operator in order to describe radiological and chemical status of the reactor coolant systems.

(ll) In the design of the postaccident sampling and analysis capability, consideration should be given to the following items:

I (a)

Provisions for purging sample lines, for reducing plateout in sample lines, for minimizing sample loss or distortion, for preventing blockage of sample lines by loose material in the RCS or containment, for appropriate disposal of the samples, and for flow restrictions to limit reactor coolant loss from a rupture of the sample line.

The postaccident reactor coolant and containment atmosphere samples should be representative of the reactor coolant in the core area and

~the containment atmosphere following a transient or accident.

The sample lines should be as short as possible to minimize the volume of fluid to be taken from containment.

The residues of sample collection should be returned to containment or to a closed system.

(b)

The ventilation exhaust from the sampling station should be filtered with charcoal adsorbers and high-efficiency particulate air (HEPA) filters.

3-11

Discussion and Conclusions Postaccident Sam lin -ALARA Evaluation'y letters dated February 29, 1980, and February 19 and February 20,'981, the applicant has submitted commitments and documentation of actions to be taken at Diablo Canyon to implement short term lessons learned items from NUREGs-0578/0737.

The applicant has completed a review of post-accident sampling which has evaluated actions to minimize personnel radiation exposure.

Procedures and sampling facility designs have been planned to minimize exposure in several ways:

a Remote Sample Station with location in low post-accident backgrdund area; accessibility from grade level; shielded sampling lines; spaciousness; minimum sample quantity; minimal effort involved in sampling; ventilation systems; remote valve control panels; and direct control room communications.

Special handling and counting techniques, including dilution, will be employed for high level samples.

The applicant's plans and actions for post-accident sampling are ALARA, and should enable post-accident sampling without excessive exposure.

Analyses will include on-line analysis of chloride,

hydrogen, oxygen, pH and conductivity.

Provisions also include sample inlet for boron concentration after dilution at the liquid sampling panel, and inlets for reactor coolant, pressurizer,

letdown, and sump samples.

Air sampling will'e provided for the containment atmosphere.

Moreover, the applicant has the P&ID drawings and an estimate of the sampling and analyses line, which we find to be acceptable.

Based on our evaluation, we find that the design meets the requirements of NUREGs-0578, 0737 and Regulatory Guide 8.8 and is therefore acceptable.

Modifications are planned for completion by January 1, 1981.

On-site verification of modifications will be conducted during routine inspection.

II.D.1 Performance Testin of Boilin Mater Reactor and Pressurized ater Reactor e )e and Safet Va ve Position Pressurized Mater Reactor and Boiling Mater Reactor licensees and applicants shall conduct testing to qualify the reactor coolant system relief and safety valves under expected operating conditions for design basis transients and accidents.

Clarification Expected operating conditions can be-determined through the use of analysis of accidents and anticipated operational occurrences referenced in Regula-tory Guide 1.70, "Standard Format and Content of Safety Analysis Reports."

2.

This testing is intended to demonstrate valve operability under various flow conditions; that is, the ability of the valve to open and shut under.

the v'arious flow conditions should be demonstrated.

3.

Not all valves on all plants are required to be tested.

The valve testing may be conducted on a prototypical basis.

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4.

The effect of piping on valve operability should be included in the test conditions.

Not every piping configuration, is required to be tested, but the configurations that are tested should produce the appropriate feedback effects as seen by the relief or safety valve.

5.

Test data should include data that would permit an evaluation of discharge piping and supports if those components are not tested directly.

6.

Additional clarifications regarding documentation submittal dates and a

requirement to qualify PWR block valves by July 1, 1982 were transmitted to near term l.icense applicants in NUREG-0737, "Clarification of TMI Action Plan Requirements,"

by letter from D. Eisenhut dated October 31, 1980.

Discussion and Conclusions The applicant has stated that it will participate in the EPRI/NSAC program to conduct performance testing of PWR relief and safety valves and associated, piping and supports.

The applicant has referenced the proposed EPRI program

("Program Plan for the Performance Verification,of PWR Safety/Relief Valves, and Systems,"

dated December 13, 1979) for the performance testing of these valves.

Additionally, by letter of February 26, 1981, the. applicant has responded to the,clarifications and requirements of NUREG-0737.

A description of the EPRI/NSAC test program was provided to NRC,by EPRI in, December 1979 and an. updated revision of the program description was'rovided in July 1980.

The staff has reviewed these descriptions and is generally in agreement that the NUREG-0737 technical requirements for relief and safety valves and associated piping and supports can be met, subject to receipt of additional information which was requested by letter of November 26, 1980 to Russell C. Youngdahl.

By letter of December 15, 1980 from R.

C.

Youngdahl to D. fisenhut, EPRI has responded to both the staff's November 26, 1980 letter and NUREG-0737.

Pacific Gas and Electric Company has referenced the above December 15, 1980 EPRI response in their letter of February 26, 1981.

The staff has not completed its review of the EPRI December 15, 1980 letter.

However, in that response EPRI has taken exception to the documentation sub-mittal dates specified in NUREG-0737 and has stated that a detailed block valve test.program cannot be resolved unti 1 after the completion of the ongoing relief and safety valve program scheduled for completion July 1, 1981.

We have recently been informed by EPRI that certain block valves "did not perform satis-factorily during testing.,

However, the block valves installed at Diablo Canyon were manufactured by,the,VELAN Company and.a VELAN block valve of the type installed at Diablo Canyon did perform satisfactorily during testing.

On completion of the review of the EPRI December 15, 1980 submittal, the staff will arrive at a generic resolutiog regarding.the NUREG-0737 required documen-tation submittal dates which will be applicable to all operating reactors.

We will,require that Pacific Gas and Electric Company provide documentation in 3"13

accordance with this schedule for the Diablo Canyon Unit relief valves, safety

valves, block valves and associated piping.

The Pacific Gas and Electric Company has committed to the requirements of this item to the extent practicable at this time.

The applicant is participating in the EPRI/NSAC PWR safety and relief valve performance verification test program and is monitoring this program to assure that the results apply to the Diablo Canyon plant specific valves and associated piping and supports.

In addition, the applicant has also committed to qualification of'- block valves'y July 1, 1982.

We believe that this commitment provides adequate assurance that the requirements for performance testing of relief valves, safety valves, block valves, and associated piping will'be satisfied.

The basis for accepting this commitment is our review to date of the EPRI/NSAC relief and safety valve test program and our continued review of this program to confirm that it is acceptable for the Diablo Canyon plant specific design.

II.F. 1 Additional Accident Monitorin Instruments Position (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 necessary to cover the ranges of interest.

Noble gas effluent monitor s with an upper'ange capacity of 10 pCi/cc (Xe-133) are considered to be practical.

Noble gas.effluent monitoring shall be provided for the total range of concentration extending from normal condition (as low as reasonable achievable (ALARA)) concentrations to a maximum of 10 pCi/cc (Xe-133).

Multiple monitors are considered to be. necessary to cover the ranges of interest.

The range capacity of individual monitors should overlap by a factor of ten.

(2)

Capability for effluent monitoring of radioiodines for the accident shall be provided with sampling conducted by adsorption on charcoal or other media, followed by onsite laboratory analysis.

Iodine gaseous effluent monitors for the accident condition are not considered to be practical at this time.

(3)

In containment radiation-level monitors with a maximum range of 10 rad/hr shall be installed.

A minimum of two such monitors that are physically separated shall be provided.

Monitors shall be developed and qualified to function in an accident environment.

(This requi,rement was later revised to provide for a photon-only measurement with an upper range of 107 R/hr.)

{4)

A continuous indication of containment pressure shall be provided in the control room of each operating reactor.

