ML16342C158

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)

ML16342C158
Person / Time
Site:	Diablo Canyon
Issue date:	08/31/1985
From:	Office of Nuclear Reactor Regulation
To:
References
TASK-2.K.3.30, TASK-2.K.3.31, TASK-TM NUREG-0675, NUREG-0675-S32, NUREG-675, NUREG-675-S32, TAC-56737, TAC-57575, NUDOCS 8508210398
Download: ML16342C158 (58)

Text

NUREG-0675 Supplement No. 32 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 U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation July 1985

ABSTRACT SSER 32 has been prepared by the NRC staff with respect to the full-power li-censing of Diablo Canyon Unit 2 (Docket 50-323).

SSER 32 addresses (1) a number of items that had been identified in earlier reports as requiring specific ac-tion prior to full-power, including allegations, seismic analyses, and comple-tion of modifications; (2) combined and common Technical Specifications for Diablo Canyon Units 1 and 2; and (3) changes to license conditions as included in the Unit 2 low-power license.

Diablo Canyon SSER 32

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'ABLE OF CONTENTS ABSTRACT.

Pacae 1

INTRODUCTION.........

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SUMMARY

AND CONCLUSIONS..........

2-1 3

TURBINE BUILDING.........

3-1 3,1 3.2 3.3 Introductson......

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Scope of Staff Review..

Findings and Conclusions...........................'-1 3"2 3"3 4

PIPEMAY STRUCTURE

4. 1 Introduction.........

4.2 Scope of Review.

4"1 4"1 4.2.1 4.2.2 4.2.3

4. 2.4 Modeling of Pipeway Connection to Auxiliary and Turbine Buildings............

Selection of Seismic Input for"the Pipeway Structure Model.

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Integration Time Step..

DE and DDE Evaluation.......................

4-2 4-3 4"3 4"4 4.3 Conclusions 5

ALLEGATIONS 6

PHYSICAL SECURITY 4"5 5-1 6-1 7

EMERGENCY PREPAREDNESS...............

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7. 1 FEMA Findings on Offsite Emergency Plans and Preparedness.......

7.2 Offsite Emergency Planning for Medical Services 7.3 Conclusions.

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TECHNICAL SPECIFICATIONS~......,........

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8. 1 Introducti on...........,.....

8.2 Unit 1 Technical Specifications Considerations.

8.3 Unit 2 Technical Specifications Considerations.

8.4 Movable Control Assemblies 8.5 Reactor Trip System Instrumentation............

8. 6 GONPRAC Composition.....

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8.7 Summary and Conclusions..........

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TABLE OF CONTENTS (Continued) 9 SMALL-BREAK LOCA (II.K.3.30 and II.K.3.31)"..

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9-1 9.1 9.2 9.3 9.4 9.5 Summary.

Introduction......

Summary of Requirements.

Evaluation.

Conclusions.................................

9-1 9"1 9"2 9-2 9"4 10 OTHER MATTERS......

10-1

10. 1 Seismically Induced Systems Interactions.......'...

10.2 Regulatory Guide 1.97 - Emergency

Response

Capabil 10.3 Masonry Walls.........,..............

10.4 Generic Letter 83-28:

Required Actions Based on Generic Implications of Salem ATWS Events 10.5 Equipment qualification...............

10.5. 1 Seismic and Dynamic gualification.-......

10.5.2 Environmental qualification..

10.6 Fire Protection...

10. 7 Plant Readiness..

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10.8 Post Accident Sampling System (II'.3)..

10.9 Inadequate Core Cooling Instrumentation System (II.F. 2)

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10"1 10-1 10-2 10-3 10-4 10-4 10-4 10"4 10-4 10-5 10-5 11 REFERENCES 11-1

~II.K.3.30 and same type of identification throughout this report refers to NUREG-0737 Items.

Diablo Canyon SSER 32 vi

1 INTRODUCTION The staff of the U.S.

Nuclear Regulatory Commission (the NRC or. th'e staff) issued on October 16, 1974, its Safety Evaluation Report (SER). concerning the application by the Pacific Gas and Electric Company (PGLE) to operate the Diablo Canyon Nuclear Power Plant, Unit 1 and Unit 2.

The -report has since been supplemented with SER Supplements No.

1 through No.

31 (SSER 1'through SSER 31) regarding further staff reviews.

SSERs 1 through 17 apply equally to both units, SSERs 18 through 27 were issued primarily with respect to Unit 1, and SSERs 28 through 31 were issued with respect to the licensing of Unit 2 on the following matters:

SSER 28:

Allegations SSER 29:

Design Verification SSER 30:

Piping and Pipe Supports SSER 31:

Miscellaneous Matters

'SSER 31 presented the staff safety evaluation of matters that resulted in license conditions in the Unit 2 low-power, license, NUREG-0737'tems and other safety issues.

SSER 31 also identified those matters that required further action or resolution prior to full-power operation, including those that were addressed in more detail in other SSERs.

This is SER Supplement No.

32 (SSER 32) which provides the staff evaluation of those matters that require an appropriate resolution prior to full-power opera-tion of Unit 2 and which updates previous staff evaluations.

The NRC Project Manager for the Diablo Canyon Nuclear Power Plant is Mr.

H.

Schierling.

He may be contacted by calling (301) 492-7100 or by writing to:

Division of Licensing, U.S. Nuclear Regulatory Commission, Washington, D. C.

20555.

Copies of this supplement are available for public inspection at the Commission's Public Document Room at 1717 H Street, N.W., Washington, D. C.

and at the California Polytechnic State University Library, Document and Maps Department, San Luis Obispo, California 93407.

Availability of all material cited is described on the inside front cover of this report.

Diablo Canyon SSER 32

2

SUMMARY

AND CONCLUSIONS The staff has evaluated those matters and specific items that require resolution prior to a decision on the issuance of a full-power license for Diablo Canyon Unit 2 as previously identified in SSER 31.

The staff also completed its evalu-ation of certain other matters not directly related to the full-power licensing.

Finally, the staff and PGRE developed a common and combined set of Technical Specifications for both units.

A summary of the staff findings for each item is provided below.

1.

Turbine Buildin Section 3

The concrete floor slab at elevation 140 feet in the Unit 2 turbine building is structurally adequate and the floor spectra are acceptable.

The issue of shear friction capacity along the construction joint has been resolved.

2.

Pi ewa Structure Section 4

The modeling, input, and integration time steps in the Hosgri evaluation of the pipe way structure are appropriate.

A license condition will be included in the full-power license requir ing PGSE to perform a confirmatory analysis, prior to startup after the first refueling, to further demonstrate the adequacy of the structure with respect to load combinations that include the DE and DDE.

3.

Alle ations Section 5

Allegations received as of June 30, 1985 have been screened in accordance with previously established criteria and have been determined not to be of sufficient safety significance to preclude the continued full-power operation of Unit 1 or the issuance of a full-power license for Unit 2.

The staff evaluations and investigations are ongoing.

4.

Seismicall Induced S stems Interactions Section

10. 1

,All modifications resulting from the SISIP have been completed.

This completes the staff requirements regarding SISIP plant modifications.

5.

Re ulator Guide 1.97-Emer enc Res onse Ca abilit Section 10.2 Regulatory Guide 1.97 equipment has been installed and tested.

Two items (radioactivity concentration monitor and boric acid charging flow meter) will be installed when commercially available.

This is acceptable to the staff.

6.

Masonr Walls (Section 10.3 PG8E has completed the required masonry wall modifications and has completed a survey of construction documents.

The staff concludes there is reasonable Diablo Canyon SSER 32 2"1

assurance that the masonry walls are constructed in accordance with design requirements.

The license condition regarding the verification of the energy-balance technique remains unchanged.

7."

Generic Letter 83-28 Section

10. 4 The staff has completed its evaluation of a number of items in Generic Letter "83-28.

'The license condition for the completion remains unchanged.

8.

E ui ment ualification Section 10v5 PG8E has completed the required modifications and the documentation of qualifi-cation records.

PG8E will implement the surveillance and maintenance program prior to full-power operation.

The staff finds this acceptable.

9.

Fire Protection Section 10.6 PG8E has concluded that the modifications pertaining to fire protection have been completed.

The license condition remains unchanged.

10; Plant Readiness Section 10.7 PG8E has completed the preoperational test of the radiation monitoring system an is on schedule regarding conditions on the master completion list.

The staff concludes that the low-power license condition may be deleted from the

. full-'power license.

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Post Accident Sam lin S stem Section 10.8 The 'PASS system has been installed and is operational.

This item is resolved.

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Inade uate Core Coolin Instrumentation S stem Section 10.9 PG8E has completed the upgrade of the ICCIS for Unit 1, appropriate Tech Specs have been developed, and ICC procedures have been implemented.

Final calibration of the Unit 2 ICCIS will take place during power ascension.

The staff finds this acceptable.

13.

'Ph sical Securit Section 6

The low-power license condition will be included in the full-power license.

There's no change.

14.

Emer enc Pre aredness Section 7

PG&E has committed to fully comply with the Commission's response to the remand by the U.S.

Court of Appeals regarding the recent decision, GUARD v.

NRC, pertaining to medical services.

FERA has confirmed that a list of medical service facilities is contained in the emergency plans of involved jurisdictions.

This is acceptable to the staff.

Diablo Canyon SSER 32 2-2

15.

Technical S ecifications Section 8

PG8E and the staff have developed a

common and combined set of Technical Specifi-cations.that will apply to both units.

Unit-specific information is identified.

These specifications will be issued for Unit 2 with the issuance of the full-power license and for Unit 1 by amendment to the Unit 1 full-power license.

PG&E has committed to submit, within 90 days after issuance of the Unit 2 full-power li-

cense, an amendment request on each unit for five specifications that were not included at this time.

The staff finds this acceptable.

The staff has requested PG8E to include specifications for the polar crane inside containment and for

'the two bridge cranes in the turbine building or provide justification for ex-cluding them.

16.

Small Break LOCA Section 9

The staff has completed its evaluation of the Westinghouse generic small break LOCA analysis regarding the requirements of NUREG-0737 Item II.3.30.

PG8E has committed to provide the Diablo Canyon plant specific analysis (i.e.,

Item II.K.3. 31) within 1 year.

The staff finds this acceptable.

17.

License Conditions The Unit 2 full-power license will include the same license conditions as the low.-power license except as follows:

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Attachment 1 to license condition 2. C.(1) will be deleted; Attachment 1

required performance of the preoperational test of the radiation monitoring system and completion of work items (see Section 10).

2.

A new license condition,

3. C.(10), will be added requiring a confirmatory analysis of the pipeway structure to demonstrate the adequacy with respect to DE and DDE load combinations (see Section 4).

In its decision regarding Unit 2 design verification, ALAB-811, the Appeal Board required the imposition of two conditions:

one requiring jet impingement analy-ses of certain lines inside containment and the second requiring a technical specification regarding component cooling water operability.

The jet impinge-ment analyses required by the first condition has been satisfied.

As discussed in Section 10.2 of SSER 29, dated March 1985, PG8E has performed the required analyses which have been reviewed and found acceptable by the staff.

With re-spect to the second condition, the technical specifications incorporated as part of the Unit 2 operating license includes the same condition on component cooling water operability as was previously imposed in the Unit 1 technical specifications.

Diablo Canyon SSER 32 2-3

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3 TURBINE BUILDING

3. 1 Introduction The staff presented its safety evaluation of the seismic design of the Unit 2 turbine building in SSER 29 (Ref. 3).

As stated in Section

5. 2. 2 of the report PG&E had not yet completed at that time its evaluation of the shear friction capacity along a construction joint in the concrete floor slab at elevation 140.

The staff concluded in Section

5. 3. 2 of the report that this issue did not impact low-power operation but must be resolved prior to full-power operation.

The staff audited the PG8E analysis in January and Nay 1985 at,.the PG8E offices

'n San Francisco and in April 1985 at the Brookhaven National Laboratories (BNL).

PGSE submitted a final report on the subject by letter dated June 6,

1985 (Ref. 7).

The floor slab at elevation 140 in the turbine building is a 12-inch thick re-inforced concrete slab with a large central cutout for the turbine pedestal and an equipment elevator cutout at column lines C and 19.

The slab is supported on a grillage of steel beams.

The diaphragm action of the slab provides torsional stiffness to the turbine building at elevation 140.

The diaphragm acts as a one bay Vierendel truss with the E-W shear walls at column lines 19 and 35 providing the primary "support.

A critical section of the diaphragm (for in-plane 'forces) occurs along, the N-S column line C between the E-W lines 19 and,21 because of the cutout for the equipment elevator shaft at the north end of the slab at column lines C and 19.

It should be noted that the condition is not the same for the Unit 1 turbine building because of the extra bays in the Unit 1 turbine building and a smaller equipment elevator shaft cutout in the elevation 140 floor slab.