Measurement and indication capability shall include three times the design pressure.of the containment f'r concrete, four times the design pressure for steel, and -5 psig for all c'ontainments.

{5)

A continuous indication of containment water level shall be provided in the control room for all plants.

A narrow range instrument shall be 3"14

provided to cover the range from the bottom to the top of the containment sump.

A wide range instrument shall also be provided to cover the range from the bottom of the containment to the elevation equivalent to a 600,000 gallon capacity.

(6)

A continuous indication of hydrogen concentration in the containment atmosphere shall

'be provided in the control room.

Measurement capability shall be provide'd over the range of 0 to 10K hydrogen concentration under both positive and negative ambient pressure.

These requirements shall be met prior to January 1, 1982.

See October 31, 1980 "letter (NUREG-0737).

Discussion and Conclusions In regard to position (1), the potential gaseous effluent release points at-Diablo Canyon, Unit Nos.

1 and 2, consist of the plant vent stack, the atmospheric steam

dump, and the steam generator blowdown tank vent.

The applicant has committed to install mid/high-level'oble gas effluent monitors at these three points by January 1,

1982.

The monitors will be designed to meet the requirements and satisfy the characteristics given in NUREG-073?

and is, therefore, acceptable.

In regard to position (2), the potential gaseous" effluent release points at Diablo Canyon, Unit Nos. 1'nd 2, are the same as for the noble gases.

The applicant has, committed to install new isokinetic probes in each release line by January 1, 1982.

The applicant will utilize the new and existing isokinetic probes and the off-line sampling systems to quantify release rates during an accident.

Silver-zeolite/charcoal cartridges will be reverse blown with air to purge interfering noble gases.

The present counting room will provide analysis.

The system for measuring radioiodine for the accident meets the requirements and satisfies the characteristics given in NUREG-0737, and is, therefore, acceptable.

In regard to position (3), Diablo Canyon will have two high-range containment radiation monitors installed; PGEE has committed to have them installed in accordance with the implementation dates of NUREG-0737.

The instruments will measure'e the range 1 to 10~ R/hr, with the capability to measure 60 Kev photons.

Separate power supplies from separate vital buses will be provided, along with control room readout and recording capability on the Post Accident Monitoring Panel.

Location in-contai,nment is such that there, are no in-containment shielding interferences or potential "hotspots" to interfere with representative readings.

The applicant has committed to provide calibration for the equipment in accordance with the positions outlined in NUREG-0737.

The seismic and environmental qualifications will comply with Regulatory Guides 1.89 and 1.100.

The commitments for Diablo Canyon meet our positions in NUREGs-0578/

0664/0694/0737 and are acceptable.

A.review of the equipment installation and operation will be performed during a routine inspection by our Office of Inspection and Enforcement.

In regard to position (4), the Diablo Canyon containment is a steel lined, reinforced concrete structure designed for a maximum pressure of 54 psig concurrent with a safe shutdown earthquake, or ?0 psig without an earthquake, The applicant is adding containment pressure transmitters with a range of 0 to 3"15

200 psig connected to control room recorders.

This instrumentation will complement the existing post-'accident containment pressure indicators which have a range from -5 to '+5 psig and meet the requirements of Appendix B; although they do not have recording capabilities.'

The transmitters have an accuracy of 0.25K of full scale with a full scale step response time of less than 200 milliseconds.

The recorders have an

'ccuracy of 0.5X of full scale with a full scale'tep response time of, 3 seconds.

We will conditio'n th'e full power license to require. installation of the containment pressure monitor meeting the requirements of. NUREG-0737 by January 1,

1982.

We conclude, based on.our evaluation an8 the license condition noted above,'hat the containment monitor is acceptable.

With respect to position (5), the applicant has stated that the existing containment sump level"instrumentation at Diablo Canyon is post-accident qualified.

The monitor range is'rom the recirculating sump bottom (elevation 88') to seven feet above the containment floor (elevation'8').

This level is approximately two feet above the worst calculated accident flooding level (elevation 96'"'which accounts for 535,000 gallons of water as described in FSAR Section 15.4. 1).

The accuracy of the loop is +2.5X, or +3 inches.

The bottom of the reactor cavity is at elevation 63'";- Wide-range'monitors will be installed and will have recorders in the control room.

4 Mutually redundant, loops are provided which are wired and separated in accordance with IEEE Class 1E requirement's'.

Each loop utilizes a Barton Model 764. dif-ferential pres'sur'e transmitter which is mounted at floor level because of reference leg limitations. 't is qua'lified for submerged p'ost-accident.

opera-'ion.

Sealed capillary sensing legs connect ea'ch transmitter to Barton Model 351 bellow sens'ors mounted in the "reactor cavity, the lowest. point in contain-ment (wide-range),

and in the containment recirculation sump,(narrow 'range).

The other lag of each transmitter is connected to sensors above the maximum flood level.

h The wide-range recorders are mounted on the Post-Accident Monitoring Panel, and the narrow range indicators are mounted on the Main Control Board.,

The narrow range indicators are used when operating pumps for recirculation and are loCated above the respective recirculation control switches.

The existing recirculation sump level instrumentation meets the requirements for narrow range, except that the level range is from the bottom of the recirculation sump-to the highest level used for the recir'culation mode.

r The wide-range containment water level instrumentation has been purchased and will be insta'lied prior to the required implementation date.

In addition to those described

above, there are three small sumps in containment (250 gallons).

Two are at the containment" floor level, and one is in the reactor cavity.

These sumps presently have auxiliary control board level indication.

The applicant has committed to add the level indication on the Post Accident Monitoring Panel.

3-16

Me will condition the full power license to require installation of the con-tainment water level monitor, meeting the requirements of NUREG-0737, by January 1, 1982.

Based on our evaluation and the, above cited condition to the

license, we conclude that the containment water level monitoring system is acceptable.

In regard to pos'ition.(6), the applicant has stated that two mutually redundant hydrogen monitors have been delivered to the Diablo Canyon Site.

They are to be mounted outside of the containment with sample lines from two widely separated points inside the containment.

They are. capable of monitoring a range 0-1QX hydrogen to an accuracy of +2X of full scale.

The readouts are mounted in the post-LOCA sampling room.

In addition, recorders are mounted on the Post-Accident Monitoring panel (PAN-1).

They will not normally be running, but can be energized and operated within 30 minutes.

The instrumentation has been procured and will be operational by the required implementation date.

Me will condition the full power license to require installation of the hydrogen monitors meeting the requirements of HUREG-0737 by January 1, 1982.

Me conclude, based on our evaluation and the license condition cited above, that the hydrogen monitors are acceptable.

II.F. 2 Inade uate Core Coolin Instruments Position Licensees shall provide a description of any additional instrumentation or controls (primary or backup) proposed for the plant to supplement existing instrumentation (including primary coolant saturation monitors) in order to provide an unambiguous, easy-to-interpret indication of inadequate core cooling (ICC).

A description of the functional design requirements for the system shall also be included.

A description 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.

Chan es to Previous Re uirements and Guidance (1)

Specify the "Design and gualification Criteria" for the final ICC monitor-ing system in section, "Clarification" (items 7, 8, and 9), Attachment 1, and Appendix A.

(2)

Specify complete documentation package to allow NRC evaluation of the final ICC monitoring systems to begin on January 1, 1981.

(3)

No preimplementation review is required but postimplementation review of installation and preimplementation review before use as a basis for operator decisions are required.

(4)

Installation of additional instrumentation is now required by January 1,

1982.

(5)

Clarification item (6) has been expanded to provide licensees/applicants with more flexibilityand diversity in meeting the requirements for determining liquid level indication by providing possible examples of alternative methods.

3"17

Discussion and Conclusions In its February 6, 1981 response to NUREG-0737 requirements the applicant pro-vided a proprietary submittal "PG8E

-Response to Item II.F. 2 Instrumentation for Detection of Inadequate Core Cooling."

The existing instrumentation in the Diablo Canyon facility for detection of inadequate core cooling (ICC) consists of redundant wide range reactor coolant

pressure, wide range coolant, temperature sensors (total 8 RTD's, one for each hot leg and one for each cold leg in each loop),

65 core exit thermocouples, and a Combustion Engineering subcooling margin monitor which has temperature input from each of the reactor coolant system hot legs (4 RTD's) and from six core exit thermocouples, and pressure input from two pressure measurements from hot legs.

PGKE is adding a reactor vessel level instrumentation system (RVLIS), designed by Mestinghouse, to supplement existing instrumentation in determining the-existence of ICC.

The description of the RVLIS in the app'licant's February 6, 1981 submittal, including hardware description, microprocessor

system, resistance temperature detectors, RVLIS valves and transmitters, hydraulic isolators, and sensors are being reviewed by the staff.

The staff has not approved the generic Mestinghouse revised analysis and guidelines required by Task Action Plan item I. C.1{3) and clarified in NUREG-0737.

Therefore the pilot monitoring discussed in it'em I.C.8, Pilot Monitoring of Selected Emergency Procedures, was conducted using interim guidelines that have been approved by the staff.

The adequacy of the vendor provided guidelines was discussed in a meeting with PG8E personnel on November 3 and 4, 1980.

The draft procedures we reviewed for I.C.8 reflected the Mestinghouse analysis of small-break LOCAs and inadequate core cooling in accordance with a license requirement and Task Action Plan

{NUREG-0660) item I.C. 1.

However, on November 10, 1980 the Mestinghouse Owners'roup provided additional guidelines for the mitigation of inadequate core cooling.

The staff has not completed its review of these guidelines.

The changes made to the draft procedures and any additional changes that may result from our review of the November 10, 1980 guidelines must be made and the Diablo Canyon operators must be trained on the changes and their bases prior to operation above 5X of the rated power level.

The Office of 18E will verify that these requirements are.satisfied.

Based on our review of the emergency procedures and our observation of the procedures being implemented on the simulator and in the plant walk-through, we have concluded that when the required changes have been made to the proce-dures as specified in Section I.C.8, the Diablo Canyon emergency operating procedures will be acceptable for operation at power levels up to 100 percent of rated power.

Future actions required by Task Action "Plan items I. C.l. a(3),

Transients and Accidents and I.C.9,'Long-Term Program Plan for Upgrading of Procedures may require future revisions to the emergency procedures.

In response to a staff request for additional information to supplement the February 6, 1981 submittal, the applicant has provided additional information on March 19, 1981 including an evaluation of the subcooling meter and documentation required by NUREG-0737 Section II,F.2.

3"18

The subcooling margin monitoring system is environmentally qualified in accord-ance with Regulatory Guide 1.89 except for the backup analog recorder which is not subject to a harsh environment and the incore thermocouple inputs which are being upgraded and will be completed during the first refueling outage.

No single failure will prevent the operator from determining the subcooling margin.

All components are redundant except for the calculator itself.

Should the calculator fail, the operator will monitor the loop temperature and pressure and then determine the degree of subcooling using steam tables and procedures specifically provided for this function.

The monitor is powered by a Class IE power source.

The staff has reviewed.and found it to be acceptable.

In the February 6,. 1981 submittal the applicant has addressed the conformance of the incore thermocouples to the requirement of Appendix B, "Design and gualification Criteria for Accident Monitoring Instrumentation,"

provided in NUREG-0737, "Clarification of'MI Action Plan Requirements."

The incore thermocouple systems will satisfy these requirements by the required implementation date except for the following:

(1)

The finalized system will be seismically and environmentally qualified in accordance with requirements of the criteria.

The applicant has proposed replacement of the incore thermocpuple in-containment connectors and junction boxes with environmentally qualified equipment during the first refueling outage since the material will not be available by the required implementation date.

We will require that the qua1ified equipment be insta11ed during the first extended plant outage following component availability.

All other aspects of the seismic and environmental qualifications will be completed by the required January 1,

1982 date, F

(2)

The 65 incore thermocouples (TCS) will be separated into two groups (one covering 32 TCS and one covering 33 TCS) with separate readout indicators to provide redundant indication as backup displays with readout ranging from 200'F to 2300'F.

Also all thermocouples signals are routed to the plant computer for the core mapping function as a primary operator display.

The qualification criteria requires that the transmission of signals for other use should be through isolation devices that are designated as a

part of the monitoring instrumentation and meet the provisions of Regula-tory Guide 1;75.

The applicant has noted that the computer has input isolation; however, it does not satisfy the requirements to be classified as Class lE and conform to more stringent isolation requirements.

Since

'the computer is a common element of this redundant indication system, full conformance to the isolation criterion has not been provided.

Therefore, we will require that the applicant provide additional analysis to justify that the plant computer is not a source of common failure for the redundant indication of core region outlet temperatures, or that there is a low probability that such a failure could occur and that if it did,-appropriate action could be taken to restore the indication system to an operable status.

We will pursue this matter further and effect a resolution prior to the required implementation date for this upgrade in inadequate core cooling instrumentation.

Prior. to January 1, 1982. the applicant has committed (1) to install the RVLIS, (2) to complete the reactor coolant pressure transmitter relocation, and 3-19

{3) complete upgrading the incore thermocouple readouts for both, the primary operator displays and the backup displays.

The primary operator displays will.

utilize an alphanumeric display located remotely from the computational system which can display all thermocouple temperatures.

All readings (ranging from 200 F to 2300~F) will be printed out on a hard copy printer.

The applicant has also committed in the. March 19, 1981 letter to provide.

additional submittals with respect to the report on the results of the portions of ongoing RVLIS test programs which are scheduled to be completed by November 1981 and the in-situ calibration document prior to January

1982, and a summary of key operator actions consistent with current procedures by April l5,'981.

The staff has reviewed the applicant's commitments stated above and has con-cluded that they are acceptable, except for issues on the applicant's proposed schedule for completing the upgrade incore thermocouple wiring and the analysis for isolation devices'.

The staff has performed an acceptance review of the documentation which was submitted on February 6 and March 18, 1981.

Based on the results of our

review, we conclude that the description of the instrumentation for detection of ICC for Diablo Canyon Facility meets the documentation requirements of NUREG-0737,Section II.F.2, and is acceptable for full-power operation.

We will condition the full-power license to require that, prior to January

1982, the issues on the isolation devices used to isolate redundant core region outlet thermocouples at the plant computer be resolved.

Further, the completion of the upgrading of the incore thermocouple wiring shall be completed during the first extended outage following component availability; and that prior.to operating above 5X of rated power the applicant must make certain procedural revisions, train the operators, on these revisions, and submit these procedures for staff review.

The ICC system will be reviewed for. acceptability after installation, testing, and calibration of the reactor vessel level instrumenta-tion is complete.

II.K. 2 Commission Orders on Babcock and Wilcox Plants=

II.K.2 Item 13 Thermal-Mechanical Re ort The Westinghouse Owners Group, of which PG&E is a member, will address the NRC requirements of detailed analysis of the thermal-mechanical conditions in the reactor vessel during recovery from small breaks with an extended loss of all feedwater.. This program is scheduled to be completed and documented to the NRC by January 1, 1982 and will consist of analyses for generic Westinghouse PWR plant groupings.

Following the completion of this generic program, PG&E has committed to supply a plant specific analysis based on the generic analysis, if required.

This schedule is consistent with the requirements of NUREG-0737, and we find the PG&E commitment acceptable.