During the PGLE stress evaluation of the turbine building for the Hosgri event, the shear stresses in this critical region were found to slightly exceed the criteria in ACI 318-73, Section 11.16.

The loads used for this stress evalua-tion were developed from the three-dimensional model of the entire turbine building.

The floor diaphragm was modeled with relatively large, plane strain, elastic elements in this analysis.

Concrete cracking was not considered when the diaphragm loads were determine'd.

PG8E prepared a more detailed model of the floor at elevation 140 to obtain a better estimate of the loads.

A finer finite element grid was used to model the concrete floor slab, and separate elements were used to model the embedded steel beams and reinforcing steel.

The nonlinear computer code FINEL was used so that the cracking of the concrete along planes of principal tensile stresses could be considered.

Peak floor nodal accelerations were obtained from the three dimensional dynamic model of the turbine building.

Diablo Canyon SSER 32 3"1

These accelerations were used by the computer program FINEL to compute nodal forces.

These accelerations were adjusted to produce forces in the uncracked FINEL model at the floor slab critical location that matched the peak forces

.computed in the three dimensional model.

These adjusted accelerations were also used as input to the cracked floor slab FINEL model so that the effect of concrete cracking on the magnitude of the loads on the critical section could be evaluated.

The forces resulting.from the FINEL model, when concrete cracking was con-

sidered, were smaller than the forces determined in the uncracked model.

The stress limits of the concrete code ACI 318-73, Section

11. 16 were satisfied using the FINEL cracked concrete model.

The floor response spectra used in the equipment evaluation was based on the-uncracked concrete floor slab.

The FINEL solution indicates that some concrete cracking occurs, but this cracking is restricted to a small region of the slab.

Since the cracking is localized, it is not expected that the overall stiffness of the slab changes enough to cause significant changes in the floor response spectra.

3.2 Sco e of Staff Review The staff has reviewed the three-dimensional model of the turbine building (PG&E Calculation 64T283) and found it to be acceptable as reported in SSER 29.

The shear problem in the elevation 140 floor slab was first discussed with PG&E during the audit on January 15-17, 1985.

At that time very preliminary calculations were available indicating that a problem might exist.

PG&E presented a plan of action that would address this matter in detail.

Progress on the PG&E.work was reviewed at an audit held at Brookhaven National Labora-tory on April 16, 1985.

Mhile the PG&E calculations were not complete at that time, the preliminary assessment by PG&E was that the problem could be satis-factorily resolved.

A detailed audit of the PG&E calculations was conducted on May 30-31, 1985, at the PG&E offices in San Francisco.

The following items were reviewed during this audit:

(a)

'PG&E Calculation 65T355 The evaluation of shear stresses at the critical section was made in this calculation.

The shear, tensile force, and moment at the critical section based on the original three dimensional model are:

2780 kips, 2940 kips and 10,400 kips-feet respectively.

Using the ACI 318-73 allowable shear for a concrete, strength of 6590 psi (the average of the 60-day test results) and a capacity reduction factor, g, of 0.85 the actual shear capacity was shown to be 2370 kips.

This capacity is less than the demand

of 2780 kips.

The shear, tensile force and moment at the critical sections, based on the FINEL model including concrete cracking is:

1850 kips, 1960 kips, and 3700 kips-feet respectively.

The shear capacity, using the 60-day strength and a $ factor of 0.85, is 2370 kips.

The shear capacity is larger than the demand of 1850 kips.

The moment and tensile capacity is larger than the demand.

Diablo Canyon SSER 32 3-2

(b)

PG&E Calculation 65T318 The FINEL model was developed in this calculation.

From the results of these calculations, it is shown that the concrete cracking develops across the critical section and the cracks generally run in the NM-SE direction.

Stresses in the cracked elements indicate that a tensile stress of about 600 psi (close to the modulus of rupture of the concrete) was the criter-ion used to decide when cracking occurred.

This indicates that at least for the problem of interest FINEL gave reasonable results.

(c)

FINEL Manual The FINEL Manual was reviewed to assess the adequacy of the FINEL Code to handle the problem of interest.

The Code employs an iterative solution starting with the uncracked solutions.

After each iteration all elements are reviewed to determine whether or not cracking occurs and if so, at what angle.

The stiffness matrix is modified for the cracked element and the next iteration is performed.

Convergence is determined when the changes in the solution between one iteration and the next are acceptably small.

Six iterations were used for the turbine building analysis.

The Code has been verified by Bechtel using comparisons with both analyt-ical and experimental data.

The comparisons for a cantilever beam and a

deep panel were particularly applicable to the problem at hand.

3.3 Findin s and Conclusions Based on the staff audits and reviews of the extensive information provided during the audits by PG8E as described

above, the staff findings are as follows:

(a)

The shear demand based on the elastic model loads are larger than the ACI capacity based on 60-day concrete strength and a g factor of 0.85.

(b)

(c)

The FINEL Code is applicable to the turbine building floor slab and the results for member loads are realistic.

When the more realistic model is

used, including the effects of cracking, the shear loads are reduced, resulting in the demand being less than the capacity.

Since cracking is restricted to a small region of the slab, one would not expect the overall =stiffness of the slab to change enough to cause sig-nificant changes in the original floor response spectra.

Based on the FINEL results which take into consideration load redistribution caused by concrete cracking, the staff concludes that the elevation 140 turbine building floor slab is structurally adequate and the floor response spectra are acceptable.

The staff concludes that the issue of shear friction capacity at the construction joint in the concrete slab at elevation 140 has been resolved.

Diablo Canyon SSER 32 3-3

4 PIPEWAY STRUCTURE

4. 1 Introduction The staff presented its safety evaluation of the seismic design of the Unit 2 pipeway structure in Section 8 of SSER 29 (March 1985).

As stated in Sec-tion 8.3 of the report the staff had not completed its evaluation and was awaiting further information from PG8E regarding the design earthquake (DE) and double design earthquake (DDE).

The staff concluded that the issue need not be resolved for low-power operation but must be resolved for full-power

'peration.

Each unit of the Diablo Canyon-plant has a steel space frame pipeway structure attached to the outside of the containment shell, turbine building, and auxi 1-iary building between elevations 87 feet and 119 feet.

The main function of the pipeway structure is to support the main-steam and feedwater lines from their point of exit from the 'containment to their entry into the auxiliary building.

The seismic Category I pipeway structure is required to satisfy load combina-tions involving dead,

seismic, and pipe rupture restraint loads.

The seismic loads include the design earthquake (DE), the double design earthquake (DDE),

and the Hosgri event.

The seismic analysis for the Hosgri event for the Unit 1 pipeway was performed by Westinghouse, the Unit 2 pipeway Hosgri analysis was performed by PG&E.

The DE and DDE evaluations for both units were performed by PG&E.

The pipeway analysis for the pipe rupture loading condition was per-formed by PG8E for both units of the plant.

The staff performed three audits of the civil/structural calculations for the pipeway structure:

January 16 - 17, February 28, and May 30 - 31, 1985.

-The first and the third audit took place in'an Francisco, California at the PG&E offices, the second audit was conducted in Monroeville, Pennsylvania at the Westinghouse offices.

PG&E supplied follow-up information as requested by the staff (Ref.

8 and Ref.

9) and submitteg a final report, including documents which previously had been reviewed during the audits on the subject (Ref. 10).

The staff evaluation of the pipeway structure was reported in March 1985 in SSER 29 (Ref. 3).

The concerns identified in SSER 29 are summarized below:

(a) modeling of the pipeway structural connections to the auxiliary and turbine buildings, (b) selection of the seismic input for the pipeway seismic analysis, (c) integration time step used in the seismic analysis for the Hosgri event, (d) procedure used to account for accidental

torsion, Diablo Canyon SSER 32 4"1

(e) strength capability of the pipeway structure under action of relative motion between containment-auxiliary-turbine buildings, and (f)

DE and DDE evaluations.

Items (d) and (e) were resolved in SSER 29, Items a, b, c, and f are discussed below.

4.2. 1 Modeling of Pipeway Connection to Auxiliary and Turbine Buildings The staff audited the PG&E calculations of the Unit 2 pipeway structure for the Hosgri event (Calculation File No.

52. 10.2).

The three dimensional dynamic model used in these evaluations incorporates a centilever beam type idealization of the containment exterior shell.

The pipeway structure is represented by an assembly of beam elements which represent structural steel members as well as, the main steam and feedwater piping.

As stated in SSER 29, the staff concluded that the connections of the structure to the auxiliary and turbine buildings are allowed to move freely in the horizontal plane.

These beams are connected to the turbine building and auxiliary building in the vertical direction only.

Since slotted holes were provided to accommodate the structural displacements in the axial direction only of the steel beams that frame into the auxiliary building and the turbine building, the in-plane freedom of the corresponding nodes was questionable.

The staff requested PG&E to justify the consistency between the modeling of these connections in the three dimensional pipeway model and the as-built condition at the plant.

During the audits PG&E provided the details of the slotted holes in both the auxiliary,and turbine buildings (Drawing No. 443375, SKC-PW-01, Sheet No.

1 to 4).

PG&E stated that the differential displacements between the pipeway struc-ture and the turbine building are a 0. 17 and t 0. 24 inches in the N-S and E-W directions, respectively.

These displacements can be accommodated by the clear-ance provided by the slotted holes.

Therefore the staff finds the turbine build-ing connection modeling is acceptable.

The pertinent differential displacements at the pipeway frame and the auxiliary building connections were not provided during the January 1985 audit.

The staff requested in SSER 29 that these dis-placements be provided to assure that there is sufficient clearance to accom-modate these movements (SSER 29, p. 8-2, Item a).

During the May 1985 audit, the pipeway structure to auxiliary building connec-tion wa's further evaluated by the staff.

The audit concentrated on the modeling assumptions used in the three dimensional pipeway structural model for a set of three beams which connect the pipeway structure with the auxiliary building.

PG&E stated that the as-built condition at the south end connection of these beams use clip angles attached to the beam web and thus the motion of the pipe-way structure is not restrained.

Thus the relative displacements between the pipeway structure and the auxiliary building can be accommodated as per as-built condition.

This condition, however, is not reflected in the three dimensional pipeway structure model.

In the model these connections are modeled as the mo-ment type.

PG&E stated that this difference is due to a modeling necessity in order to avoid instability in the computer program.

Furthermore, these members do not provide structural support for the pipeway structure.

The modeling ap-proximation should not affect the overall results of the pipeway structure seis-mic evaluation performed for the Hosgri event.

Therefore, the staff considers the issue of modeling of the pipeway connection to the auxiliary building resolved.

Diablo Canyon SSER 32 4-2

4.2.2 Selection of Seismic Input for the Pipeway Structure Model The pipeway structure of each unit is attached to the auxiliary and turbine building as well as to the containment shell.

Thus, the seismic input associ-ated with the pipeway seismic model is not the same at the various attachment points.

Both PG&E and Westinghouse employed a single input motion in their Hosgri evaluation of the pipeway structure of Unit 2 and Unit 1, respectively.

This motion was taken to be the Hosgri input at the base of the containment structure.

The staff requested PG&E to justify the selection of the input motion.

PG&E provided connection details between the auxiliary building and pipeway struc-ture which showed that no input motion can be transmitted between these two

'tructures.

Furthermore, for the same

reason, no horizontal input is expected from the turbine building to the pipeway structure.

To justify the use of the vertical input from the containment structure a response spectra comparison was made by PG&E which showed that the containment spectra enveloped the corres-ponding turbine building spectra at the locations of interest.

Thus, in th'is

case, the input choice is conservative.

The staff requested additional justification of possible transmission of seismic loads from the auxiliary building through the piping systems (SSER 29, p. 8-3, Item b).

According to the methodology used by both PG&E and Westinghouse, a

single input motion was applied to the three dimensional pipeway structure models.

These are coupled models of the pipeway structure with the major piping systems.

Since support conditions were assigned at the nodes of the piping systems in the auxiliary building, input motions are applied at these locations.

Thus, the staff questioned the adequacy of the assumption that the piping systems (main steam and feedwater lines) incorporated into the three dimensional model were excited at the auxiliary building snubber locations with the containment Hosgri response spectra.

The above concern applies to both units and was discussed dur'ing the audit '

in the Westinghouse office in Monroeville, Pennsylvania on February 28, 1985.

A local frequency was calculated by the staff using the main steam line cross-sectional properties and the distance between the attachment points i.e.,

snubber locations at the auxiliary building and the pipeway structure.

It was shown that the natural frequency of,this, segment of the main steam line is approxi-mately 3 Hertz.

At this frequency the containment spectra envelop the corres-ponding auxiliary building spectra.

Thus, the use of containment input spectra't the snubber locations in the auxiliary building is conservative in this case.

Based on the above, the staff concludes that the input transmitted through the piping systems into the pipeway structure would not affect the results of the seismic analyses for the Hosgri event.

This conclusion applies to both units.