II.K.2 Item 17 Voidin in RCS The Westinghouse Owners Group, of which PG&E is a member, is addressing the potential for void formation in the reactor coolant system (RCS) during natural 3-20

circulation cooldown conditions, as described in Westinghouse Letter NS-TMA-2298 PG8E believes the results of this effort will address the NRC requirement for analysis to the NRC before January 1, 1982.

This schedule is consistent with the requirements of NUREG-0737, and we find the PGRE commitment acceptable.

II.K. 2 Item 19 Benchmark Anal sis Se uential AFE Flow The transient analysis

code, LOFTRAN, and the present small-break evaluations analysis

code, WFLASH, are both undergoing benchmar king against plant informa-tion or experimental test facilities.

These codes are also being compared with each other.

The Westinghouse Owners Group, of which PG8E is a member, will provide a report addressing the benchmarking of these codes by January 1,

1982.

This schedule is consistent with the requirements of NUREG-073?,

and"we fi'nd the PG8E commitment acceptable.

II.K.3 Final Recommendations of 880 Task Force II.K.3.2 Re ort on Overall Safet Effect of Power-0 crated Relief Valve

~IS Position (1)

The licensee should submit a report for staff review documenting the various actions taken to decrease the probability of a small-break loss-of-coolant accident (LOCA) caused by a stu'ck-open power-operated relief valve (PORV) and show how those actions constitute sufficient improvements in reactor safety.

(2)

Safety-valve failure rates based on past history'of the operating. plants designed by the specific nuclear steam supply system (NSSS) vendor should be included in the report submitted in response to (1) above.

Discussion and Conclusion In a letter dated February 13, l981 the applicant has proposed to provide the report required by NUREG-0737, II.K.3, Item 2, in parch 1981.

It has stated that the Westinghouse Owners Group, of which PGEE is a member, is in the process of developing a report in response to this item.

Although the appli-cant's planned schedule is two months later than the NUREG-0737 schedule for implementation of this item, we conclude that the schedule as proposed satisfactorily meets the intent of implementation of this item, and is therefore acceptable.

II.K.3 Item 5 Automatic Tri of RCPs The applicant has referenced WCAP-9584, an analysis performed by the Westing-house Owners Group using"the Westinghouse small-break evaluation model (WFLASH) to show that ample time is available for the op'erator to trip reactor coolant 3-21

t pumps following certain small breaks.

In addition, the owners gr'up is support-ing a best-estimate study using the NOTRUMP computer code to demonstrate that tripping the reactor coolant pump at the worst trip time after a small-break will lead to acceptable results.

Using both of these analysis methods (WFLASH and NOTRUMP), the Westinghouse Owners Group has performed post-test predictions of LOFT experiment L3-6.

The input data and model to be used with WFLASH on LOFT L3-6 was submitted by the owners group to the staff on December 1, 1980.

The LOFT prediction using WFLASH is presently under review.

PGKE has committed to supply justification that these models and analyses apply to Diablo Canyon, within 3 months after NRC approval of the models.

This schedule is consistent with the requirements of NUREG-0737, and we find the PG8E committment acceptable.

If indicated by results of the approved analyses, we will require Diablo Canyon to implement an automatic trip of the RCPs on a schedule consistent with the intent of NUREG-0737.

II.K.3.17 ECCS Outa es Diablo Canyon has provided a commitment to adopt a program to identify future outage dates and lengths of outages.

This commitment has included an identifi-

'ation of the listings to be included in their reports and a further commitment to submit a greater detailed breakdown of the reported data.

This proposed listing agrees with the required areas identified in NUREG-0737 and is accept-able for full-power operation.

II.K.3 Item 25 Effect of Loss of A-C Power On Pum Seals In a letter dated February 13,

1981, PG8E identified that component cooling water pumps and containment isolation valves in the component cooling system providing cooling water to the RCP thermal barriers have emergency onsite backup power and their operability would not be lost with a loss of A-C power.

We find this capability acceptable to meet the requirement for full-power operation.

II.K.3 Item 31 Com liance with CFR 50.46 In a submittal dated February 26, 1981, the applicant has committed to submit a new analysis (using a new and approved Westinghouse model in accordance with the NRC schedule) if the conclusions of NUREG-0737, Item II.K. 3. 30 indicate that the present small break LOCA analysis for Diablo Canyon is not in conformance with 10 CFR 50.46.

We find this commitment acceptable for full-power operation.

3"22

III.

EMERGENCY PREPARATIONS AND RADIATION PROTECTION III.A.1.2 U

rade Emer enc Su ort Facilities Position Provide radiation monitoring and ventilation systems, including particulate and charcoal filters, and otherwise increase the radiation protection to the onsite technical support center to assure that personnel in the center will not receive doses in excess of 5 rem to the whole body or 30 rem to the thyroid for the duration of the accident.

Provide direct 'display of plant safety system parameters and call up display of radiological parameters.

For the near-site Emergency Operations Facility, provide shielding against direct radiation, ventilation isolation capability, dedicated communications with the onsite technical support center and direct display of radiological and meteorological parameters.

This requirement shall be met by January 1, 1981, although the safety parameter information requirements will be staged over -a longer period of time.

Discussion and Conclusions The above requirements will be revised, upon Commission approval,,by those set forth in NUREG-0696, "Functional Criteria for Emergency Response Facilities."

The revised requirements have been published in final form in February 1981.

The NUREG document specifies the functional criteria necessary for the design and implementation of the Technical Support Center and the Emergency Operations Facllltyo Emergency facilities,needed to support an emergency response have been'rovided including a Technical Support Center, Emergency Operations Facility and Opera-tions Support Center.

Each will be activated for an Alert or higher emergency classification.

The Technical Support Center is being constructed on the upper level of the flying buttresses on the west side of the Unit 2 turbine building. It will be used as the assembly point for utility, vendor, NRC, or other personnel who would be directly involved in 'assessment of plant accident response and mitigation. It has the capability to display plant status conditions, and it is habitable to the same'egree as the Control Room.

The temporary Offsite Recov'ery Center (Emergency Operations Facility) is currently'ocated in a trailer next to the County Sheriff's office.

PG8E has committed to build a permanent EOF which will satisfy all the habitability and data display criteria (NUREG-0696).

It will be used to evaluate and coordinate emergency and reentry/recovery operations on a continuing basis by PG8E, Federal and State officials.

The Operations Support" Center'(assembly area) is located in the security building and will be the assembly point'for unassigned personnel.

It is provided with telephone facilities as well as radio communications, two evacuation kits containing portable radiological monitoring equipment, and other equipment useful in an evacuation.

3-23

We require that the above cited centers be completed in accordance with the requirements of NUREG-0696 and NUREG-0737.

III.A.2 Lon -Term Emer enc Pre aredness Position Each nucle'ar,facility shall upgrade its emer'gency plans to provide reasonable assurance that adequate protective measures can and will be taken in the event of 'a radiological'mergency.

'Specific criteria to meet this requirement is delineated in NUREG-0654 (FEMA-REP-l), "Criteria for Preparation and Evaluation of Radiological Emergency

Response

Plans and Preparation in Support of Nuclear Power Plants."

Discussion and Conclusions See Appendix B, to this supplement.

3-,2$

4 CONCLUSIONS Based on our evaluation of the application as set forth in our Safety Evaluation Report issued in October 16, 1974 and Supplement Nos.

1-13 and our evaluation as set forth in this supplement, we conclude that the operating license can be issued to allow power operations at full rated power 3338 and 3411 megawatts thermal for Units 1 and 2, respect'ively subject to license conditions which will require further Commission approval and license amendments before the

- stated condition can be removed.

In addition, actions related*to emergency preparedness matters (Section 2, III.A.l. 1) must be, completed prior to issuance of a full power license.