4. 2. 3 Integration Time Step The integration time step used by both Westinghouse and PG&E in their seismic evaluation of the pipeway structures for Unit 1 and Unit 2, respectively, was 0.01 second.

Since response computations are generally sensitive to the choice of the integration time interval, the staff requested PG&E to justify the time step used in the evaluations (SSER 29, p. 8-3, Item c).

Diablo Canyon SSER 32 4-3

Westinghouse used the computer code WECAN for the Hosgri evaluation of the Unit 1 pipeway structure.

For the corresponding Unit 2 evaluation, PG&E used the Bechtel computer code BSAP.

In these evaluations the time history method was employed.

The integration procedure used in the BSAP code was the Nigam-Jennings type.

This method is based on the exact solution to the second order differen-tial equation of motion for a linear segment type of input.

Experience has shown that this procedure yields accurate results provided that the integration time interval is properly choosen.

As a general rule, the time intervals should be taken as a fraction of the period of interest.

The staff requested PG&E to show the effect of the time step interval on the results.

In response to the staff request, PG&E presented comparative studies in which response spectral curves were calculated'ith smaller time steps.

Floor response spectra were computed with low and high damping (i.e.,

2X and 7X).

The response spectra curves generated with the time step of 0.01 seconds were found to be in very good agreement with those computed with the smaller time step of 0.003 seconds.

n Based on the above, the staff concludes that no significant approximation is introduced in the pipeway structure response evaluations due to the time inter-val used in the analysis.

This conclusion applies to both units.

4.2.4 DE and DDE Evaluation Prior to the staff audit in January

1985, a particular structural evaluation had, not been performed for the pipeway structure of either unit with respect to the DE and DDE loading conditions.

PG&E explained this based on the results obtained from the Hosgri event and pipe rupture loading evaluations.

However, load combinations involving,DE and DDE may control the design of some'embers of the pipeway structure due to differences in the criteria (i.e.,

damping

values, stress allowables, modeling procedures) for the three earthquake eval-uations.

Thus, the staff requested PG&E to verify that the stresses in the pipeway structure satisfy the FSAR design criteria for loading conditions for the DE and DDE.

Subsequent evaluations for the pipeway structure of both units were performed by PG&E and reviewed by the staff.

The DE and DDE evaluations for Unit 1 and Unit 2 are documented in the following calculations:

1.

Calculation No. 2151C-2:

Diablo Canyon Unit 1, gualification of Unit 1 Pipeway for DE and DDE 2.

Calculation No. 1149C-1:

Diablo Canyon Unit 2, Pipeway Structure Unit 2, Evaluation and Acceptance of Selected Critical Members in the Pipeway Structures due to DE, DDE and Rupture Loading.

In these calculations the members of the pipeway structure were evaluated for load combinations involving dead load, DE, DDE, and pipe rupture restraint load.

The allowable stresses are the same for both units.

The objective of these calculations was to compute stress ratios (i.e.,

demand divided by capacity) at selected members of the pipeway structure in both units for a set of load com-binations including the DE and DDE loads.

Individual seismic analyses for DE and ODE were not performed to compute stresses in the pipeway structure.

In-

stead, they were computed by using spectral ratios of the DE and DDE spectra to the Hosgri spectra and the stress ratios based on ipformation from the de-tailed Hosgri seismic analyses.

Diablo Canyon SSER 32 4-4

The Unit 2 pipeway DE and DDE evaluations were performed for a set of 4 load combinations as committed in the FSAR.

A review of member stresses was per-formed for the dead load, Hosgri.and pipe rupture analyses.

As a result of this review two radial bents (2B -and 3B) were identified to contain critical members and were used in the DE and ODE evaluation.

Radial members of the pipe-way structure'were found to be=.generally more critical than tangential or ver-tical members.

Following the selection of bents 2B and 3B, spectral ratios were used to compute stress ratios for several members.

From the Unit 2 pipeway evaluation one structural member (elevation 108' 3")

in bent 2B reached the shear capacity for the load combination involving dead, OE and OOE and pipe rupture load.

However, some of this load should be redis-tributed to the adjacent members.

This situation was not found for bent 3B.

With respect to the DE and DDE evaluations for the Unit 1 pipeway structure, the computation of the stress ratios at six selected bents was based on conver-sion factors.

During the May 1985 audit, the staff requested a justification of the screening criteria used to establish these factors.

Furthermore, PG&E was requested to provide bounds with regard to the magnitude of the approxima-tion associated with the conversion factors.

Following the May 1985 audit, PG&E provided additional information which is based on Unit 1-pipeway specific data.

The latter calculations were found to be more appropriate than those presented previously.

Specifically, a larger number of structural members were incorporated into the DE and DDE evaluation procedure.

These members were selected on the basis of the Westinghouse results from the Hosgri evaluation.

According to the screening criteria used in the selection process, all members with stress ratios equal to or greater than 0.90 were used in the DE and DDE evaluation.

Furthermore, the original conversion factors are not utilized.

Instead, new ratios of DE and DDE to Hosgri were computed.

The values found for these ratios are less than one, except for three members (1103,

828, and 496).

However, the computed stress ratios were less than one for, the all members, including the above three members.

The maximum stress ratio found in this evaluation was 0.97.

Moreover, the stress ratios associated with members controlled by the pipe rupture load are generally less than those of the members for which the DE and DDE govern.

Based on the computed stress ratios the staff concludes that the pipeway members selected in the DE and ODE evaluations for Unit 1 satisfy the design allowables.

Based on the above, the staff concludes that, in general, the procedures used by PG&E in the DE and DDE evaluations of the Unit 1 and Unit 2 pipeway structures are acceptable.

The procedure used by PG&E to identify the critical members in the pipeway structure for the DE and DDE evaluation is based on the results of the Hosgri seismic analysis and involves various approximations.

The staff has determined that the approximations give reasonable results for the stresses in the members of the pipeway structure.

However, a further confirmatory anal-ysis is required to demonstrate the accuracy of the results.

4.3 Conclusions Based on the staff audits and review of the extensive information provided by PG&E as described

above, the staff concludes:

Diablo Canyon SSER 32

(1)

The modeling of the pipeway structure connections to the auxiliary and turbine buildings is acceptable.

(2)

The choice of the input to the pipeway structure model is not expected to affect the response results obtained for the pipeway structure.

Any input to the pipeway structure through the piping system is not expected to alter the results of the Hosgri evaluation.

(3)

The integration time step used in the Hosgri evaluations has no significant impact on the response computations.

(4)

The procedures used in the DE and'DDE evaluations are found to be generally acceptable.

However, the computations of the DE and DDE stresses for the pipeway structural members involve various approximations.

Results from individual seismic DE and DDE analyses are not presently available.

There-fore, the degree of the approximation should be further assessed.

Accord-ingly, the staff will include the following as a condition in the Unit 2 full-power license.

Pi ewa Structure DE and ODE Anal ses Prior',to start-up, following the first refueling outage, PG8E shall complete a

confirmatory analysis for the pipeway structure to further demonstrate the ade-quacy of the pipeway structure for load combinations that include the'esign earthquake and the double design earthquake.

Since the same approach in the DE and DDE evaluations was used for both Unit 1 and Unit 2 pipeway structures the conclusion from this confirmatory analysis should be applicable to both units.

Diablo Canyon SSER 32

5 ALLEGATIONS SSER 28 was issued in April 1985 to support the low-power licensing of Diablo Canyon Unit 2 (Ref. 2). It addressed the status of the staff review for both units as of March 31, 1985 for allegations that had been received as of March 1, 1985. 't that time 1665 allegations had been received, 1377 of which were resolved for Unit 1 and 1147 were resolved for Unit 2.

SSER 28 concluded that the staff effort regarding all 1665 allegations was sufficiently complete to conclude that none of the allegations, either individually or collectively, indicated problems of such magnitude that should preclude continued operation-of Unit 1 or full-power licensing of Unit 2.

The staff reaffirms the above conclusion at this time.

By letter dated March 14, 1985, the Government Accountability Project (GAP) submitted an amendment (Ref. 11) to its previous petition submitted pursuant to 10 CFR 2.206 (Ref. 12).

The affidavits contained in the March 14, 1985 submit-tal included 55 additional allegations.

In accordance with the Commission's Statement of Policy (50 FR 11030),

the staff has completed a screening review and has determined that none of these 55 allegations is of such significance as to preclude continued full-power operation for Unit 1 or issuance of a full-power license for Unit 2.

As of June 30, 1985, 1720 allegations have been received.

As of that date 1466 have been resolved for Unit 1 and 1395 have been resolved for Unit 2.

Addi-tional allegation information is presented in Table 5-1.

The resolutions of allegations completed since the issuance of SSER 28 have not yet been documented.

These resolutions, along with future resolved allegations, 'will be issued in another report.

In summary, the staff reaffirms its conclusion, as stated in SSER 28, regarding the 1665 allegations received as of March 1, 1985.

The staff has reached the same conclusion regarding the 55 allegations received in the Mar'ch 14, 1985 GAP submittal.

The staff is continuing its investigation, inspection and evalu-ation of those allegations not yet resolved.

Diablo Canyon SSER 32 5-1

Table 5-1 Allegation status as of June 30, 1985 with respect to Unit 2 Resolved Open

, Total Region V

1052 Office of Investigations Office of Inspector and Auditor Total 0

1395

"71 of these have been resolved for Unit l.

Office of Nuclear Reactor Regulation 329 31 1083 95" 424 197 211 325 1720 Diablo Canyon SSER 32 5-2

6 PHYSICAL SECURITY The staff safety evaluation for the" PG&E industrial security plan for Diablo Conyon Unit 2 was provided in SSER 31, Section 4.9.

A license condition was included in the Unit 2 low-power license DPR-81:

Ph sical Protection:

The Pacific Gas

& Electric Company shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans, including amendments made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p).

These plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled:

"Diablo Canyon Power Plant Physical Security Plan,"

with revisions submitted through August 30, 1984; "Diablo Canyon Power Plant Guard Training and gualification Plan," with revisions submitted through August 29, 1984; and "Diablo Canyon Power Plant Safeguards Contingency Plan,"

with revisions submitted through August 29, 1984.

This condition will likewise be included in the Unit 2 full-power license.

The approved security, contingency and guard training and qualification plans for the Diablo Canyon Plant consist of the following amendments, revisions and changes:

1.

Diablo Can on Power Plant Securit Plan Ref.

13 Revision 11 dated May 31, 1982 (PG&E letter May 27, 1982) as revised by:

June 21, 1982 (PG&E letter June 21, 1982)

August 12, 1982 (PG&E letter August 12, 1982)

February 4, 1983 (PG&E letter February 14, 1983)

April 19, 1984 (PG&E letter April 19, 1984)

August 29, 1984 (PG&E letter August 29, 1984)

August 30, 1984 (PG&E letter August 30,'984) 2.

Diablo Can on Power Plant Guard Trainin and uglification Plan (Ref.

14 Revision 1 dated July, 1980 (PG&E letter July ll, 1980) as revised by:

Revision 2 dated August 24, 1984 (PG&E letter August 29, 1984) 3.

Diablo Can on Power Plant Safe uards Contin enc Plan Ref.

15 Revision 1 dated May 1, 1980 (PG&E letter May 12, 1980) as revised by:

Revision 2 dated February 4, 1983 (PG&E letter February 4, 1983) and documents dated:

August 3, 1983 (PG&E letter August 3, 1983)

August 29, 1984 (PG&E letter August 29, 1983)

The above listed documents apply equally to Diablo Canyon Unit 1 and Unit 2.

Diablo Canyon SSER 32 6"1

7 EMERGENCY PREPAREDNESS

7. 1 FEMA Findin s

on Offsite Emer enc Plans and Pre aredness The staff addressed the Federal Emergency Management Agency (FEMA) findings on offsite emergency plans and preparedness in Section 4.23 of SSER 31.

The on-site plans and procedures have been reviewed by the NRC.

As stated in SSER 31 the offsite plans have been evaluated by FEMA (Ref. 16).

The NRC and FEMA have observed the exercises conducted to date for the Diablo Canyon facility; includ-ing the last one on October 30, 1984 (Ref.

47 and 48).

As discussed in SSER 31 a condition was included in the Unit 2 low-power license regarding the completion of procedures in the FEMA rule, 44 CFR Part 350.

This license condition will be included in the full-power license.

7.2 Offsite Emer enc Plannin for Medical Services In a recent decision, GUARD v.

NRC, 753 F.2d 1144 (D.C. Cir. 1985), the U.S.

Court of Appeals. vacated the Commission's interpretation of 10 CFR 550.47(b)(12) to the extent that a list of facilities was found to constitute adequate arrange-ments for medical services for members of the public offsite exposed to dangerous levels of radiation.

The Commission has now provided guidance to be followed in determining compliance with this regulation pending its determination of how it will proceed in response to the Court's remand.

In particular, the Commission directed that Licensing Boards, and in uncontested

cases, the staff, should con-sider the uncertainty attendant to the Commission's interpretation of this regu-lation, especially in regard to its interpretation of the term "contaminated injured individuals."

In GUARD, the Court left open to the Commission the dis-cretion to reconsider whether that term should include members of the offsite public exposed to dangerous levels of radiation and, thus, whether arrangements for this population of individuals are required at all.

For this reason, the Commission observed that it may reasonably be concluded that "no additional actions should be taken now on the strength of the present interpretation of

- that term."

Accordingly, the Commission observed that it can be found "that any deficiency which may be found in complying with a finalized post GUARD planning standard (b)(12) is insignificant for the purposes of 10 CFR 550.47(c)(1)."

In this regard, the Commission, as a generic matter, noted the low probability of accidents which might result in exposure of mem-bers of the offsite public to dangerous levels of radiation as well as the slow development of adverse reactions to over exposure.

See, Emergency Planning; Statement of Policy, 50 FR 20892, May 21, 1985.

Consistent with the foregoing Statement of Policy, PG8E has committed, by let-ter dated June 28, 1985 (Ref. 17), to fully comply with the Commission's response to the Court's remand.

By letter dated July 5, 1985 (Ref. 49)

FEMA confirmed that the emergency plans of the involved offsite response jurisdictions contain a list of medical service facilities.

Accordingly, on the basis of the factors identified by the Commission in its Statement of Policy, the staff has determined that the requirements of Diablo Canyon SSER 32

10 CFR 550.47(c)(1) have been satisfied so as to warrant issuance of the operating license pending further action by the Commission with respect to the requirements of 10 CFR 550.47(b)(12).

7.3 Conclusions On the basis of its review of the FEMA findings and determination on the ade-quacy of state and local emergency plans and preparedness, and based on the previous staff assessment of the adequacy of the PG&E onsite emergency plans and preparedness, the staff.concludes that the onsite and offsite emergency preparedness provides reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency at the Diablo Canyon Nuclear Power Plant.

Diablo Canyon SSER 32 7-2

8 TECHNICAL SPECIFICATIONS

8. 1 Introduction The Technical Specifications for Diablo Canyon Unit 1 full-power operation were issued on November 2, 1984, as Appendix A to the Unit 1 Operating License DPR-80 (Ref. 51).

The Technical Specifications for Diablo Canyon Unit 2 low-power operation were issued on April 26, 1985, as Appendix A to the Unit 2 Operating License DPR-81 (Ref. 52).

In May 1985 the staff identified a number of Unit 2 Technical Specifications that required further resolution for full-power opera-tion (Ref. 53).

The staff and PG&E have been pursuing the development of common and combined Technical Specifications for Unit 1 and Unit 2 since late 1984.

Diablo Canyon Unit 1 and Unit 2 are nearly identical.

There are some differences due to the mirror image, unit-specific layout, core thermal ratings, and equipment ratings.

The controls for both units are located in one common control room.

As stated in SSER 31, the operating staff will be licensed on both units, with a desig-nated crew assigned and dedicated to each unit on each shift.

PG&E concluded that the Technical Specifications for both units can and should be identical to a great extent.

Accordingly, PG&E proposed one set of Technical Specifications, common to both units, with unit-specific specifications due to differences in design and operations appropriately identified.

8.2 Unit 1 Tech S ec Considerations In January 1985 PG&E submitted License Amendment Request LAR 85-01 (Ref.

54) for Unit 1 to include the Unit 2 specific Technical Specifications information and make changes necessary for a two unit operation (i.e.,

minimum shift crew composition and diesel generator surveillance requirements).

The chan'ges neces-sary for two unit operation were issued on April 26, 1985 as Amendment 1 to the Unit 1 license (Ref. 66).

.PG&E subsequently submitted further changes to the Unit 1 Technical Specifications in Revisions 1 through 4 of LAR 85-01 to include Radiological Effluent Technical Specifications (RETS) and Diesel Fuel Oil Speci-fications, make typographical, editorial and similar type changes, improve con-sistency within the specifications, and to further update the specifications with common and unit-specific information (Ref.

55 through Ref. 58).

PG&E requested further changes to the Technical Specifications for both units in LAR 85-03 regarding movable control assemblies (Ref. 58),

LAR 85-04 regarding reactor trip, system instrumentation (Ref.

59) and LAR 85-06 regarding a reor-ganization of the PG&E Nuclear Power Generation organization (Ref. 61).

Finally, in LAR 85-05 (Ref.

60) and LAR 85-07 (Ref.

62 and 63)

PG&E requested Unit 1 Technical Specifications changes as a result of certain issues that had been raised by the staff with regard to full-power operation of Unit 2 (Ref. 53).

The above Technical Specifications changes for Unit 1 have been evaluated by the staff and are expected to be issued as an amendment to the, Unit 1 operating license DPR-80 at about the time of issuance of the Unit 2 full-power license.

These changes will revise the current Unit.1 Technical Specifications to be identical to those to be issued with the Unit 2 full-power license with unit-Diablo Canyon SSER 32 8-1

specific information clearly identified.

These combined and common Technical Specifications will apply to both units.

8.3 Unit 2 Tech S ec Considerations By letter dated May 14, 1985, PG&E submitted License Amendment Request LAR 85-02 (Ref.

58) which requested changes to the Unit 2 Technical Specifications (Ref.

52) to (1) remove typographical errors, provide further clarification, improve con-

, sistency, adjust nomenclature, and improve format and legibility and (2) include Unit 1 specific information and specifications as needed for a common and com-bined set of Technical Specifications for Units 1 and 2.

The staff has reviewed these proposed changes and concludes that they will improve the overall effective use of the specifications and will identify appropriate Unit 1 specific informa-tion in the common and combined full power Technical Specifications.

The staff addressed the matter of common and-combined Technical Specifications in a letter to PG&E, dated May 15, 1985 (Ref. 53).

In the letter the staff requested also further consideration'f 17 issues that, had been identified dur-ing the preparation of the Technical Specifications for the Unit 2 low-power license.

The staff met with PG&E on May 15, 1985 to discuss the issues (Ref. 70),

and PG&E provided additional information (Ref.

60, and 65).

The staff reviewed the information and discussed further these issues with PG&E.

As a result, these issues were resolved, appropriate Technical Specifications were developed, and will be included in the common and combined Technical Specifications for both units except as follows.

The final resolution for five specifications was achieved after the required Federal Register Notice for the Technical Specifications changes for the Unit 1 license amendment had been prepared (Ref.

71) as discussed in Section 8. 2.

The five specifications are (1) administrative controls for startup test reports (2)

~ reactor coolant system relief valves (3) loose-part detection system instrumentation (4) bases for electrical power systems (5) reactor coolant system pressure/temperature limits Rather than issuing the combined Technical Specifications with substantially different requirements for Unit 1 and Unit 2 the staff determined that for an interim period the current low-power Technical Specifications of Unit 2, as reflected in the combined Technical Specifications, provide, sufficiently con-servative requirements.

PG&E, by letter dated June 20, 1985 (Ref. 64), identi-fied the resolution to'he five specifications and committed to submit an appro-priate license amendment request for both units within 90 days after issuance of the Unit 2 full-power license.

The staff finds this acceptable.

As a result of its review of the Diablo Canyon Unit 1 and Unit 2 Technical Specifications, the staff determined that specifications should be included for the polar crane in the. containment buildings and for the two bridge cranes in the turbine building or appropriate ju'stification should be provided for not including them.

At this time, the parking location for the polar crane is restricted to pre-clude jet impingement from a postulated pipe rupture.. This restriction should Diablo Canyon SSER 32 8"2

be included in the specifications or PG&E shall demonstrate that the crane can withstand the jet impingement forces from a postulated pipe rupture.

The two bridge cranes in the turbine building are currently restricted by rail clamps from entering the end bays in the turbine building.

The restriction for placing the two cranes relative: to each other relies on visual observations imposed by administrative procedure.

This restriction should be included in the specificiations or PG&E shall demonstrate that no restriction is required.

The staff has requested PG&E to submit, in addition to the above five items, a

license amendment request within 90 days of issuance of the Unit 2 full-power license regarding these restrictions or provide a justification why they are not required (Ref. 77).

The staff finds acceptable, for the interim, the use of the current administrative procedures.

As stated in Section 8.2, PG&E requested further changes to the current Unit 1 and Unit 2 Technical Specifications regarding movable control assemblies (Ref. 58),

reactor trip system instrumentation (Ref.

59) and reorganization of the PG&E Nuclear Power Generation organization (Ref. 60).

The staff evaluation of these proposed

changes, which are applicable to both units, is presented in separate sections below.

8.4 Movable Control Assemblies This evaluation applies to Unit 1 and Unit 2.

8.4. 1 Introduction By letter dated May 14,

1985, PG&E requested in LAR 85-03 changes to the Techni-cal Specification

3. 1.3. 1, "Movable Control Assemblies" for Diablo Canyon Units 1 and 2 (Ref. 58).

The changes will allow continued operation for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (to perform repairs) for a condition in which there are multiple control rods which cannot be moved by their drives because of electrical power failure but which are still within alignment and trippable for insertion under gravity.

8.4.2 Evaluation The proposed Technical Specification changes contain Action Statements which address inoperable and misaligned moVable control assemblies as follows:

Action a. is unchanged and addresses a mechanical failure of one or more full length rods that leaves the rod(s) immovable and/or untrippable.

This is the classic stuck rod situation, requiring application of the stringent fix-or-shutdown requirement.

Action b.

has been deleted and replaced with the new Action c.

and d.

Action c.

has been relabeled to Action b. with the content unchanged and addresses one full-length rod trippable but inoperable due to causes other than addressed by Action a.

Action c.

A new Action c.

has been added to address multiple inoperable rods and replaces the old Action b.

The change now allows operation to continue for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (to perform repairs) for a condition in which Diablo Canyon SSER 32 8-3

there are multiple immovable rods that are still trippable and within alignment.

Action d.

has been added to replace that part of the old Action b. that addressed multiple misaligned rods.

,Although these Action statements have been realigned and relabeled, there is no alteration of the content and requirements of the present Technical Specifications except for the new Action c.

The present Technical Specifications require the reactor to be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for the condition of new Action c.,

which is when there are multiple inoperable rods which are not stuck or untrippable.

The intent and basis for Technical Specification

3. 1.3. 1 is to provide require-

'ments which either directly in the Specification or in support of other Specifi-cations ensure that:

. 1) acceptable power distribution limits are maintained,

, 2) the minimum shutdown margin is maintained with allowance for a stuck rod,

3) the potential effects of rod misalignment on associated accident analyses are limited, and 4) the trip reactivity assumed in the accident analysis will be available.

Appropriate provisions have been made in the proposed change. to ensure that these requirements continue to be maintained.

Specifically, the proposed change requires that within one hour after entering the Action state-ment the remainder of the rods in the bank(s) with the inoperable rods are aligned to within 2 12 steps of the inoperable rods.

This relates to ensuring conformance to items 1 and 3 above.

Further, it is explicitly specified (per-haps redundantly) that the rod sequence and rod insertion limit Technical Speci-fications must be maintained as applicable.

This ensures conformance with items 2 and 4 above.

The intent of the proposed Technical Specification change is to allow continued operation of the unit in the case of electrical failures which prevent moving of more than one control rod.

The rods remain trippable, so they can provide their safety function in shutting down the reactor, if needed.

The staff finds the proposed change acceptable

because, as indicated above, the safety requirements for the control rods remain intact.

Control of the power plant can still be effected with boration systems, if the inoperable rods are in the control bank or the rod sequence requirements potentially could not be maintained.

If the first regulating banks or the shutdown banks become inoper-able while inserted in the core, the insertion limit Specification requires the power level to be 0.

If the last two banks become inoperable while at part power the insertion limits allow operation only at part power.,

Operation at part power or at full power with the regulating bank at its insertion limit for extensive periods of time is not accounted for in the power distribution analy-

sis, which is assumed as an input condition for the LOCA analysis.

Therefore the time span of the proposed change is limited to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Normal operation of the power plant assumes all of the control rods are out of the core, except the regulating bank may be inserted a small amount to allow for control of the axial flux difference and day to day reactivity burnout (which may also be controlled by boron concentration).

In Attachment C to a letter dated December 21, 1984 from E.

P.

Rahe, Jr.

(Mestinghouse) to C.

0.

Thomas (NRC),

a description of the movable control assembly mechanisms is provided, along with a discussion of failures which Diablo Canyon SSER 32 8-4

might occur (Ref. 69).

The attachment provides guidance on the type of elec-trical failures which may render control assemblies inoperable but still trippable.

It will be used in the preparation of the bases of the Technical Specifications.

In preparing the proposed

change, PG&E also deleted the words "step counter" from the Unit 1 Technical Specifications to be consistent with Unit 2 and to allow other means for determining group demand position to satisfy this Specifi-cation.