These actions include completion of outstanding items in the State, local and site emergency plans and conduct of a satisfactory emergency response exercise.

Moreover, other matters identified in this supplement that we require to be completed prior to the issuance of a full power license, must be completed prior to exceeding 5 percent of rated power.

We conclude that the construction of Diablo Canyon Unit has been substantially completed in accordance with the requirements of Section 50.57(a)(1) of 10 CFR Part 50, and that construction of the facility has been monitored in accordance with the inspection program of the Commission s staff.

Construction of Diablo Canyon, Unit 2 is expected to be complete late 1981.

Subsequent to the issuance of the operating license for full rated power for Diablo Canyon, Units 1 8 2 the facility may then be operated only in accordance with the Commission's regulations and the conditions of the operating license under the continuing surveillance of the Commission's staff.

We conclude that the activities authorized by the license can be conducted without endangering the health and safety of the public, and we reaffirm our conclusions as stated in our Safety Evaluation Report and its supplements.

4-1

July 31, 1980 August 1, 1980 August 1, 1980 August 26, 1980 August 30, 1980 APPENOIX A CHRONOLOGY Letter to applicant regarding interim criteria for shift staffing.

Letter to applicant forwarding "Functional Criteria for Emergency Response Facilities,"

NUREG-0696 (draft).

Letter from applicant advising that information regarding evacuation times was provided in "Evacuation Times Assessment Study for Diablo Canyon Nuclear Power Plant, April 1980" submitted April ll, 1980.

Letter to applicant transmitting request for additional information concerning emergency plan.

Letter to applicant requesting clarification of issue discussed in Emergency

Plan, Revision 2.

September 3, 1980 Letter from applicant concerning control room habitability.

September 5, 1980 September 19, 1980 September 22, 1980 October 2, 1980 October 9, 1980 October 17, 1980 October 21, 1980 October 31, 1980 Letter to applicant transmitting preliminary clarification of TMI Action Plan Requirements.

Letter -to applicant transmitting addendum to clarification letter on TMI Action Plan Requirements.

Letter from applicant transmitting information on shielding review, containment monitors, and iodine sampling.

Letter from applicant forwarding comments on "Criteria for Emergency Planning Facilities,"

NUREG-0696 (draft).

Letter from applicant regarding auxiliary feedwater system analysis.

Letter from applicant (to IE) providing information to clarify February emergency plan.

Meeting with applicant, FEMA, State and county officials to (1) discuss completion dates and review matters of the state and local emergency plans and (2) discuss location of the emergency operations facility.

Letter to applicant transmitting NUREG-0737, "Clarification of TMI Action Plan Requirements."

November 13, 1980 November 25, 1980 December 2, 1980

'ecember 2, 1980 December 8, 1980 December 9, 1980 December 16, 1980 January 26, 1981 February 3, 1981 February 6, 1981 February 6, 1981 February ll, 1981 February 13, 1981 February 18, 1981 February 20, 1981 February 26, 1981 March 4, 1981 Letter to applicant concerning final. regulations on emergency planning.

Letter from applicant transmitting "Evacuation Times Assessment for the'iablo Canyon Nuclear Power Plant,"

September 1980.

Letter from applicant forwarding information regarding its upgraded meteorological program.

I Letter fr'om applicant providing information in response to requirements specified in Supplement No.

10 to SER.

Letter from applicant providing information on emergency operating procedure's.

Letter to applicant transmitting NUREG-0654/FEMA-REP 1 on emergency plans.

Letter from applicant transmitting proposed licenses to permit fuel loading and operation at 5X.

Letter from applicant transmitting information on full power license requirements.

Letter to applicant providing correction to 12/22 ltr on

'UREG"0612.

4 Letter from applicant forwarding "Special Low Power Tests:

Final Safety Evaluation Report."

Letter from applicant forwarding responses to "Instrumentation for Detection of Inadequate Core Cooling" of NUREG-0737 (proprietary and nonproprietary versions).

Letter from applicant transmitting response to "Reactor Coolant System Vents" of NUREG-0737.

Letter from applicant providing various information in response to staff request on TMI-related matters.

Letter to applicant concerning post-TMI requirements for the Emergency Operations Facility.

Letter from applicant transmitting "Diablo Canyon Unit 1, Radiation Shielding Review."

Letter from applicant providing information which responds to specific action plan requirements.

I'ssuance of Supplement No.

12 to Safety Evaluation Report.

A-2

March 5, 1981 Letter to applicant transmitting "Functional Criteria for Emergency Res pons e Faci 1 ities, " NUREG-0696.

March 12, 1981 Letter from applicant concerning radiological plans of state and local entities.

March 13; 1981 March 13, 1981 March 13, 1981 March 19, 1981 March 19, 1981 March 26, 1981 Letter from applicant concerning Westinghouse review. of cer ta in erne rgen cy procedures.

Letter from applicant providing information. in response to request by staff on TMI-related matters.

Letter from applicant concerning evaluation of potential complicating factors for emergency response planning due to effects of earthquakes.

Letter from applicant providing additional information in regard to Items I.D.l, I.'D.2 and II.B.2 of NUREG-0737.

Letter from applicant providing additional information in regard to Item II.F.2 of NUREG-0737.

Letter from applicant providing additional information on Item I.C.7 of NUREG-0737.

March 27, 1981 Letter from applicant providing additional information on meteorology in regard to emergency preparedness.

A-3

l

APPENDIX B EMERGENCY PREPAREDNESS EVALUATION BY THE DIVISION OF EMERGENCY PREPAREDNESS OFFICE OF INSPECTION AND ENFORCEMENT U.

S.

NUCLEAR REGULATORY COMMISSION IN THE MATTER OF DIABLO CANYON POWER PLANT UNITS 1 8( 2 DOCKET NUMBERS 50-275, 323

C I

V

'I

,I 1

4 I

L

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'k

INTRODUCTION The Pacific Gas and Electric Company (hereinafter referred to as the Applicant, The Company, or PG8E) filed with the Nuclear Regulatory Commission a revision to the Diablo Canyon Power Plant Units l and 2 Emergency Plan dated February

1980, as amended (hereinafter referred to as the Plan).

The Commission's staff conducted a review of this Plan.

The'staff s review also included a site visit to the facility and a public meeting during the week of November 27, 1979.

The Plan was reviewed against the criteria of the Planning Standards in Part II of the "Criteria for Preparation and Evaluation of Radiological Emergency

Response

Plans and Preparedness in Support of Nuclear Power Plants,"

NUREG-0654, Rev.

1, November 1980.

In addition, the staff has requested all California nuclear plant licensees and applicants to provide analyses on the effects of earthquake on their emergency plans, specifically in terms of the utilities capabilities to insure availability of personnel and equipment to the sites.

(Letter, R.

Tedesco to M. Furbush, dated December 16, 1980.)

This Emergency Preparedness Evaluation Report lists each standard in order, followed by a summary of applicable portions of the Emergency Plan as they apply to the Standard.

The final section of this report provides our review results and conclusions.

t At a later date a supplement will be issued setting forth the findings and determinations of the Federal Emergency Management Agency (FEMA) as to whether State and local emergency response plans are adequate and capable of being implemented.

EVALUATION A.

Assi nment of Res onsibilit Or anization Control P

annwfn tandard Primary responsibilities for emergency response by the nuclear facil'ity

licensee, State and local organizations within the Emergency Planning Zones have been assigned, the emergency responsibilities of the various supporting organizations have been specifically established, and each principal response organization has staff to respond and to augment its initial response on a continuous basis.

A licant's Emer enc Plan Evaluation The shift foreman is initially designated as the Site Emergency Coordinator.

Mhen an abnormal condition arises it is his responsibility to determine if the abnormality meets any of the emergency classifications specified in the Plan and to implement the Plan, if necessary.