This change is also acceptable because it does not alter the require-ment for the Specification, but only allows the possibility of other means for determining the group demand position.

8.5 Reactor Tri S stem Instrumentation This evaluation applies to Unit 1 and Unit 2.

8.5. 1

Background

On February 21, 1985, the NRC issued its Safety Evaluation Report on the Westinghouse Technical Specification Optimization Program for'increased sur-veillancee intervals and out-of-service times for testing and maintenance of the reactor trip system (RTS) (Ref. 67).

The Optimization Program proposal was set forth in WCAP-10271, "Evaluation of Surveillance Frequencies and Out-Of-Service Times for the Reactor Protection Instrumentation System" Supplement 1 (Ref. 68).

By letter dated May 20, 1985, PG&E Company submitted proposed Technical Specifi-cations as LAR 85-04 for Diablo Canyon Units 1 and Unit 2 based on the Optimiza-tion Program (Ref. 59).

8.5.2 Evaluation The proposed Technical Specification changes include an increase in the sur-veillance interval of reactor trip system channels from monthly to quarterly.

The proposed quarter ly surveillance intervals are annotated to indicate that the quarterly tests are to be performed on a staggered test basis.

In addition the surveillance intervals for those reactor trip sytem channels which also provide signals to the Engineered Safeguards Actuation System (ESFAS) have been annotated to indicate that the more restrictive ESFAS.surveillance requirements apply to these channels.

These changes are consistent with the staff's generic evaluation of the Optimization Program and are, therefore, acceptable.

The surveillance frequency for channels which would be tested prior to plant startup was noted as prior to startup in lieu of quarterly as recommended under the Optimization Program.

Also, the startup surveillance was annotated to indicate that, it is to be performed prior.to startup if not performed in the previous 31 days.

The channels for which testing prior to startup applies are:

A.

Power Ran e

Neutron Flux Low Set oint The low setpoint channels are required to be tested prior to entry into Mode 2

and for operation in Mode 1 below the P-10 interlock setpoint.

For these. tran-sitional operating

modes, the surveillance test would be conducted prior to startup and not on a routine basis as would be implied by specifying a quarterly Diablo Canyon SSER 32

test frequency.

Thus, the surveillance frequency is noted as prior to startup, S/U(1).

In the Table 4. 3-1 Notation, note (1) is revised to indicate if not performed in the previous 31 days, instead of 7 days.

B.

Intermediate Ran e

Neutron Flux The intermediate range channels are required to be tested for the same transi-tional modes as the power range low setpoint channels.

On 'the same basis the surveillance frequency is noted as prior to startup, rather than quarterly.

C.

Source Ran e

Neutron Flux The source range channels are required to be tested in Mode 2 below the P-6 interlock setpoint and in Modes 3, 4, and 5.

The source range channels, in addition to initiating reactor trip, provide the high flux alarms at shutdown'hich alert the operator of reactivity changes caused by a boron dilution event.

Therefore, the surveillance requirements for the source range channels apply to Modes 3, 4, and 5.

The surveillance frequency is noted as prior to startup and as quarterly.

D.

Reactor Tri S stem Interlock RTSI Intermediate Ran e Neutron Flux P-6 The P-6 interlock channels are required to be tested prior to entry into Mode 2 and for operation in this mode below the P-6 interlock setpoint.

For this transitional operating mode, the surveillance test would be conducted prior to startup and not on a routine basis as would be implied by specifying a quarterly test frequency.

Thus, the surveillance frequency is noted as prior to startup.

E.

RTSI Low Power Reactor Tri s Plock P-7 The surveillance frequency is noted as prior to startup, consistent with that specified for the P-10 and P-13 channels as noted below.

F.

RTSI Power Ran e Neutron Flux P-8 The currently specsfied monthly test interval is annotated to indicate that when the plant is at a power level greater than the channel trip setpoint, the surveillance requirement is satisfied by verifying that the interlock permissive logic is in its required state.

This provision was included in the Technical Specifications such that power reductions below the channel setpoint would not be required for the sole purpose of meeting the surveillance requirement.

For

example, the previous monthly surveillance requirement for the P-8 channels would require a power reduction below its setpoint had this annotation to the surveillance requirement not been included.

Likewise, this same annotation is used for the P-10 channels and precludes the necessity of power reductions for testing.

This annotation to the surveillance requirement only verifies the status of the permissive logic and does not address verification of channel setting or oper-ability.

Those aspects would be verified following a refueling shutdown and prior to startup.

With a monthly surveillance interval the annotated surveil-lance at power would in all likelihood have expired during a refueling shutdown (i.e.,

exceeded 31 days) and testing would be required prior to entry into Modes 2 or 1, as applicable to the permissive channels.

However, this situation

'4 Diablo Canyon SSER 32 8"6

may not be true if a quarterly surveillance interval is specified, i.e., the annotated surveillance performed at power may not have exceeded a 92 day sur-veillance interval during a refuelinq shutdown.

Therefore, since the only comprehensive tests are actually performed prior to startup, the surveillance interval is stated as prior to startup.

The status of interlock permissives at the logic and channel level are individ-ually indicated on status monitoring displays in the control room.

As such they are routinely checked and particular attention is given to this information during operational mode changes.

The fact that the permissive status indication is readily available and can be routinely verified constitutes a different con-sideration with respect to the availability of trip channels which must change state on the occurrence of an event and for which the function unavailability is dependent on the surveillance interval.

It is concluded that maintaining the requirement for the verification of the permissive logic status is not safety significant.

Therefore, since the surveillance frequency has been noted prior to startup, note (8) is no longer applicable to these channels.

G.

RTSI Low Set oint Power Ran e Neutron Flux P-10 The surveillance frequency is noted as prior to startup, on the same basis as for the power range P-8 channels noted above.

H.

RTSI Turbine Im ulse Chamber Pressure P-13 The survei11ance frequency is noted as prior to startup.

The bases with regard to note (8) is the same as for the power range interlock channels noted above.

However, unlike the power range channels which can only be tested when the measurement signal is below the setpoint, the turbine impulse chamber pressure channels include features which would permit them to be tested without reducing reactor power.

However, since the P-13 channels and hence P-13 logic only pro-vide inputs to the P-7 logic which are diverse to the P-10 inputs to the P-7 logic, it is concluded that there is sufficient justification for excluding testing as currently noted by. the provision on note (8) when operating above the P-13 setpoint.

Further, since the channel state is indicated, its state can be readily determined.

Thus, it is concluded that the additional surveil-lance as would occur by noting a quarterly surveillance frequency annotated with note (8) is not warranted.

By letter dated July 18, 1985, PGKE requested a change to exclude the require-ment to test the P-13 channels prior to startup from Modes 2 or 3 (Ref. 76).

A startup from Mode 2 (power less than 5 percent) or from Mode 3 (not standby) would occur only following plant operation at power and subsequent reduction in power to these modes.

Such plant star tup would be due to unusual circum-stances and not on a routine basis.

It is concluded that this change is accept-able since the additional testing which may result under these conditions would not have an impact on plant safety due to considerations related to the operability of interlocks noted above.

This exception to the startup tests is included as "Note 8" to the surveillance requirements.

Mith regard to the inclusion of the startup test, requirement, wherein channels would be tested prior to plant startup if not tested in the previous 31 days, the staff concludes that this is appropriate in view of the increase in the Diablo Canyon SSER 32 8-7

surveillance frequency from that currently required for startup tests if not performed in the previous 7 days.

I.

Out of Service Times for Testin and Maintenance For those RTS channels which provide input signals to ESFAS, the ACTION column of Table 3.3-1 is annotated, as applicable, with note (1) and the Table 3.3-1 notation includes note (1) to indicate that the applicable Modes and ACTION statement for these channels under the ESFAS specification requirements are applicable.

This is consistent with the staff s evaluation for the Optimization Program and is, therefore acceptable.

8.5.3 Conclusions The staff concludes that the proposed Technical Specifications are acceptable for Diablo Canyon Units 1 and 2.

8.6 GONPRAC Com osition PG&E requested in License Amendment Request LAR 85-06 of'May 30, 1985 (Ref. 61),

changes to the Technical Specifications for Diablo Canyon Units 1 and 2.

The LAR proposes to change the composition of the PG&E General Office Nuclear Plant Review and Audit Committee (GONPRAC) to reflect changes to the PG&E Nuclear Power Generation Department.

These organizational changes have been reviewed by the staff during the ful,l-power license review for Diablo Canyon Unit 2 and were found acceptable.

The proposed changes to the composition of the GONPRAC delete two members (i.e., Project Manager - Diablo Canyon, and Technical Assistant to the Vice President - Nuclear Power Generation) and add three new members (i.e.,

Manager

- Nuclear Engineering and Construction Services, Director - Nuclear Administra-tion and Support Services',

and Director - Nuclear Regulatory Affairs).

The staff finds that the proposed changes meet the acceptance criteria of'ection 13.4 of NUREG-0800, the Standard Review Plan, and concludes that the changes to the composition of the GONPRAC are acceptable.

8.7 Summar and Conclusions PG&E submitted a number of license amendment requests for changes to the Techni-cal Specifications for the Unit 1 full-power license and the Unit 2 low-power license in order to (1) establish common and combined Technical Specifications for,both units (2) make editorial and format changes (3) revise the specifications'for movable control rod assemblies, reactor trip system instrumentation, and PG&E Nuclear Power Generation organization, and (4)'esolve certain'full-power considerations that had been identified by the staff during the Unit. 2 low-power.licensing.

II Diablo Canyon SSER 32 8"8

Based on the staff review of the information provided by'- PG8E and based on further discussion with PG8E the staff has developed common and combined Technical Specifications for Units 1 and 2 that include the above considera-tions.

The Technical Specifications for six issues identified in Section 8.3 will be revised after issuance of the specifications.

I Diablo Canyon SSER 32 8"9

V l

k,j V

9 SMALL-BREAK LOCA (II.K.3.30 and II.K.3.31)

9. 1

~Summar The staff reported in SSER 31 on the status of its evaluation regarding the small-break LOCA methodology in accordance with the requirements of NUREG-0737, Item II.K.3.30 for the Westinghouse generic effort, and Item II.K.3.31 for the Diablo Canyon plant specific analysis.

As stated in SSER 31, PG8E is a member of the Westinghouse Owners Group (WOG) for this effort.

In May 1985 the staff completed its evaluation of the Westinghouse small-break LOCA model NOTRUMP, as discussed below and informed Westinghouse accordingly (Ref. 18).

This completes Item II.K.3.30 for Diablo Canyon Units 1 and 2.

By letter dated September 16, 1983 PG8E committed to submit its Diablo Canyon plant specific small-break LOCA analysis within one year of staff approval of the Westinghouse generic model (Ref. 19).

On November 2, 1983 the NRC provided clarification in Generic Letter 83-35 (Ref.

20) and proposed a generic resolution for Item II.K.3.31.

The staff requires that future plant specific analyses performed for Diablo Canyon by Westinghouse for reloads or Technical Specification amendments should be based on the NOTRUMP Code.

9.2 Introduction NUREG-0737 is a report transmitted by a letter from D.

G. Eisenhut, Director of the Division of Licensing, NRR, to licensees of operating power reactors and applicants for operating reactor licenses forwarding TMI Action Plan require-ments which have been approved by the Commission for implementation (Ref. 21).

Section II.K.3.30 of Enclosure 3 to NUREG-0737 outlines the Commission require-ments for the industry to demonstrate its small break loss of coolant accident (SBLOCA) methods continue'o comply with the requirements of Appendix K to 10 CFR Part 50.

The technical issues to be addressed were outlined in NUREG-0611,

-"Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in Westinghouse-Designed Operating Plants" (Ref. 22).

In addition to the concerns listed in NUREG-0611, the staff requested licensees with U-tube steam generators to assess their computer codes with the Semiscale S-UT-08 experimental results.

This request was made to validate the code's ability to calculate the core coolant level depression as influenced by the steam generators prior to loop seal clearing.

In response to TMI Action Item II.K.3.'30, the Westinghouse Owners Group (WOG) elected to reference the Westinghouse NOTRUMP code as their new licensing small break LOCA model (Ref.

23 and Ref.

24).

Referencing the new computer code did not imply deficiencies in WFLASH to meet the Appendix K requirements.

The decision was based on desires of the industry to perform licensing evaluations with a computer program specifically designed to calculate small break LOCAs with greater phenomenological accuracy than capable by WFLASH, The following docume'nts the staff evaluation of the WOG response to TMI Action Item II.K.3.30 confirmatory items.

Diablo Canyon SSER 32 9-1

9.3 Summar of Re uirements NUREG-0611 required licensees and applicants with Westinghouse NSSS designs to address the following concerns:

(a)

Provide confirmatory validation of the small break LOCA model to ade-quately calculate the core heat transfer and two-phase coolant level during core uncovery conditions.

(b)

Validate the adequacy of modeling the primary side of the steam generators as a homogeneous mixture.