There is continuous (24-hour) communication capability between the Plant and Federal,

State, and local response organizations to ensure rapid transmittal of accurate notification information and emergency assessment data.

Responsibility for the overall direction of the on-site emergency response organization is vested in the Site Emergency Coordinator.

gualified B-l

members of the station staff who report directly to him have been assigned specific responsibilities for the major elements of emergency response.

Written agreements have been executed with agencies and organizations which will provide support.

These include the State of California Depart-ment of Public Health, State of California Department of Forestry, San Luis Obispo County Department of Administrative Management, U.S; Coast

Guard, French Hospital, St. Francis Hospital, San Luis Ambulance=Service and the Westinghouse Electric Corporation.

PG8E will periodically review these agreements with all the parties involved.

B.

Onsite Emer enc Or anization P

annwfn Standard On-shift responsibilities for emergency response are unambiguously defined, adequate staffing to provide initial facility accident response in key functional areas is maintained at all times, timely augmentation of response capabilities is available, and the interfaces among various onsite response activities and offsite support and response activities are specified.

A licant's Emer enc Plan Evaluation The Shift Foreman on duty is designated as the Site Emergency, Coordinator until relieved by the most senior member of station management.

The authorities and responsibilities of the Site Emergency Coordinator have been clearly specified, including those that cannot be delegated.

The Site Emergency Coordinator can immediately and unilaterally declare an emergency and make the necessary notifications and recommendations to the authorities responsible for implementing offsite emergency measures.

Station staff emergency assignments have been made and the relationship between the emergency organization and normal staff complement is shown in the Plan.

Positions or titles of shift and plant staff personnel, both on and offsite, who are assigned major emergency functional duties are listed.

The proposed shift staffing for one-and two-unit operations satisfy the requirements in Table B-1 of NUREG-0654 for nuclear power plant emergencies.

While PG&E has not committed to provide additional offshift personnel within'0 minutes of an accident because of the remote location of the site, it has committed to provide more than 26 additional people within 60 minutes.

The duties and responsibilities of corporate management personnel who will augment the plant staff have been established; a long-term General Office Emergency Organization framework is in place.

Interfaces between and among the company corporate staff,.station staff, governmental and private sector organizations and technical contractor groups have been specified along with services to be provided.

C.

Emer enc Res onse Su ort and Resources P annwfn tandard Arrangements for requesting and effectively using assistance resources have been

made, arrangements to accommodate State and local staff at the B-2

licensee's Emergency Operations Facility have been

made, and that organiza-tions capable of augmenting the planned response have been identified.

A licant's Emer enc Plan Evaluation

- Arrangements for requesting and using outside resources have been

made, including authority to request implementation of the Federal Radiological Monitoring and Assessment Plan by either a representative of PG8E's

'Department of Nuclear Generation, or the Site Emergency Coordinator.

Also, assistance is available from the reactor vendor (Westinghouse Electric Corporation), California Polytechnic Institute Laboratory, LFE Environmental Analysis Laboratory, and PGKE's own Research Laboratory.

The Offsite Recovery Center (Emergency Operations Facility) will be activated for the more serious emergency classifications having or potent-ially having environmental consequences (Alert, Site Emergency, General Emergency).

The facility can accommodate representatives from Federal, State and local government agencies, as well as representatives from contractor and other support groups.

It will be the central point for providing information needed by primary response agencies for implementation of offsite protective actions.

D.

Emer enc Classification S stem annwfn Standard A standard emergency classification and action level scheme is in use by the nuclear facility licensee, including facility system and effluent parameters; State and local response organizations will rely on informa-tion provided by the licensee for determinations of minimum initial offsite response measures.

A licant's Emer enc Plan Evaluation The applicant has established four standard emergency classes

- Notifica-tion of Unusual Event, Alert, Site Emergency and General Emergency.

The initiating conditions used for recognizing and declaring an emergency class are based on specific measurable parameters or observable conditions

.defined as Emergency Action Levels (EAL).

The applicant has incorporated appropriate initiating conditions as set forth in NUREG-0654 for each class of emergency.

The California, State and San Luis Obispo plans are still under revision.

Consequently, we are unable to determine if the State and local emergency classification scheme is consistent with that of the applicant and are awaiting the FEMA findings.

E.

Notification Methods and Procedures P annwfn tandard Procedures have been established for notification of State and local response organizations and for notification of emergency personnel by all response organizations; the content of initial and followup messages to response organizations and the public has been established; and means to provide early warning and clear instruction to the populace within the plume exposure pathway Emergency Planning Zone have been established.

8-3

A licant's Emer 'enc Plan Evaluation Procedures have.been established for notification of State and local response organizations in case of emergency.

The Site Emergency Coordinator has been given the sole authority and responsibility to initiate prompt notification to these agencies.

Initial notification is made to the County Sheriff s office, which will in turn notify other county authorities.

The California State Office of Emergency Services (OES) is also notified, and it will in turn alert other State agencies.

The information to be reported to the offsite agencies in the event of an emergency has been predetermined in accordance with the recommendations in NUREG-0654 and is included as part of the Plan and implementation pro-cedures.

A form, which contains preworded messages and blanks, is available for use in initial notification of offsite authorities.

The message contents include the emergency classification, brief description of the incident, radiological release information, meteorological data and recommended actions.

The applicant is installing a prompt aler'ting and noti,fication system in accordance with Appendix 3 of NUREG-0654, within a 10-mile minimum radius of the plant.

The system consists of about 52 electro-mechanical sirens and is capable of notifying 100K of the population within 5'iles, and 90K of the population between 5 and 10 miles of the plant, within 15 minutes after notification to the County Sheriff. 'ctuation of the'system is the decision of the County OES.

In the pre-accident public information

program, the populace will be instructed that the sirens are simply alerting devices and that people should turn on radios to predesignated stations for further instructions.

Installation of the system will be completed by July 1, 1981, in accordance with 10 CFR 50, Appendix E.

A pre-accident public information program will be implemented.

This

'program aims to provide the resident and transient populations within the 10-mile EPZ with information on topics such as emergency classes and protective measures.

F.

"Emer enc Communications annwfn tandard" Provisions exist for prompt communications among principal response organizations, to emergency personnel and to the public.

A licant's Emer enc Plan Evaluation The station communication system is designed to provide secure,'edundant and diverse communications to all essential onsite and offsite locations during normal and accident conditions.

Onsite systems are comprised of an intercom system, UHF and VHF two-way radio systems, and a direct dial telephone system.

Offsite systems are comprised of both commercial and leased telephone lines, UHF and VHF two-way radio systems, and a direct dial telephone system which'rovides communication to all PG8E facilities.

Two separate commercial telephone lines are dedicated to NRC communications.

These telephones plus other systems are located in plant areas manned 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day.

The Emergency Coordinator will, in emergency situations, communicate directly with the County Sheriff's Office, the State OES and the NRC duty officer.

These offices are manned 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day.

Communi-cations among the Control Room, Technical Support Center, Operations Support Center and Offsite Recovery Center, are available via the above mentioned onsite and offsite communications systems.

Some systems (such as the commercial telephone lines) are continually used and are therefore kept in constant working order; systems that are not used continually will be tested periodically.

G.

Public Information Plann)n Standard Information is available to the public on a periodic basis on how they will be notified and what their initial actions should be during an emergency; the principal points of contact with the news media for dis-semination of information (including physical location) are established in advance; and procedures for coordinated dissemination of information to the public are established.

A licant's Emer enc Plan Evaluation The applicant's public information program will consist of general infor-mation on warning procedures and protective actions.

This information will be provided to the public in various forms such as pamphlets, adver-tisements, or bill inserts such that all topics areas will be covered annually.

In addition, the applicant will, in cooperation with State and local agencies, provide such information in periodic public meetings and via radio and television announcements.

In an emergency, the Offsite Recovery Center will serve as the principal point of interaction between the station, governmental authorities and corporate management for exchange of information.

The Public Information Recovery Manager, who will be stationed in the Offsite Recovery Center, will coordinate the dissemination of information with the PG8E Corporate Headquarters.

All information released to the news media will be approved by the Recovery Manager in charge of the Offsite Recovery Center activities.

Me will evaluate the applicants'Ublic information program in a supplement to this report.

The applicant will conduct an annual information program in seminar format to acquaint representatives of the news media and the general public with the emergency plans, information concerning radiation and points of contact for release of information in an emergency.

H.

Emer enc Facilities and E ui ment ann)n tan ar Adequate emergency facilities and equipment to support the emergency response are provided and maintained.

B-5

A licant's Emer enc Plan Evaluation Emergency facilities needed to support an emergency response have been provided including a Technical Support Center, Emergency Operations Facility and Operations Support Center.

Each will be activated for an Alert or higher emergency classification.

The Technical Support Center is being constructed on the upper level of the flying buttresses on the west side of the Unit 2 turbine building.

It will be used as the assembly point for utility, vendor, NRC, or other personnel who would be directly involved in assessment of plant accident response and mitigation. It has the capability to display plant status conditions, and it is habitable to the same degree as the Control Room.

E The temporary Offsite Recovery Center (Emergency Operations Facility) is currently located in a trailer next to the County Sheriff's office.

PG8E has committed to build a permanent EOF which will satisfy all the habitabil-ity and data display criteria (NUREG-0696).

It will be used to evaluate

'nd coordinate emergency and re-entry/recovery operations on a continuing basis by PG8E, Federal and State officials.

The Operations Support Center (assembly area) is located in -the security building and will be the assembly point for unassigned personnel.

It is provided with telephone facilities as well as radio communications, two evacuation kits containing portable radiological monitoring equipment, and other equipment useful in an evacuation.

The Plan provides a listing of the emergency equipment stored at various strategic locations around the, facility.

Stored emergency equipment will be inventoried and surveyed periodically.

Equipment resources are provided to replace those that may be removed for servicing and calibration.

Onsite monitoring systems and instrumentation used to initiate emergency measures or provide continuing assessment have been identified.

These include meteorological and seismic instrumentation, radiological monitors, process monitors, fire detection

systems, and portable dose rate and radiation detection instruments.

The meteorology program has been reviewed and we find that the methods, systems and equipment described by the applicant satisfactorily meets

'milestones j., 2, and 3 of NUREG-0737, III.A.2 and NUREG-0654 Appendix 2, Revision 1 criteria for the issuance of a full power license.

The applicant has made provisions for offsite monitoring including an extensive thermoluminescence dosimeter (TLD) network, real time radiation monitors and portable radiation monitoring instruments for use by the offsite field assessment teams.

I.

Accident Assessment Plannsn Standard Adequate

methods, systems and equipment for assessing and monitoring actual or potential offsite consequences of a radiological emergency condition are in use.

B-6

A licant's Emer enc Plan Evaluation The applicant has identified the instruments that will be used to identify and assess an accident at the Diablo Canyon Power Plant.

In addition, the applicant has provided methods to project actual or potential offsite consequences using plant parameters, onsite meteorologicial conditions, and process radiological monitor indications.

The portion of the DCH relating to transport and diffusion of gaseous effluents in consistent with characteristics of the Class A model as required by NUREG-0737, Item III.A.2, Milestone 3 (ii).

A high range containment monitor capable of detecting 10~ R/hr is available for assessing the gross activity within containment.

This information, together with the predetermined activity levels resulting from various nuclide releases from the coolant and the fuel, will aid the operators in assessing the status and extent of core degradation in the event of a serious accident.

In addition to having the capability to project offsite consequences from measured in-plant parameters, the applicant has also established a field

-'onitoring capability.

Field monitoring teams will be employed whenever there are potential offsite radiological consequences.

These teams will have use of portable radiation monitors and air samplers, as well as a

well-e'quipped mobile emergency response laboratory.

The,air samplers used to monitor radioiodine can detect a level as low as 10-e uCi/cc.

J.

Protective Res onse annwfn tan ar A range of protective actions has been developed for the plume exposure pathway for emergency workers and the public.

Guidelines for the choice of protective actions during an emergency, consistent with Federal

guidance, are developed and in place, and protective actions for the ingestion exposure pathway appropriate to the locale have been developed.

A licant's Emer enc Plan Evaluation The applicant has established onsite protective responses for employees,*

contractor personnel, and members of the general public who may be within the exclusion area at the time of an emergency.

These responses consist of warning and notification, relocation and accountability, and protective

actions, Onsite warning and notification will be by means of various alarm systems, station public address

system, or by members of the security force depending on the location of the individuals within the exclusion area.

In the case of a Site or General Emergency, personnel within the protected area will be relocated and an initial accountability completed within thirty minutes.

The Site Emergency Coordinator will authorize the site evacuation.

Evacuation can take place on the plant access road in the southerly direction or on an alternate route in the northerly direction if radiation levels prevent use of the former. 'dditi,onal onsite protective measures include the use of individual respiratory protection, protective

clothing, and radioprotective drugs.

The Plan provides criteria for recommending offsite protective measures depending on the projected doses to the populace.

The particular recom-mendation may be sheltering or evacuation depending on the magnitude of B-7

the projected

dose, the meteorological conditions, the nature of the

release, and the consideration of evacuation time estimates for the affected sector(s).

K.

Radiolo ical Ex osure Control P ann)n Standards Means for controlling radiological exposure are established for emergency workers.

The means for controlling radiological exposures shall include exposure guidelines consistent with EPA Emergency Worker and Lifesaving Activity Protective Action Guides.

A licant's Emer enc'lan Evaluation The applicant has established a radiation protection program for control-,

ling radiological exposures in the event of an emergency.

Emergency exposure guidelines have been provided for the various categories of radiation workers.

These guidelines are consistent with the EPA Emergency Worker and Life Saving Activity Protective Action Guides.

The Plan clearly specifies that only the Site Emergency Coordinator is authorized to permit emergency exposures in excess of 10 CFR Part 20 limits.

The applicant has established 24-hour-a-day dose determination capability for emergency personnel.

Dose records will be maintained to ensure that the exposure history is current.

Onsite contamination control measures for personnel, equipment, and access control are provided.

The criteria for decontamination of personnel and equipment are specified in the Plan, together with the criteria for permitting return of areas and items to normal use.

Provisions have been established for decontaminating onsite personnel including provisions for extra clothing and decontaminants suitable for the type of contamination expected.

Reserve supplies of clothing and decontaminants are stored onsite.

L.

Medical and Public Health Su ort lannsn Standard Arrangements are made for medical services for contaminated injured individuals.

A licant's Emer enc Plan Evaluation PG8E has made arrangements with the French Hospital in San Luis Obispo, and St.

Francis Hospital in San Francisco to provide medical assistance to site personnel injured in accidents involving radioactive contamina-tion.

The hospitals have designated space for the care and treatment of contaminated patients.

The plant has a first aid facility located in the Auxiliary Building for providing medical assistance to all site personnel.

The facility can provide first aid treatment for minor injuries and emergency aid for more serious injuries.

An agreement has been made with a physician to serve 8-8

as chief consultant on nuclear medicine during normal operation and emergencies.

A written agreement has been made with the San Luis Ambulance Service for transporting injured contaminated personnel.

M.

Recover and Reentr Plannin and Postaccident 0 erations Plannsn Standard General plans for recovery and reentry are developed.