(c)

Validate the condensation heat transfer model and affects of non-condensible gases.

(d)

Demonstrate, through noding studies, the adequacy of the SBLOCA model to calculate flashing during system depressurization.

(e)

Validate the polytropic expansion coefficient applied in the accumulator model.

(f). Validate the SBLOCA model with LOFT tests L3-1 and L3-7.

In addition, validate the model with the Semiscale S-UT-08 experimental data.

Detailed responses to the above items are documented in WCAP-10054, "Westing-house Small-Break ECCS Evaluation Model Using the NOTRUMP Code" (Ref.

23) and WCAP-10079, "NOTRUMP, A Nodal Transient Small-Break and General Network Code" (Ref. 24).

9.4 Evaluation The staff has reviewed and evaluated the information provided in the above topical reports in response to the above requirements.

The staff evaluation for each item is as follows:

(a)

Core Heat Transfer Models The Westinghouse Owners Group (WOG) referenced the NOTRUMP computer code as their new computer program for small-break loss of coolant accident (SBLOCA) evaluation.

NOTRUMP was benchmarked against core uncovery experiments con-ducted at the Oak Ridge National Laboratory (ORNL).

These tests were performed under NRC sponsorship.

The good agreement between the calculations and the data confirmed the adequacy of the drift flux model used for core hydraulics as well as the core heat transfer models of clad temperature predictions.

The staff finds the core thermal-hydraulic models in NOTRUMP acceptable.

This item is resolved.

(b)

Steam Generator Mixture Level Model NUREG-0611 requested licensees and applicants with Westinghouse designed NSSSs to justify the adequacy of modeling the primary system of the steam generators as a homogeneous mixture.

This question was directed to the WFLASH Code.

NOTRUMP, the new SBLOCA licensing code, models phase separation and incorporates flow regime maps within the steam generator tubes.

The adequacy of this model Diablo Canyon SSER 32 9-2

was demonstrated through benchmark analyses with integral experiments, in parti-cular with Semiscale Test S-UT-08.

The staff finds the steam generator model in NOTRUMP acceptable.

This item is resolved.

(c)

Noncondensible Affects on Condensation Heat Transfer 1

~

NUREG-0611 requested validation of the condensation heat transfer correlations in the Westinghouse SBLOCA model and an assessment of the consequences of non-condensible gases in the primary coolant.

The condensation heat transfer model used in NOTRUMP is based on steam experiments performed by Westinghouse on a 16-tube PWR steam generator model.

For two-phase conditions, an empirical cor-relation developed by Shah is applied.

The staff finds the condensation heat transfer correlation in NOTRUMP acceptable.

/

The influences of noncondensible gases on the condensation heat transfer was demonstrated by degrading the heat transfer coefficient in the steam generators.

The heat transfer degradation was calculated using a boundary layer approach.

For this calculation, the noncondensible gases generated within the primary coolant system were collected and deposited on the surface of the steam genera-tor tubes.

The sources of noncondensibles considered were:

(1) air dissolved in the RWST (2) hydrogen dissolved in the primary system (3) hydrogen in the pressurizer vapor space (4) radiolytic decomposition of water With a degradation factor on the heat transfer coefficient, the limiting SBLOCA was reanalyzed for a typical PWR.

The WOG, thereby, concluded that formation of noncondensible gases in quantities that may reasonably be expected for a 4-inch cold leg break LOCA presents no serious detriment on the PWR system re-sponse in terms of core uncovery or system pressure.

-What perturbation was observed was minor in'nature.

I The staff finds acceptable the Westinghouse submittal on the influences of non-condensible gases on design bases SBLOCA events.

The staff conclusion"is based on the limited amount of noncondensible gases available during a design basis SBLOCA event, as well as results obtained from Semiscale experiments which reached similar conclusions while injecting noncondensible gases in excess amount expected during a SBLOCA design basis event.

This item is resolved.

(d)

Nodalization Studies for Flashing During Depressurization As a consequence of the staff's experience with modeling SBLOCA events with NRC developed computer codes (in particular the TMI-2 accident),

the staff questioned the adequacy of the nodalization in the licensing model to calculate the depressurization of the primary system.

The staff therefore requested'vali-dation of the Westinghouse Evaluation Model to properly calculate the depres-surization expected during a SBLOCA event.

Through nodalization studies and validation of the NOTRUMP licensing model with integral experiments (e. g.,

LOFT and Semiscale),

Westinghouse demonstrated the acceptability of the nodalization and nonequi librium models.

The staff finds the Westinghouse model acceptable for calculating depressurization during SBLOCA events.

This item is resolved.

Diablo Canyon SSER 32 9"3

(e)

Accumulator Model WFLASH, the previous Westinghouse small-break loss of coolant accident (SBLOCA) analysis

code, applied a polytropic gas expansion coefficient of 1.4 to the nitrogen in the accumulators.

The WOG was requested to validate this accumula-tor model,in light of data obtained through the LOFT experimental programs for SBLOCAs.

Westinghouse reviewed the applicable LOFT data and determined the need to perform full scale accumulator tests.

Based upon these tests, Westinghouse modified the polytropic expansion coefficient to a more realistic value.

Of interest is Westinghouse's conclusion that the selection of either a high or low expansion coefficient had negligible effect on the calculated peak clad tempera-ture (PCT).

This insensitivity is only appropriate to'NOTRUMP, with its non-equilibrium assumptions.

The staff finds acceptable the polytropic expansion coefficient in the NOTRUMP code.

This item is resolved.

(f)

Code Validation Following the Three Mile Island event in 1979, staff analyses with NRC developed computer codes led to concerns that detailed nodalization was required to simu-late realistic systems responses to postulated SBLOCAs.

As a consequence, li-censees and applicants with Westinghouse plants were requested to validate their licensing tools with integral experiments.

Specifically, the NRC requested that the computer codes be validated with the LOFT L3-1 and L3-7 experimental data.

In addition, the staff also requested that the code be benchmarked with the Semiscale S-UT-08 experimental data.

Westinghouse performed the above benchmark analyses.

For the LOFT tests, Westinghouse showed good agreement between the NOTRUMP calculations and the experimental data.

For the S-UT-08 test, Westinghouse de'monstrated that NOTRUMP did a reasonable job calculating the experimental data.

However, this required a more detailed nodalization of the steam generators then used in the licensing model.

With the less detailed licensing nodalization, the pre-loop-seal-clearing core level depression phenomenon, as observed in the S-UT-08 data, was not conservatively calculated for very small breaks.

However, the calculated peak clad temperature was demonstrated to be higher (i.e.

more conservative) with the coarse nodalization.

The staff, therefore, finds acceptable the NOTRUMP computer code and the associated nodalization for SBLOCA design basis evaluation.

This item is resolved.

9.5 Conclusions The Westinghouse Owners Group (WOG), by referencing WCAP-10079 and WCAP-10054 (Ref.

23 and 24),

has identified NOTRUMP as the new thermal-hydraulic computer program for calculating small break loss of coolant accidents (SBLOCAs).

The staff finds acceptable the use of NOTRUMP as the new Westinghouse licensing tool for calculating SBLOCAs for Westinghouse NSSS designs.

The responses to NUREG-0611 concerns, as evaluated within this

SER, have also been found acceptable.

This SER completes the requirements of TMI Action Item II.K.3.30 for licensees and applicants with Westinghouse NSSS designs who were members of the WOG and referenced WCAP-10079 and WCAP-10054 as their response to this item.

Diablo Canyon SSER 32 9-4

Mithin one year of receiving this SER, the licensee is required to submit plant specific analyses with NOTRUMP, as required by THI Action Item II.K.3.31.

Per generic letter 83-35, compliance with Action Item II.K.3.31 may be submitted generically.

The staff requires that the generic submittal include validation that the limiting break location has not shifted away from the cold legs to the hot or pump suction legs.

Diablo Canyon SSER 32 9-5

ll

10 OTHER MATTERS In addition to those matters addressed in separate sections of this report the staff also identified in SSER 31 a number of, matters that required completion prior to fuel loading, criticality or power ascension.

The status and staff evaluation of these items is presented below.

In addition, SSER 31 addressed other matters with a longer term completion requirement for which the PG&E effort and/or staff review was still ongoing.

The status of some of those is also presented.

10. 1 Seismicall Induced S stems Interactions In SSER 31, Section 4.2 the staff reported on the PG&G seismically induced sys-tems interaction program (SISIP).

The PG&E SISIP Final Report (Ref.

29 and Ref.

30) was updated by letters dated April 22 and May 16,

1985, (Ref.

31 and Ref.

32) which documented the results of the SISIP for Unit 2.

The interaction data sheets (IDS) for Unit 2 will be submitted later.

PG&E provided a status of the Unit 2 modifications (Ref.

33 and Ref 34) and by letter dated July 9, 1985 confirmed that all modifications have been completed (Ref. 28).

This completes the staff requirement regarding plant modifications as stated in Section 8.2 of SSER 11.

On June 6, 1985 the NRC Office of Inspection and Enforcement issued IE Informa-tion Notice No. 85-45:

"Potential Systems Interaction Involving the Movable In-Core Flux Mapping System Used in Westinghouse Designed Plants" (Ref. 35).

The notice identifies a potential interaction, caused by a seismic event, be-tween the nonsafety related portions of the in-core flux mapping system as the interaction source and the instrumentation tubing seal table as the interaction target.

As stated in the Notice, multiple failures of the flux mapping tubing and/or fittings constitute an unanalyzed small break loss of coolant accident.

PG&E advised the staff that this SISI was identified in March 1982 during the program and is included in the SISIP Final Report (Ref.

29 through 32).

Modi-fications, including verification walkdowns, were completed in February 1984 for Unit 1 and June 1985 for Unit 2.

PG&E provided detailed information on this interaction (Ref. 36).

The staff audited the PG&E documentation and in-spected the modifications made for Unit 2 (Ref. 37).

Based on the information provided by PG&E and the staff audit, the staff concludes that the PG&E modifi-cations are appropriate and acceptable and resolve this systems interaction.

10.2 Re ulator Guide 1.97 - Emer enc Res onse Ca abilit In SSER 31, Section 4.5 the staff found acceptable viding instrumentation for postaccident monitoring accordance with Regulatory Guide 1.97.

By letters July 18, 1985 (Ref. 34, 50, and 54, respectively),

the instrumentation identified in SSER 31 has been the PG&E schedule for pro-of certain variables in dated June 6 and 21, and PG&E informed the staff that installed and fully tested.

In its schedule of January 25, 1985 (Ref. 75),

PG&E identified the radioactiv-ity concentration monitor for circulating primary coolant and the boric acid charging flow works as two items that are currently not commercially available Diablo Canyon SSER 32 10-1

as qualified equipment.

In its submittal of July 18, 1985 (Ref. 74),

PG8E committed to install these items during the first refueling outage following that availability.

The staff finds this acceptable.

The staff safety evaluation of masonry walls for Unit 2 was provided in Sec-tion 4.7 of SSER 31 (April 1985).

The staff required that necessary modifica-tions be completed prior to full-power operation.

In addition, as discussed in SSER 31, a condition was included in the Unit 2 low-power license regarding the verification of the energy-balance technique, prior to start-up following the first refueling outage.

PGLE informed the staff that the required masonry wall modifications have been completed (Ref.

33 and 34).

During staff audits in October and November 1984 (Ref.

25 and Ref.

26) at the PG8E office in San Francisco and the plant site, the staff reviewed documents related to the masonry wall construction to determine if:

(a) the documentation is sufficient to fulfillthe requirements of "special inspection" for masonry walls, and (b) the masonry walls are constructed in accordance with the design documents.

The construction documents included material test reports, photographs and con-struction logs'f five walls.

In addition, the staff also discussed the con-struction and inspection practices used in the masonry wall construction with the construction personnel involved in those activities.

The staff requested PG&E to further undertake an extensive survey of the construction documents and report to the staff its conclusions regarding the above two objectives.

By letter dated May 23, 1985 (Ref. 27),

PG8E provided details of its investi-gation.

These details indicate the following:

(1)

The construction records provide the details of construction for various walls located in the turbine and auxiliary buildings.

They document the various stages of construction, such as hauling rein-forcing steel and blocks to the construction area, placing of rein-forcing steel and blocks, filling block cells with grout, and pro-viding drypack at the top of the wall.

I (2)

The construction logs indicate that proper communication channels were established between construction and engineering.

The inspector assigned for blockwall construction contacted engineers on the project for approval of changes in design and documentation.

(3)

Blocks, in-fill grout, and reinforcing steel were tested for compli-ance with the appropriate standards during construction of a majority of the block walls.

Since the walls were not classified as Class I at the time of construction, test results for some of the walls are not available.

However, many test results are available and they indicate that the construction materials met or exceeded specification requirements.

Diablo Canyon SSER 32 10-2

PGLE further noted that during the implementation of recent wall modifications it was necessary to locate the reinforcing steel in the masonry walls to avoid cutting the bars.

This effort has also confirmed that the reinforcing steel was provided in accordance with the design drawings.

(During the site visit in November 1984, the staff observed the marked locations of the reinforcing bar on faces of walls.)

Based on the above PGKE findings and, based on the results of its audits, the staff concludes that the masonry walls at the Diablo Canyon plant meet the intent of "sp'ecial inspection" requirements as specified in "Code Requirements for Concrete Masonry Structures" of the American Concrete Institute (ACI 531-79) and there is a reasonable assurance that they were constructed in accordance with the design documents.

The remaining issue regarding the masonry wall at Diablo Canyon Unit 1 and Unit 2 is the license condition as discussed in SSER 31.

The staff will report its activities regarding the license condition in a future report.

10.4 Generic Letter 83-28:

Re uired Actions Based on Generic Im lications of Salem ATWS Events The staff reported on the status of the PG8E actions regarding Generic Letter 83-28 (GL 83-28) in Section 4.0 of SSER 31.

PG8E'has completed the'equired actions for high priority items (Ref. 38).

The completion of all items in accordance with the PG&E schedule (Ref.

39 and Ref. 40) has been made a Unit 2 low-power license condition (Ref.

6) which will be retained in the Unit 2 full-power license.

The staff has requested additional information, by letter dated June 20, 1985 (Ref. 41), regarding the following GL 83-28 items:

(1)

Item 2.2. 1:

Equipment Classification (2)

Item 2. 2. 2:

Vendor Interface (3)

Item 3. 2. 3:

Post Maintenance Testing (4)

Item 4.5.3:

On-Line Functional Testing Intervals The staff has completed its evaluation of the following items and has found acceptable the respective actions by PG8E:

(1)

Item 1. 1 Post-Trip Review (Program Description and Procedure)

(Ref. 42)

(2)

Item 1.2 Post-Trip Review (Data and Information Capability) (Ref. 43)

(3)

Item 4.1 Reactor Trip System Reliability (Vendor-Related Modifications)

(Ref. 44)

(4)

Item 4.2. 1:

Periodic Maintenance (Ref. 45)

(5)

Item 4.2.2:,,

Trending of Parameters (Ref.

45)

(6)

Item 4.3 Reactor Trip System Reliability (Shunt Trip Attachment)

(Ref; 46)

Diablo Canyon SSER 32 10-3

(7)

Item 4.5. 1:

On-Line Functional Testing (Ref.. 44)

PG8E advised the staff that its review of the remaining items is on schedule.

The staff evaluation of other GL 83-28 items is ongoing.and the results will be

'eported in future transmittals to PG8E.

10.5 E ui ment uglification 10.5. 1 Seismic and Dynamic gualification In SSER 31 Section

4. 11.1 the staff, stated that the RCP upper bearing cooler and the spray additive tank were yet to be seismically qualified by completing modifications prior to fuel load.

PG8E has completed these, modifications as stated in a letter dated April 25, 1985 (Ref. 33).

PG&E committed to complete, prior to fuel load, the as-built piping analyses and necessary modifications resulting from nozzle load differences and to confirm the as-built configura-tions of safety-related electrical and mechanical equipment.

The staff required that PG8E complete and maintain all pertinent documentation of the Unit 2 quali-fication program in an auditable form in a central filing system.

By letter dated July 9, 1985, (Ref. 28),

PG8E informed the staff that the documentation has been completed in the PG8E San Francisco office.

10.5.2 Environmental gualification

, In SSER, 31 Section 4. 11. 2 the staff concluded that the environmental qualifica-tion program is acceptable based on the PG8E commitment that the surveillance and maintenance program be implemented prior to full-power operation.

The pro-gram is being fully applied and implemented for each system as it is turned over from the construction to the operations organization.

By letter dated July 9, 1985, (Ref. 28)

PG8E informed the staff that the, program is expected to be implemented by July 31, 1985, i.e., prior to full-power operation.

The staff requires that PG8E inform the staff accordingly.

10.6 Fire Protection In SSER 31 Section 9.6.1.21 the staff found acceptable the PG8E commitment to complete the Appendix R modifications and conduct an emergency lighting walkdown prior to Unit 2 fuel load.

By letter dated April 25, 1985 (Ref.

33)

PG8E informed the staff that these actions have been completed.

The staff also found acceptable the PG8E commitment to make any necessary modifications to the Ruskin fire damp-ings prior to fuel load.

By letter dated June 6, 1985 (Ref. 34),

PG&E informed the staff that the modifications had been completed prior to fuel load.

This resolves the required actions identified in SSER 31.

The l.icense condition on fire protection, as discussed in SSER 31 Section 9.6. 1.2.22 and included in the Unit 2 low-power license, will also be included in the Unit 2 full-power license.

10.7 Plant Readiness The staff reported in SSER 31 on the Diablo Canyon Unit 2 readiness regarding fuel loading and low-power operation.

The low-power license DPR-81 includes in Attachment 1 to License Condition C.(l) the requirement that PG&E complete (1) the Preoperational Test 38.4, "Radiation Monitoring System" and (2) resolve Diablo Canyon SSER 32 10"4

work items as listed on the PG&E Master Completion List.

The NRC Region V

office review of the PG&E adherence and completion of construction, maintenance and preoper ational testing as identified in Attachment 1 to Operating License DPR-81 indicates that PG&E has complied fully with the first requirement and

.continues

'to comply with the second requirement.

Items on the Master Completion List are reviewed and tracked by the NRC Resident Inspector to verify completion prior to mode changes.

Items required prior to entering Mode 5 and Mode 6 have been satisfactorily completed.

PG&E informed the staff that items for Modes 1, 2, 3, and 4 are pending in accordance with the startup schedule (Ref. 28).

Items for Mode 1 will be completed prior to exceeding 5 percent power.

There-fore, the staff concludes that the license conditions in Attachment 1 are no longer'equired and will not be included in the full-power license.

., 10.8 Post Accident Sam lin S stem II.B.3)

In SSER 31, Section 4. 19 the staff found acceptable the Unit 2 post accident sampling system (PASS) regarding the requirements in NUREG-0737 for Item II.B.3, based on the PG&E commitment that the system will be fully operational prior to Unit 2 criticality and maintenance and surveillance procedures will be imple-mented.

By letters of April 25 (2 letters) and June 6, 1985 (Ref. 72, 33 and 34, respectively)

PG&E informed the staff that the PASS is operable as demon-strated by (1) initial demonstration of functionality, (2) procedures for opera-tion, (3) training of technicians, (4) acceptance criteria for sampling, and (5) maintenance capability and surveillance procedures.

The staff concludes that the requirements for Item II.B.3 in NUREG-0737 have been met.

10.9 Inade uate Core Coolin Instrumentation S stem II.F.2 In SSER 31, Section

4. 10 the staff reported on its evaluation of the inadequate core cooling instrumentation system (ICCIS) for Diablo Canyon Units 1 and 2

regarding the requirements in Item II.F.2 of NUREG-0737.

The staff identified certain actions for both units to be completed by PG&E in accordance with a specified schedule.

The following is an update of these actions.

(a)

Subcoolin Mar in Monitor SMM One single channel SMM each is installed for Unit 1 and Unit 2 which must be

operable, in accordance with the Technical Specifications.

Two backup methods to manually calculate the subcooling margin are available to the operator as discussed in SSER 31.

This is acceptable.

(b)

Core Exit Thermocou le S stem CET As a result of the environmental qualification of the junction boxes and con-nectors for the CET, modifications were required for Units 1 and-2.

These modifications were completed for Unit 1 during the April 1985 outage (Ref. 50) and for Unit 2 prior to fuel loading (Ref. 34).

This completes the required action for the CET.

(c)

Reactor Vessel Level Instrumentation S stem RVLIS The RVLIS for Unit 1 has been installed, tested and calibrated as stated in SSER 31.

The system has been installed for Unit 2, and will be tested and Diablo Canyon SSER 32 10-5

calibrated after fuel loading and during power ascension (Ref.

33, 34, 73).

PG8E shall notify the staff when the RVLIS is 'declared fully operational.

The current Technical Specifications for Unit j. do not include the RVLIS, while for Unit 2 the system is included in Table 3. 1-10 of the specifications, with two channels required and one channel as the minimum operable.

PG8E requested the same specifications for Unit l in License Amendment Request LAR 85-01, Rev.

4 (Ref. 58).

The change request will be included in the common and combined Tech Specs for'oth units and is acceptable to the staff.

PG8E has revised and implemented the ICC procedures (Ref. 33).

The staff finds this acceptable.

In summary, the action items identified in SSER 31 have been completed by PG8E and are acceptable to the staff.

Final calibration of the RVLIS will be com-pleted during power ascension.

Diablo Canyon SSER 32 10-6

11 REFERENCES U.S. Nuclear Regulatory Commission, October 12, 1974, "Diablo Canyon, Units 1 and 2 Safety Evaluation Report."

2.

3.

5.

6.

7.

8.

9 ~

10.

12.

13.

14.

U.S. Nuclear Regulatory Commission, April 1985, "Diablo Canyon, Units 1

and 2 Safety Evaluation Report, Supplement 28,"

SSER 28.

U.S. Nuclear Regulatory Commission, March 1985, "Diablo Canyon, Units 1 and 2 Safety Evaluation Report, Supplement 29,'"

SSER 29.

U.S. Nuclear Regulatory Commission, April 1985,- "Diablo Canyon, Units 1 and 2 Safety Evaluation Report, Supplement 30,"

SSER 30.

U.S.

Nuclear Regulatory Commission, April 1985, "Diablo Canyon, Units 1 and 2 Safety Evaluation Report, Supplement 31,"

SSER 31.

U.S. Nuclear Regulatory Commission, issued April 26, 1985, "Diablo Canyon Unit 2 Low Power License DPR-81."

PG&E letter OCL-85-206, June 6, 1985, "Diablo Canyon Unit 2, Turbine Building Operating Deck - Final Report."

PG&E letter DCL-85-137, April 3, 1985, "Diablo Canyon Units 1 and 2, NRC Staff Audits of Pipeway Structure."

PG&E letter DCL-85-158, April 19, 1985, "Diablo Canyon Units 1 and 2, Pipeway Structures."

PG&E letter DCL-85-207, June 10, 1985, "Diablo Canyon Units 1 and 2, Pipeway Structure - Final Report."

Government Accountability Project (GAP), March 14,

1985, Supplement to Petition Pursuant to 10 CFR 2.206 regarding Diablo Canyon Nuclear Power Plant, Units 1 and 2.

Government Accountability Project (GAP), July 27, 29, 30 and 31, 1984, and November 15, 1984, letters and submittals regarding Diablo Canyon Nuclear Power Plant, Units 1 and 2, Petition and Supplements Pursuant to 10 CFR 2.206.

"Diablo Canyon Power Plant Security Plan,"

PG&E letters dated May 27, 1982, June 21, 1982, August 12, 1982, February 14, 1983, DCL-84-149, April 19, 1984, DCL-84-294, August 29, 1984, and DCL-84-297, August 30, 1984.

"Diablo Canyon Power Plant Guard Training and gualification Plan,"

PG&E letters dated July ll, 1980, and DCL-84-294, August 29, 1984.

Diablo Canyon SSER 32

15.

16.

17.

18.

19.

20 ~

21.

22.

23.

24.

"Diablo Canyon Power Plant Safeguards Contingency Plan,"

PG&E letters dated May 12,

1980, February 4, 1983, August 3, 1983, and DCL-84-294, August 29, 1984.

FEMA Memorandum from R.

W.

Krimm (FEMA) to E.

L. Jordan (NRC), July 11, 1984, "Interim Finding on State of California Nuclear Power Plant Emergency

Response

Plan (CNPPERP)"

(see also NRC letter, August 16,

1984, same subject).

PG&E letter DCL-85-231, June 28, 1985, "Diablo Canyon Unit 2, Compliance with Emergency Planning Standard 10 CFR 50.47(b)(12)."

U.S.

Nuclear Regulatory Commission letter from C.

C.

Thomas (NRC) to E.

P.

Rahe (Westinghouse),

May 21, 1985, "Acceptance for Referencing of Licensing Topical Report WCAP 10054(P) - Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code."

PG&E letter, September 16, 1983, "Diablo Canyon Units 1 and 2 Clarifica-tion of Status of Compliance with License Condit'ions 2 AC(8) 1(2) and 2.C.(8)0."

Generic Letter 83-35, November 2, 1983, "Clarification of TMI Action Plan Item II.K.3.31."

NUREG-0737, October 1, 1980, "Clarification of TMI Action Plan Requirements."

NUREG-0611, January 1, 1980, "Generic Evaluation of Feedwater Transients and Small Break Loss of Coolant Accidents in Westinghouse-Designed Operat-ing Plants."

Westinghouse Report WCAP 10054, December 22,

1982, "Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code."