A lica'nt's Emer enc Plan Evaluation The PG8E Corporate Emergency

Response

Plan is intended to support the Diablo Canyon Power Plant in the execution of its emergency plan.

The corporate organization will consist of experienced company management headed by the Senior Vice President, Operations, and supervisory personnel who. have the authority to assure the best available u'se of company resources to assist in rap'id recovery.

The corporate organization will provide:

1.

2.

3.

4.

5.

Technical and operational support planning for recovery operation Radiological field monitoring and data assessment Logistics support for emerge'ncy p'ersonnel Management level interface with governmental authoritieq Release of information to news media coordinated with governmental authorities Any decision to relax protective measures will be 'made by the Emergency Coordinator using guidance from the Recovery Manager, and the senior NRC representative at the scene.

Mhenever'a recovery operation is to be initiated or any change is to be made in the licensee's organizational structure, the. State will be notified.

N.

Exercises and Dri.lls Plann)n Standard Periodic exercises wi 11 be conducted to evaluate major portions of emergency response capabilities, periodic drills will be 'conducted to develop and maintain key skills, and deficiencies identified as a

result of exercises or drills will be corrected.

'I A licant's Emer enc Plan Evaluation An exercise will be conducted, prior to issuarice of an operating license and with offsite response agencies.

Thereafter, exercises will be con-ducted annually.

Although the State plan will be exercised annually, it may be done separate from the licensee exercise in some years due to the existence of other nuclear power reactor facilities within the State' jurisdiction.

Me,require that the scenario used for the various exercises contain at least the essential elements as set forth in NUREG-0654; Arr'angements will be made for official Federal, State and local 'observers and a critique will'be held after each exercise.

PG8E will review and 8-9

resolve any identified deficiencies, and ensure that appropriate actions have been taken to correct the deficiencies.

In addi,tion.to the exercises, various drills will be conducted covering communications, fires, medical emergencies, health physics and radio-logical monitoring.

Depending on the particular drill, the frequency varies from monthly to annually in accordance with recommendations in NUREG-0654.

Minimum requirements for each of the drills will be as

'tated in NUREG-0654.

Deficiencies resulting from evaluation of the drills will be handled by station management and as discussed above for exercises.

Radiolo ical Emer enc Res onse Trainin Plannin Standard Radiological emergency response training.'is provided to those who may be called upon to assist in an emergency; I

'k A licant's Emer enc Plan=Evaluation

'I l

C The applicant will provide training on the Emergency Plan and procedures to all-permanent plant personnel.

This includes assignment of duties and responsibilities, location and use of assembly

areas, and familiarization with alarms and communications-"systems.

In addition, those personnel having specific response roles's part of the onsite:emergency organiza-tion are given-specialized training in accordance with their expected duties.

These areas include emergency response coordination,and direction, radiological monitoring, first aid and fire fighting.

K f

Training is also provided for those offsite organizations whose services may be r'equired'in an emergency, such as medical support personnel and'.

the'an Luis Obispo County officials."

Res onsibilit for the Plannin Effort:

Develo ment Periodic Review and Distr)but>on of Emer enc Plans Plannin Standard II Responsibilities for'mergency plan development, review and distribution are established and that planners are properly trained.

A licant's Emer enc Plan Evaluation The Vice President, Nuclear Power Generation, has the overall authority and responsibility for radiological emergency response planning at the corporate level.

The Supervising Nuclear Generation Engineer (Personnel and Environmental Safety) is the Emergency Planning Coordinator responsible for maintenance of the emergency plan.

Provisions exist for biennial review and revision of the emergency plan and its implementing procedures.

Me require, however, that PG&E perform this review annually, and update telephone numbers in emergency procedures quarterly.

In addition, the critiques of drills and exercises shall be used as bases for changes and revisions.

Any changes to these documents

will be provided to the organizations and individuals having a responsibi-lity for implementing the emergency. plan.

We also require that the overall emergency preparedness program be audited annually by an independent

group, as defined in NUREG-0654.

The audit will include the emergency plan and procedures, training, readiness

testing, emergency equipment, and interfaces with State, and local governments.

CONCLUSIONS ON LICENSEE EMERGENCY PLAN Based on our review of the Applicant's Emergency Plan against the criteria in NUREG-0654, Rev.

1, November 1980, titled "Criteria-for Preparation and Evaluation and Radiological Emergency

Response

Plans and Preparedness in Support of Nuclear Power Plants,"

we conclude that the Diablo Canyon emergency plan, when revised in accordance with the commitments made, provides an adequate planning basis for an acceptable state of emergency preparedness and will meet the requirements of 10 CFR 50 and Appendix E thereto.

However, the Emergency Plan must be revised to address the final criteria and implementation schedule for the emergency centers and their functions as previously discussed.

h PGLE has been requested to explicitly address protective action determination and implementation during an earthquake in the revised site plan.

In addition, FEMA has been requested as part of their review of State and local emergency plans to review the planning efforts for the areas around the site to assure that protective actions to be recommended by PG8E during earthquakes could be implemented and are adequate.

After receiving the findings and determinations made by FEMA on State and local emergency response

plans, and after reviewing the revised site plan from

PG8E, a supplement to this report will provide the staff's overall conclusions on the status of emergency preparedness for the Diablo Canyon Power Plant and related Emergency Planning Zones.

The final NRC approval of the state of emergency preparedness for the Diablo Canyon site will be made following implementation of the'mergency plans to include development of procedures, training and qualifying of personnel, installation of equipment and facilities, and a joint exercise of all the plans (site, State and local).

I II 1

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4

NRC FORM 335 I7.77)

U.S. NUCLEAR REGULATORY COMMISSION BIBLIOGRAPHICDATASHEET

1. REPORT NUMBER /Assigned by DDCJ NUREG-0675 Supplement No. Q

4. TITLE ANDSUBTITLE (Add Vo/ume Na, IfappropriaseJ Safety Evaluation Report Related to the Operation of Diablo Canyon Nuclear Power Hant, Units 1 and 2 2, (Leave blank J

3. RECIPIENT'S ACCESSION No.

7. AUTHORIS)

5. DATE REPORT COMPLETED

9. PERFORMING ORGANIZATION NAME AND MAILINGADDRESS (include Zip CodeJ Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Coamission Washington, D,C, 20555 MONTH A ril DATE REPORT ISSUED MONTH A ril

6. (Leave blank J

8. (Leave blankJ YEAR 19S1 YEAR 19S1

12. SPONSORING ORGANIZATION NAME AND MAILINGADDRESS (include Zip Coda J

10. PROJECT/TASK(WORK UNITNo.

Same as 9. above 11 CONTRACT No

13. TYPE OF REPORT PERIOD COVERED (Inclusive dalesJ

15. SUPPLEMENTARY NOTES Docket Nos, 50-275 and 50-323

16. ABSTRACT (200 words or lessJ

14. (Leave blankJ Supplement No. 34'to the Safety Evaluation Report for Pacific Gas and Electric Company' application for licenses to operate the Diablo Canyon Nuclear Power Plant (Docket Nose 50-275 and 50-323) located in San M.s Obispo County, California has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission.

Supplement No 14 delineates our evaluation of the full power Mal-related issues.

'7.

KEY WORDS AND DOCUMENT ANALYSIS 17a. DESCRIPTORS 17b. IDENTIFIERS/OPEN.ENDED TERMS

18. AVAILABILITYSTATEMENT Unlimited NRC FORM 336 I7-77) 19; SECURITY CLASS (7his,reporsJ Unclassified

20. SECURITY CLASS (This paueJ Unclassified

21. No. OF PAGES

22. PRICE S

aVS, OOVSRNMENT PRINTINO OFFICII 1981

$41-742/795 1-3