Westinghouse Report WCAP 10079, November 12,

1982, "NOTRUMP, A Nodal Tran-sient Small Break and General Network Code."

/

25.

U.S.

Nuclear Regulatory Commission letter from G.

W. Knighton (NRC) to J.

D. Shiffer (PG&E), November 29, 1984, "NRC Trip/Audit Report-Unit 2."

26.

27.

U.

S. Nuclear Regulatory Commission letter from G.

W. Knighton (NRC) to J.

D. Shiffer (PG&E), January 3, 1985, "NRC Staff Audits."

PG&E letter DCL 85-189, May 23, 1985, "Diablo Canyon Units 1 and 2-Masonry Walls."

28..PG&E letter DCL-85-234, July 9, 1985, "Diablo Canyon Unit 2 - Activities Requiring Completion Prior to Full Power Operation."

29.

PG&E letter DCL-84-172, May 7, 1984, "Diablo Canyon Units 1 and 2-Seismically Induced Systems Interaction Program."

30.

PG&E letter DCL-84-319, October 2, 1984, "Diablo Canyon Units 1 and 2-Seismically Induced Systems Interaction Program."

Diabl o Canyon SSER 32 11-2

31.

32.

33.

34.

'5.

36.

37.

38.

39.

40.

41.

42.

43.

44.

45.

PG&E letter DCL 85-162, April 22, 1985, "Diablo Canyon Units 1 and 2-Seismically Induced Systems Interaction Program."

PG&E letter DCL 85-186.;

May 16, 1985, "Diablo Canyon Units 1 and 2-Seismically Induced Systems Interaction Program."

PG&E letter DCL-85-168, April25, 1985, "Diablo Canyon Unit 2.- Readiness for Operation."

PG&E letter DCL-85-204, June 6, 1985, "Diablo Canyon Unit 2 - Activities Completed Prior to Fuel Load."

U.S. Nuclear Regulatory Commission, IE Information Notice 85-45, June 6, 1985, "Potential Seismic Interaction Involving the Movable In-Core Flux Mapping System Used in Westinghouse Designed Plants."

PG&E letter DCL-85-242, July 18, 1985, "IE Information Notice No. 85-45:

Potential Seismic, Interaction Involving the Movable In-Core Flux Mapping System Used in Westinghouse Designed Plants."

J U.S. Nuclear Regulatory Commission, memorandum F.

Coffman to A. Thadani, July 22, 1985, "Trip Report - July 9, 1985 Audit of Systems Interactions at Diablo Canyon-2."

PG&E letter DCL-85-151, April 18, 1985, "Diablo Canyon Units 1 and 2:

Generic Letter 83 Additional Information."

PG&E letter DCL-85-025, January 24, 1985, "Diablo Canyon Units 1 and 2:

Updated Status and Schedule to Generic Letter 83-28 and Additional Informa-tion,"

PG&E letter DCL-85-105, March 13, 1985, "Diablo Canyon Units 1 and 2:

Generic Letter 83-28, Items 4. 2. 3 "and 4. 2. 4, Status Report on Breaker Life Cycle Testing."

U.S. Nuclear Regulatory Commision letter from G.

W. Knighton (NRC) to J.

D. Shiffer (PG&E), June 20,

1985, "Generic Letter 83-28 Request for Additional Information."

U.S.

Nuclear Regulatory Commission'letter from G.

W. Knighton (NRC) to J.

D. Shiffer (PG&E), May 15, 1985, "Diablo Canyon Units 1 and 2-Generic letter 83 Item 1. l."

U.S. Nuclear Regulatory Commission letter from G.

W. Knighton (NRC) to J.

D. Shiffer (PG&E), June 25, 1985,=Generic letter 83-28, Item 1.2."

U.S. Nuclear Regulatory Commission letter from G.

W. Knighton (NRC) to J.

D. Shiffer (PG&E), July 8, 1985, Generic letter 83-28, Items 4. 1 and 4.5.1."

U.S. Nuclear Regulatory Commision letter from G.

W. Knighton (NRC) to J.

D. Shiffer (PG&E), June 24, 1985, "Generic Letter 83-28, Items 4.2.1 and 4.2.2."

Diablo Canyon SSER 32 11-3

46.

47.

48.

49.

50.

51.

52.

53.

55.

56.

57.

58.

59.

60.

U.S. Nuclear Regulatory Commission letter from G.

W. Knighton (NRC) to J.

0. Schuyler (PG&E), May 16, 1984, "Automatic Shunt Trip for Scram Breakers."

FEMA memorandum from R.

W.

Krimm (FEMA) to E.

L. Jordan (NRC), December 17, 1984, "Exercise Report for the October 30, 1984, Exercise of the Offsite Radiological Emergency Preparedness (REP) Plans for the Diablo Canyon Nuclear Power Plant."

U.S. Nuclear Regulatory Commission letter from F.

A. Wenslawski (NRC) to J.

D. Shiffer (PG&E), November 29, 1984, Inspection Report No. 84-29.

FEMA memorandum from R.

W.

Krimm (FEMA) to E.

L. Jordan (NRC), July 5, 1985, "Medical Services Information for the Diablo Canyon Nuclear Power Station."

PG&E letter DCL-85-224, June 21, 1985, "Diablo Canyon Units 1 and 2,"

Activities Completed."

NUREG-1102, November 1984, "Technical Specifications, Diablo Canyon Nuclear Power Plant, Unit, No. l."

NUREG-1132, April 1985, "Technical Specifications, Diablo Canyon Nuclear Power Plant, Unit No. 2."

U.S. Nuclear Regulatory Commission letter from H.

L. Thompson, Jr.

(NRC) to J.

D. Shiffer (PG&E), May 15, 1985, "Diablo Canyon Technical Specifications."

PG&E letter DCL-85-028, January 29, 1985, "Diablo Canyon Unit 1 - License Amendment Request 85-01, Technical Specification Changes."

PG&E letter DCL-85-148, April 12, 1985, "Diablo Canyon Unit 1 - License Amendment Request 85-01, Revision 1, Radiological Environmental Monitoring."

PG&E letter DCL-85-164, April 24, 1985, "Diablo Canyon Unit 1 - License Amendment Request 85-01, Revision 2, Technical Specifications (Applicable to Radiological Effluents)."

PG&E letter DCL-85-184, May 14, 1985, "Diablo Canyon Unit 1 - License Amendment Request 85-01, Revision 3, Diesel Fuel Oil Specification."

PG&E letter DCL-85-185, May 14, 1985, "Diablo Canyon Units 1 and 2-License Amendment Requests:

85-01, Revision 4, Unit 1 Technical Specifica-tions; 85-02, Unit 2 Full Power Technical Specifications; and 85-03, Units 1 and 2, Movable Control Rod Assemblies."

PG&E letter DCL-85-187, May 20, 1985, "Diablo Canyon Units 1 and 2-License Amendment. Request 85-04, Technical Specification 3/4.3.1, Reactor Trip System Instrumentation."

PG&E letter DCL-85-197, May 30, 1985, "Diablo Canyon Unit 1 - License Amendment Request 85-05, Technical Specification Changes Sections 6.8. 1,

6. 9.1. 2, 4. 3. 3. 3. 1, 3. 3. 3. 6, 3.7. l. 1, and 4. 7.7. 1. "

Diablo Canyon SSER 32 ll"4

61.

62.

63.

64.

65.

66.

67.

68.

69.

70.

71.

72.

73.

75.

PGEE letter DCL-85-198, May 30, 1985, "Diablo Canyon Units 1 and 2-License Amendment Request 85-06, Technical Specification Change Section 6.5.2.2, GONPRAC Composition."

PG&E letter DCL-85-199, May 31, 1985, "Diablo Canyon Unit 1 - License Amendment Request 85-07, Technical Specification

Changes, Specifications 3.8.2.1 and 3.8.2.2."

PGKE letter DCL-85-214, June 14, 1985, "Diablo Canyon Unit 1 - Supplemental Information to License Amendment Request 85-07."

PGLE letter DCL-85-221, June 20, 1985, "Diablo Canyon Units 1 and 2-Additional Information Regarding the Combined Units 1 and 2 Technical Specifications.,"

PG8E letter DCL-85-188, May 21, 1985, "Diablo Canyon Units 1 and 2-Combined Technical Specifications."

U.S.

Nuclear Regulatory Commission, April 26, 1985, "Amendment No.

1 to License No.

DPR-80, Diablo Canyon Unit 1, Docket No. 50-275."

U.S.

Nuclear. Regulatory Commission, February 21, 1985, letter from C. 0.

Thomas (NRC) to J.

J.

Sheppard (W Owners Group).

Westinghouse Report WCAP 10271, Supplement 1, 'July 1983, "Evaluation of Survellance Frequencies and Out-Of-Service Times for the Reactor Protec-tion System."

Westinghouse letter, December 21,

1984, from E.

P.

Rahe, Jr.

(W) to C.

0.

Thomas (NRC) re specification and bases for movable control rod assemblies.

U.S. Nuclear Regulatory Commission memorandum, May 30, 1985 from H. Schierling (NRC), "Diablo Canyon Units 1 and 2 - NRC/PG8E Meeting on May 15, 1985."

U.S. Nuclear Regulatory Commission, letter from H. Schierling (NRC) to J.

D. Shiffer (PG8E),

June 14, 1985, "Issuance of Notices of Considera-tion of Issuance of Amendments."

PG8E letter DCL-85-169, April 25, 1985, "Diablo Canyon Unit 2, Operability of SENTRY PASS - NUREG-0737, Item II.B.3."

PG8E letter 'DCL-85-155, "Diablo Canyon Units 1 and 2, NUREG-0737, Item II.F.2 - Instrumentation for Detection of Inadequate Core Cooling / Unit 2 Implementation Report."

PG8E letter DCL-85-243, July 18, 1985, "Diablo Canyon Unit 2, Additional Information on Equipment gualification and Regulatory Guide 1.97 - Activi-ties Completed."

PG8E letter DCL-85-024, January 25, 1985, "Diablo Canyon Units 1 and 2, Regulatory Guide 1.97 Compliance."

Diablo Canyon SSER 32 11-5

76.

PG&E letter DCL-85-244, July 18, 1985, "Diablo Canyon Units 1 and 2, Cer-tification of Diablo Canyon Units 1 and 2 Full Power Technical Specifica-tions."

77.

U.S. Nuclear Regulatory Commission, Letter from G.W. Knighton (NRC) to J.D. Shiffer (PG8E), July 27, 1985, "Technical Specification for Cranes."

<U.S.

GOVERNMENT PRINTING OPPICEi1985+61-721i20260 Diablo Canyon SSER 32 11-6

NRC FORM 33S 12.84 1 NRCM 1102, 3201, 3202 SEE INSTRUCTIONS ON THE REVERSE.

2. TITLE AND SVBTI'TLE U.S. NUCLEAR REGULA'TORY COMMISSION BIBLIOGRAPHIC DATA SHEET

1. REPORTNUMBER JAssipntddt TIDC tdd Vos. No.,iltnyl NUREG-0675 Supplement No.

32

3. LEAVE BI.ANK Safety Evaluation Report Related to the Operation of Diablo Canyon Nuclear Power'Plant-,

Units 1 and 2

4,DATE REPORT COMPLFTEO S. AUTHOR(SJ MONTH August YEAR 1985

8. DATE REPORT ISSUED MONTH August YEAR 1985 2, PERFORMING ORGANIZATIONNAME ANO MAILINGADDRESS Ilntsodt Zrp Codtl Division of Licensing Office of Nuclear Reactor Regulation

.,U.S. Nuclear Regulatory Commission Washington, D.C.

20555 IO, SPONSORING ORGANIZATIONNAME ANO MAILINGADDRESS Jlntlodt Zip CodtJ

8. PROJECT/TASKJWORK UNIT NUMBER B. FIN OR GRANT NUMBER 114, TYPE OF JIEPORT Same as 7.

above Safety Evaluation Report Is. PERIOD CO V FRED llntlossrt dtstsl

12. SUPPLEMENTARY NOTES Docket Nos.

50-275 and 50-323 13, ABSTRACT J200words OrltuJ Supplement 32 to the Safety Evaluation Report for the application by Pacific Gas and Electric Company for licenses to operate Diablo Canyon Nuclear Power Plant, Units 1

and 2 (Docket Nos. 50-275/323) has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission.

This supplement provides the staff evaluation of those matters that require an appropriate resolution prior to full-power operation of Unit 2 and updates previous supplements to the Safety Evalu-ation Report.

14,DOCUMENT ANALYSIS 4, KEYWORDS/DESCRIPTORS

15. AVAILABILITY S'TATEMENT B. IDENTIFIERSIOPEN ENDED TERMS Unlimited

18. SECURITYCLASSIFICATION JTnrs ptttl Unclassified IFtnS rtPOrli Unclassified I 7, NVMBFR OF PAGES 18 PRICE

i U

f I'